RS-15-011, Proposed Alternative to the Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
| ML15030A175 | |
| Person / Time | |
|---|---|
| Site: | LaSalle  |
| Issue date: | 01/29/2015 |
| From: | Gullott D Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| RS-15-011 | |
| Download: ML15030A175 (15) | |
Text
Exelon Generation AS-15-011 January 29, 2015 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-11 and NPF-18 NRC Docket Nos. 50-373 and 50-374
<1300W111f eld Ro3d Warren\\11 I<? IL 60555 630 657 2000 Of11ce 1 o CFR 50.55a
Subject:
Proposed Alternative to the Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1}, Exelon Generation Company, LLC (EGC), is requesting relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," for LaSalle County Station (LSCS), Units 1 and 2. Specifically, Relief Request 13R-14 proposes an alternative to the examination requirements contained in Table IWB-2500-1 concerning nozzle-to-vessel welds and nozzle inner radii sections. The details of the 10 CFR 50.55a request are provided in the Attachment.
The attached relief request evaluated the proposed alternative and concludes that it provides an acceptable level of quality and safety.
The Federal Register Notice (FAN) published November 5, 2014, contains the rulemaking that amends 10 CFR 50.55a to incorporate by reference Regulatory Guide (AG) 1.147, Revision 17, "tnservice Inspection Code Case Acceptability, ASME Section XI, Division 1." As stated in the FAN, licensees may use the Code Cases listed in AG 1.147 as altematives to engineering standards for the construction, inservice inspection, and inservice testing of nuclear power plant components. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR)
Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," is listed in AG 1.147, Table 2, "Conditionally Acceptable Section XI Code Cases." The Condition associated with Code Case N-702 is as follows:
The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ML13071A240) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case.
January 29, 2015 U. S. Nuclear Regulatory Commission Page2 In the section of the FAN associated with NRC Responses to Public Comments on Draft Regulatory Guides, the NRC responses to comments specific to Code Case N-702 start on page 9 of 40 (79 FR 65783). An excerpt from the FRN is included as follows:
Licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, licensees should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by addressing the conditions and limitations specified in Section 5.0 of the NRC Safety Evaluation for BWRVIP-241.
Therefore, based on the above statements, EGC is requesting relief from the ASME Code,Section XI requirements, and is including information to support that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 are met.
This relief applies to the third 10-year lnservice Inspection interval. The third interval for LSCS, Units 1 and 2, began on October 1, 2007, and will conclude September 30, 2017. The third 10-year lnservice Inspection interval complies with the ASME B&PV Code,Section XI, 2001 Edition through 2003 Addenda.
EGC requests approval of this request by January 29, 2016, to support the LSCS Unit 1 spring 2016 refueling outage (L 1 R16). There are no regulatory commitments contained within this letter.
Should you have any questions concerning this letter, please contact Ms. Lisa A. Simpson at (630) 657-2815.
Respecttully, David Gullott Director - Licensing & Regulatory Affairs Exelon Generation Company, LLC
Attachment:
10 CFR 50.55a Relief Request 13R-14 cc:
NRC Regional Administrator, Region Ill NRC Senior Resident Inspector, LaSalle County Station Illinois Emergency Management Agency-Division of Nuclear Safety
ATTACHMENT 1 O CFR 50.55a Relief Request 13R* 14 12 pages follow
10 CFR 50.55a RELIEF REQUEST 13R-14 (Page 1 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1), Exelon Generation Company, LLC (EGC), is requesting relief from the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," for LaSalle County Station (LSCS), Units 1 and 2. Specifically, Relief Request 13R-14 proposes an alternative to the examination requirements contained in Table IWB-2500-1 concerning nozzle-to-vessel welds and nozzle inner radii sections.
1.0 ASME CODE COMPONENTS AFFECTED Code Class:
1
Reference:
ASME Section XI, Table IWB-2500-1 Examination Category:
8-D (Inspection Program B)
Item Number:
83.90 and 83.100 Component Numbers:
Reactor Vessel Nozzles: N1, N2, N3, N5, N6, N7, NB, N9, N16, and N18 (See Enclosure 1 for complete list of nozzle identifications) 2.0 APPLICABLE CODE EDITION ANO ADDENDA The third 10-year lnservice Inspection Program at LSCS, Units 1 and 2, is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV)
Code,Section XI, 2001 Edition through the 2003 Addenda. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2001 Edition is implemented, as required and modified by 10 CFR 50.55a(b)(2)(xv).
