RS-20-072, Supplemental Information Regarding Relief Request I4R 02 for the Fourth Inservice Inspection Interval

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Supplemental Information Regarding Relief Request I4R 02 for the Fourth Inservice Inspection Interval
ML20174A530
Person / Time
Site: Clinton Constellation icon.png
Issue date: 06/22/2020
From: Simpson P
Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-20-072
Download: ML20174A530 (12)
v • d • e


Text

4300 Winfield Road Warrenville, IL 60555 630 657 2000 Office RS-20-072 10 CFR 50.55a June 22, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Clinton Power Station, Unit 1 Facility Operating License No. NPF-62 NRC Docket No. 50-461

Subject:

Supplemental Information Regarding Relief Request I4R 02 for the Fourth Inservice Inspection Interval

Reference:

Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, "Relief Requests Associated with the Fourth Inservice Inspection Interval," dated December 16, 2019 In the Reference, in accordance with 10 CFR 50.55a, "Codes and standards," paragraph (z)(1),

Exelon Generation Company, LLC (EGC) requested NRC approval of the several relief requests associated with the fourth Inservice Inspection (ISI) interval for Clinton Power Station (CPS),

Unit 1. One of the relief requests (i.e., Relief Request I4R-02) was related to alternative requirements for reactor pressure vessel nozzle-to-vessel welds and nozzle inner radius section. Relief Request I4R-02 has been revised to correct the value of the nozzle inner radius (i.e., ri) for the reactor recirculation outlet nozzle (N1). Revision 1 of Relief Request I4R-02 is provided in the Attachment.

There are no regulatory commitments contained within this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

Respectfully, Patrick R. Simpson Sr. Manager Licensing

Attachment:

10 CFR 50.55a Relief Request I4R-02, Revision 1 cc: NRC Regional Administrator, Region III NRC Senior Resident Inspector - Clinton Power Station

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 1 of 8)

1. ASME Code Component(s) Affected Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.90 and B3.100

Description:

Alternative Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section (IWB-2500, Table IWB-2500 Inspection Program)

Component Number: Nozzles N1, N2, N3, N5, N6, N7, N8, N9, and N16 (See the Enclosure for a complete list of nozzle identification numbers)

2. Applicable Code Edition The Fourth Ten-Year Interval of the Clinton Power Station, Unit 1 (CPS) Inservice Inspection (ISI) Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code,Section XI, 2013 Edition. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2013 Edition is implemented as required (and modified) by 10 CFR 50.55a(b)(2)(xiv) and (xviii).

3. Applicable Code Requirement

The applicable requirement is contained in Subsection IWB, Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzle in Vessels - Inspection Program," Class 1 Reactor Vessel nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Numbers B3.90, "Nozzle-to-Vessel Welds," and B3.100, "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 nozzles identified in the Enclosure are full penetration welds.

4. Reason for Request

NRC Regulatory Guide (RG) 1.147, Revision 19 conditionally accepts the use of ASME Code Case N-702 (N-702) (Reference 3). This code case provides an alternative to performing examination of 100% of the nozzle-to-vessel welds and inner radii for Examination Category B-D nozzles with the exception of the Feedwater and Control Rod Drive Return Line (CRDRL) nozzles. The alternative is to perform examination of a minimum of 25% of the nozzle inner radii and nozzle-to-shell welds, including at least one nozzle from each system and nominal pipe size, excluding the Feedwater and CRDRL Nozzles.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 2 of 8)

N-702 has been approved for use in RG 1.147, Revision 19 with conditions as noted below:

The applicability of Code Case N-702 for the first 40 years of operation must be demonstrated by satisfying 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).

The use of Code Case N-702 in the period of extended operation is prohibited. If VT-1 is used, it shall utilize ASME Code Case N-648-2, "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI Division 1," with associated required conditions specified in Regulatory Guide 1.147.

Note: This code case was previously approved with conditions, the conditions have been revised for Revision 19 of Reg. Guide 1.147.

The analyses in BWRVIP-108NP and BWRVIP-241 were based on predicted fatigue crack growth over the initial licensed operating period and assumed additional fatigue cycles in evaluating fatigue crack growth.

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, and the reduction in examination scope could provide a dose savings of as much as 25 Rem for the entire Fourth ISI Interval.

5. 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 the Enclosure for a list of RPV Examination Category B-D nozzles applicable to this relief request). As an alternative, for all welds and inner radii identified in Table 5-1, CPS proposes to examine a minimum of 25 percent of the nozzle-to-vessel welds and inner radii sections, including at least one nozzle from each system and nominal pipe size, in accordance with N-702. For the nozzle assemblies identified in the Enclosure, this would mean 25 percent from each of the groups identified in Table 5-1 during the 120-month interval.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 3 of 8)

Table 5-1 Clinton Power Station, Unit 1 RPV Examination Category B-D Nozzle Summary Minimum Total Comments Group Number to be Number Results1 Examined 20" Recirculation Outlet 2 1 One (1) nozzle was examined in Nozzles the Third ISI Interval.

(N1) No rejectable indications.

10" Recirculation Inlet 10 3 Three (3) nozzles were Nozzles examined in the Third ISI (N2) Interval.

No rejectable indications.

24" Main Steam 4 1 One (1) nozzle is scheduled to Nozzles be examined in the Third ISI (N3) Interval.

No rejectable indications.

12" Core Spray 2 1 One (1) nozzle was examined in Nozzles the Third ISI Interval.

(N5) No rejectable indications.

10" Low Pressure 3 1 One (1) nozzle was examined in Coolant Injection the Third ISI Interval.

Nozzles No rejectable indications.

(N6) 6" Head Spray Nozzles 2 1 Two (2) nozzles were examined (N7 and N8) in the Third ISI Interval.

No rejectable indications.

4" Jet Pump 2 1 One (1) nozzle was examined in Instrumentation the Third ISI Interval.

nozzles No rejectable indications.

(N9)

Vibration 1 1 One (1) nozzle was examined in Instrumentation Nozzle the Third ISI Interval.

(N16) No rejectable indications.

Note:

1. The nozzle-to-vessel weld and inner radius examinations are performed together.

The examinations in Table 5-1 will be scheduled