3.0 APPLICABLE CODE REQUIREMENTS The applicable requirement is contained in Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzle in Vessels - Inspection Program B.w Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number 83.90, "Nozzle-to-Vessel Welds," and 83.100, "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles are examined each interval.
All of the nozzle assemblies identified in Enclosure 1 are full penetration welds.
10 CFR 50.55a RELIEF REQUEST l3R-14 (Page 2 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1) 4.0 REASON FOR REQUEST The Federal Register Notice (FAN) published November 5, 2014, contains the rulemaking that amends 1 O CFR 50.55a to incorporate by reference Regulatory Guide (AG) 1.147, Revision 17, "lnservice Inspection Code Case Acceptability, ASME Section XI, Division 1.'
1 As stated in the FAN, licensees may use the Code Cases listed in AG 1.147 as alternatives to engineering standards tor the construction, inservice inspection, and inservice testing of nuclear power plant components. Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," is listed in RG 1.147, Table 2, "Conditionally Acceptable Section XI Code Cases." The Condition associated with Code Case N-702 is as follows:
The applicability of Code Case N-702 must be shown by demonstrating that the criteria in Section 5.0 of NRC Safety Evaluation regarding BWRVIP-108 dated December 18, 2007 (ML073600374) or Section 5.0 of NRC Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ML13071A240) are met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRC prior to the application of the Code Case.
In the section of the FAN associated with NRC Responses to Public Comments on Draft Regulatory Guides, the NRC responses to comments specific to Code Case N-702 start on page 9 of 40 (79 FR 65783). An excerpt from the FAN is included as follows:
Licensees who plan to request relief from the ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, licensees should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by addressing the conditions and limitations specified in Section 5.0 of the N RC Safety Evaluation for BWRVI P-241.
The proposed alternative provides an acceptable level of quality and safety based on the technical content of BWRVIP-108 and BWRVIP-241, as endorsed by the NRC SEs.
5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE In accordance with 10 CFR 50.55a(z)(1 ), relief is requested from performing the required examinations on 100 percent of the nozzle assemblies identified in Tables 5-1 and 5-2 below (see Enclosure 1 for a list of RPV Examination Category 8-D Nozzles tor which this relief request is applicable). As an alternative for all welds and inner radii identified in Tables 5-1 and 5-2, EGC proposes to examine a minimum of 25 percent of the LSCS, Units 1 and 2, nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with Code Case N-702.
10 CFR 50.55a RELIEF REQUEST 13R*14 (Page 3 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
For the nozzle assemblies identified in Enclosure 1, this would mean 25 percent from each of the groups identified in Tables 5-1 and 5-2.
B 0 N I S a e 5-.
mt xammation ategorv.
ozze ummarv Group Total Minimum Number Comments*
Number to be Examined Reactor Recirculation Two (2) nozzles inspected Outlet (N1) 2 1
this interval; No rejectable indications Reactor Recirculation Three (3) nozzles Inlet (N2) 10 3
inspected this interval; No rejectable indications Main Steam Four (4) nozzles (N3) 4 1
inspected this interval; No rejectable indications Core Spray Two (2) nozzles inspected 2
1 this interval; (NS and N16)
No rejectable indications Nozzles On Top Head No nozzles have been 3
1 inspected this interval; (N7, N8 and N18)
No rejectable indications Jet Pump Instrument Two (2) nozzles inspected 2
1 this interval; (N9)
No rejectable indications Residual Heat Removal Three (3) nozzles (N6) 3 1
inspected this interval; No rejectable indications
- The nozzle-to-vessel weld and inner radius exams are performed together.
10 CFR 50.55a RELIEF REQUEST 13R-14 (Page 4 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
Table 5*2: LSCS, Unit 2 RPV Examination Cateaorv B*D Nozzle Summarv Group Total Minimum Number Comments" Number to be Examined Reactor Recirculation Two (2) nozzles Outlet (N1) 2 1
inspected this interval; No rejectable indications Reactor Recirculation One ( 1) nozzle inspected Inlet {N2) 10 3
this interval; No rejectable indications Main Steam Four (4) nozzles (N3) 4 1
inspected this interval; No rejectable indications Core Spray No nozzles have been 2
1 inspected this interval; (N5 and N16)
No rejectable indications Nozzles On Top Head No nozzles have been 3
1 inspected this interval; (N7, N8, and N18)
No rejectable indications Jet Pump Instrument No nozzles have been 2
1 inspected this interval; (N9)
No rejectable indications Residual Heat Removal Three (3) nozzles
{N6) 3 1
inspected this interval; No rejectable indications
- The nozzle-to-vessel weld and inner radius exams are performed together.
The exams in Tables 5-1and5-2 will be scheduled in accordance with Section XI, IWB-2412, Inspection Program B.
Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item No. 83.100, "Nozzle Inside Radius Section"). EGC may utilize Code Case N-648-1 with associated RG 1.147 conditions for the nozzles selected for examination. Volumetric examinations of the inside radius section of those reactor vessel nozzles selected for examination will be completed tf Code Case N-648-1 is not applied.
10 CFR 50.55a RELIEF REQUEST 13R-14 (Page 5 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
Electric Power Research Institute (EPRI) Technical Report (TA) 1003557, "BWAVIP-108:
Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," provides the technical basis for Code Case N-702.
BWRVIP-108 determined that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure (LTOP) event are very low (i.e., <1 x 1 o.e for 40 years) with or without lnservice Inspection. The report concluded that inspection of 25 percent of each nozzle type is technically justified. The BWRVIP-108 report was approved by the NRC in a Safety Evaluation (SE) dated December 19, 2007 (i.e., ADAMS Accession No. ML073600374), and requires additional criteria to be met in order to apply the technical basis of BWRVIP-108 for the reduction of inspection coverage of the RPV nozzles and nozzle-to-vessel shell welds.
BWRVIP-108 was supplemented by EPRI TR 1021005, "BWRVIP-241: Boiling Water Reactor Vessel and Internals Project Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and was approved by the NRC in a SE dated April 19, 2013 (i.e., ADAMS Accession No. ML13071A240). This report revised the acceptance criteria associated with the NRC additional criteria.
As stated in the BWRVIP-241 NRC SE, Section 5.0, "Conditions and Limitations," each licensee who plans to request relief from ASME Code,Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radii sections may reference the BWRVIP-241 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to its units in the relief request by demonstrating that the following general and nozzle-specific criteria are satisfied:
(1)
The maximum RPV heatup/cooldown rate is limited to less than 115 °F/hour.
LSCS, Units 1 and 2, Technical Specifications (TS) 3.4.11, "Reactor Coolant System (RCS) Pressure and Temperature (Pff) Limits," provides a Surveillance Requirement limiting heatup and cooldown rates to s 100 °F in any one-hour period. This heatup/cooldown rate is also described in the LSCS Updated Final Safety Analysis Report (UFSAR),
Section 5.2.3.3.1.7, "Operating Limits During Heatup, Cooldown and Core Operations."
10 CFR 50.55a RELIEF REQUEST 13R*14 (Page 6 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
For the recirculation inlet nozzles (N2), the following criteria must be met:
(2)
(pr/t)/CRPV S 1.15 p = RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t = RPV wall thickness (inch), and CRPV = 19332; The calculation for the LSCS, Units 1 and 2, N2 Nozzle results in a maximum value of 1.064, which satisfies this criteria.
(3}
[p(ro2 + ri2)/(ro2
- ri2)]/CNoZZLE S 1.47 p = RPV normal operating pressure (psi},
ro =nozzle outer radius (inch},
ri =nozzle inner radius (inch), and CNOZZLE = 1637:
The calculation for the LSCS, Units 1 and 2, N2 Nozzle results in a maximum value of 1.134, which satisfies this criteria.
For the Recirculation Outlet Nozzles (N1), the following criteria must be met:
(4)
(pr/t)/CAPV S 1.15 p = RPV normal operating pressure (psi),
r = RPV inner radius (inch),
t = RPV wall thickness (inch), and CRPV = 16171; The calculation for the LSCS, Units 1 and 2, N1 Nozzle results in a value of 1.025 for Unit 1 and a value of 1.272 for Unit 2. The Unit 1 results satisfy the criteria; however, the Unit 2 results are greater than 1.15.
(5)
[p(ro2 + ri2)/(ro2
- ri2)]/CNOZZLE S 1.59 p = RPV normal operating pressure (psi),
ro =nozzle outer radius (inch),
n = nozzle inner radius (inch), and CNOZZLE = 1977.
The calculation for the LSCS, Units 1 and 2, N1 Nozzle results in a maximum value of 1.114, which satisfies the criteria.
Based upon the above information, all LSCS RPV nozzle-to-vessel shell or head full penetration welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles on Unit 2, meet the general and nozzle-specific criteria in BWRVIP-241.
10 CFR 50.55a RELIEF REQUEST 13R*14 (Page 7 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
BWRVIP-241 Section 6.0 notes that for plants having recirculation outlet nozzles with Condition 4 greater than 1.15, such as for LSCS, Unit 2, a plant-specific analysis following the approach described in this report may be able to justify values greater than 1.15.
Because the Unit 2 N1 nozzles did not meet the BWRVIP-241 criteria, a bounding analysis was performed to qualify all the Unit 1 and Unit 2 nozzles. This analysis is contained in Design Analysis L-003976, "Probability of Failure Analysis for Reactor Pressure Vessel Nozzles." The methods approved in BWRVIP-108 and BWRVIP-241 were applied, using the U2 N1 nozzle as the limiting nozzle geometry. A finite element model was developed, and the stresses caused by thermal transients and internal pressure were determined. The thermal transients evaluated were those associated with the LSCS reactor vessels, and the bounding transients were chosen based on their temperature ranges and rate of change. The resultant through-wall stresses at the locations of interest were used in a Probabilistic Fracture Mechanics (PFM) calculation.
The Probability of Failure (PoF) was calculated based on operation for 60 years and assumes no inspections were performed in the initial 40 years of operation. Although the current licenses for LSCS, Units 1 and 2, expire after 40 years, the use of the 60-year duration provides additional margin in the analysis.
To address the elevated fluence issue of certain nozzles in the belt-line region of the reactor vessel, the fluence associated with the Unit 2 N6 nozzle (Low Pressure Coolant Injection nozzle) at the end of 60 years of operation was used as input. This beltline nozzle has the highest fluence of all the Unit 1 or Unit 2 nozzles, and although the current licenses for LSCS, Units 1 and 2, expire after 40 years, the use of the 60-year fluence provides additional margin in the analysis.
The VIPERNOZ computer program, as used in BWRVIP-108 and BWRVIP-241, was used in the LSCS analysis. The same assumptions used in BWRVIP-108 and BWRVI P-241 were used in the LSCS analysis, such as the assumed number of stress corrosion initiation and fabrication flaws, the flaw size distribution, etc.
The bounding load cases analyzed included the following:
- 1. Unit pressure
- 2. Turbine Generator Trip-SCRAM
- 3. Loss of Feedwater Pumps/Isolation Valves Close The number of thermal cycles used in the analysis was based on the LSCS reactor pressure vessel thermal cycle diagrams.
The results of the analysis are shown in the following table.
10 CFR 50.55a RELIEF REQUEST l3R*14 (Page 8 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
PoF per year from L TOP PoF per year from Normal events for 25% lnservice Operating Condition for 25%
Maximum Inspection for period of lnservice Inspection for period PoF Extended Operation (Zero of Extended Operation (Zero per year**
Inspection for initial 40 years)'"
Inspection for initial 40 years)
Nozzle 1.4E-9 4.2E-7 5.0E-6 Blend Radii Nozzle-to-
<<2.0E-10 3.3E-9 5.0E-6 shell weld
- Values include 1 E-3 probability of L TOP occurrence.
- Technical Basis for Revision of Pressurized Thermal Shock (PTS) Screening Limit in the PTS Rule (10 CFR 50.61), NUREG-1806, Volume 1, August 2007.
Therefore, since the bounding nozzle, the N1 on Unit 2, has been shown to meet the NRC safety goal of 5E-6 per year, and all other nozzles meet the plant specific applicability criteria from the BWRVIP-241 report and are bounded by the Unit 2 N1 nozzle analysis, the application of Code Case N-702 to all the Unit 1 and Unit 2 nozzles listed in Enclosure 1 is acceptable.
6.0 DURATION OF PROPOSED ALTERNATIVE The third interval for LSCS, Units 1 and 2, began on October 1, 2007, and will conclude September 30, 2017. The proposed alternative will be used for the remainder of the third 10-year interval of the LSCS lnservice Inspection Program.
7.0 PRECEDENT
- 1)
Letter from M. Khanna (U. S. Nuclear Regulatory Commission) to M. J. Pacilio (EGC), "Peach Bottom Atomic Power Station, Units 2 and 3 - Requests for Relief 14A-51 and 14R-52 (TAC Nos. ME5392, ME5393, ME5394 and ME5395), dated January 24, 2012 (ADAMS Accession No. ML112770217)
- 2)
Letter from H. K. Chernoff (U. S. Nuclear Regulatory Commission) to M. J. Pacilio (EGC), "Limerick Generating Station, Units 1 and 2-Proposed Alternative Request RA-13R-14, Nozzle-to-Vessel Weld and Inner Radii Examinations (TAC Nos. ME3306 and ME3307), dated September9, 2010 (ADAMS Accession No. ML102390467)
- 3)
Letter from S. J. Campbell (U. S. Nuclear Regulatory Commission) to C. G. Pardee (EGC), "Dresden Nuclear Power Station, Units 2 and 3 -
Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations (TAC Nos.
ME0882 and ME0883), dated November 3, 2009 (ADAMS Accession No. ML092940436)
10 CFR 50.55a RELIEF REQUEST 13R*14 (Page 9 of 12)
Request for Relief for Alternate Examination Requirements for Nozzl&-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
E I 1 A r bl LSCS U "t 1 N nc osure. 1001ca e m
ozzes Component Category Item Nominal System Pipe ID Number Number Size N1A Nozzle B-D 83.90 Recirc Outlet 24" N1AIR*
8-D 83.100 Recirc Outlet 24" N1B Nozzle B*D 83.90 Recirc Outlet 24" N1BIR B-D 83.100 Recirc Outlet 24" N2A Nozzle 8-D B3.90 Recirc Inlet 12" N2AIR 8-D B3.100 Recirc Inlet 12" N2B Nozzle 8-D 83.90 Recirc Inlet 12" N2BIR 8-D B3.100 Recirc Inlet 12" N2C Nozzle 8-D B3.90 Recirc Inlet 12" N2CIR 8-D B3.100 Recirc Inlet 12 11 N2D Nozzle 8-D B3.90 Recirc Inlet 12" N2DIR 8-D 83.100 Recirc Inlet 12" N2E Nozzle 8-D 83.90 Recirc Inlet 12" N2EIR 8-D 83.100 Recirc Inlet 12" N2F Nozzle B-D 83.90 Recirc Inlet 12" N2FIR 8-D B3.100 Recirc Inlet 12" N2G Nozzle B-D 83.90 Recirc Inlet 12" N2GIR B-D 83.100 Recirc Inlet 12" N2H Nozzle 8-D 83.90 Recirc Inlet 12" N2HIR B-D 83.100 Recirc Inlet 12" N2J Nozzle B-D 83.90 Recirc Inlet 12" N2J IA B-D 83.100 Recirc Inlet 12" N2K Nozzle B-D 83.90 Recirc Inlet 12" N2KIR B-D 83.100 Recirc Inlet 12" N3A Nozzle B-D 83.90 Main Steam 26" N3AIR 8-D 83.100 Main Steam 26" N3B Nozzle B*D 83.90 Main Steam 26" N381R B-D B3.100 Main Steam 26" N3C Nozzle 8-D 83.90 Main Steam 26" N3CIR 8-D 83.100 Main Steam 26" N3D Nozzle 8-D 83.90 Main Steam 26" N3DIR B-D 83.100 Main Steam 26" NS Nozzle B*D 83.90 Core Sorav 12" NSIR B-D 83.100 Core Sprav 12" N6A Nozzle B*D 83.90 LPCI**
12" N6AIR B-D 83.100 LPCI 12" N68 Nozzle B-D 83.90 LPCI 12" N6BIR B-D 83.100 LPCI 12" N6C Nozzle B*D 83.90 LPCI 12"
10 CFR 50.55a RELIEF REQUEST 13R-14 (Page 10 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
Component Category Item Nominal System Pipe ID Number Number Size N6CIR B-D 83.100 LPCI 12" Reactor Core N7 Nozzle B-0 83.90 Isolation 6"
Coolinq Reactor Core N71R B-0 83.100 Isolation 6"
Coolinq NS Nozzle 8-D 83.90 Head Vent 4"
NSIA B-0 83.100 Head Vent 4"
N9A Nozzle 8-D 83.90 Jet Pump 6"
Instrumentation N9AIR 8-D 83.100 Jet Pump 6"
Instrumentation N98 Nozzle 8-D B3.90 Jet Pump 6"
Instrumentation N981R 8-D 83.100 Jet Pump 6"
Instrumentation N16 Nozzle 8-D B3.90 Core Sorav 12" N161R 8-0 83.100 Core Spray 12" N18 Nozzle 8-0 B3.90 Spare 6"
N181R 8-D 83.100 Soare 6"
- IA - Inner Radius
10 CFR 50.55a RELIEF REQUEST 13R-14 (Page 11 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a{z)(1)
E I
1 A I'
bl LSCS U *t 2 N nc osure :
1DD'ICa e n1 ozzes Component Category Item Nominal System Pipe ID Number Number Size N1A Nozzle B-D 83.90 Recirc Outlet 24 11 N1AIR*
8-D 83.100 Recirc Outlet 24 11 N18 Nozzle 8-D 83.90 Recirc Outlet 24" N1BIR B-D 83.100 Recirc Outlet 24 11 N2A Nozzle B-D 83.90 Recirc Inlet 12 11 N2AIR 8-D 83.100 Recirc Inlet 12" N28 Nozzle B-D B3.90 Recirc Inlet 12 11 N2BIR 8-0 83.100 Recirc Inlet 12 11 N2C Nozzle 8-0 83.90 Recirc Inlet 12" N2CIR 8-D B3.100 Recirc Inlet 12" N2D Nozzle 8-D 83.90 Recirc Inlet 12" N201R 8-D 83.100 Recirc Inlet 12" N2E Nozzle 8-0 83.90 Recirc Inlet 12" N2EIR 8-D 83.100 Recirc Inlet 12" N2F Nozzle B-0 83.90 Recirc Inlet 12" N2FIR B-0 83.100 Recirc Inlet 12 11 N2G Nozzle 8-D 83.90 Recirc Inlet 12" N2GIR 8-0 83.100 Recirc Inlet 12" N2H Nozzle 8-D 83.90 Recirc Inlet 12" N2HIR 8-D 83.100 Recirc Inlet 12" N2J Nozzle 8-0 83.90 Recirc Inlet 12" N2J IR 8-0 83.100 Recirc Inlet 12" N2K Nozzle B-D 83.90 Recirc Inlet 12" N2KIR B-0 83.100 Recirc Inlet 12" N3A Nozzle 8-0 B3.90 Main Steam 26" N3AIR 8-D B3.100 Main Steam 26" N38 Nozzle B-0 83.90 Main Steam 26
N381R 8-0 B3.100 Main Steam 26" N3C Nozzle 8-0 83.90 Main Steam 26" N3CIR 8-D 83.100 Main Steam 26" N30 Nozzle 8-D 83.90 Main Steam 26 11 N3DIR 8*0 83.100 Main Steam 26" N5 Nozzle 8-0 83.90 Core Sorav 12" NSIR 8-0 83.100 Core Sorav 12" N6A Nozzle 8-0 83.90 LPCI**
12 11 N6AIR 8-D B3.100 LPCI 12" N68 Nozzle B-0 83.90 LPCI 12
N681A B-0 83.100 LPCI 12" N6C Nozzle 8-D B3.90 LPCI 12"
10 CFR 50.55a RELIEF REQUEST 13R*14 (Page 12 of 12)
Request for Relief for Alternate Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)
Component Category Item Nominal System Pipe ID Number Number Size N6CIR B-D 83.100 LPCI 12" Reactor Core N7 Nozzle B-D 83.90 Isolation 6"
Coolino Reactor Core N71R 8-D 83.100 Isolation 6"
CoolinQ N8 Nozzle 8-0 83.90 Head Vent 4"
NSIR 8-0 83.100 Head Vent 4"
N9A Nozzle 8-0 83.90 Jet Pump 6"
Instrumentation N9AIR B-0 83.100 Jet Pump 6"
Instrumentation N98 Nozzle 8-0 83.90 Jet Pump 6"
Instrumentation N9BIR 8-D 83.100 Jet Pump 6"
Instrumentation N16 Nozzle B-0 83.90 Core Sorav 12" N161A B-D 83.100 Core Sprav 12" N18 Nozzle 8-0 83.90 Spare 6"
N18 IR B-0 83.100 Soare 6"
- IA - Inner Radius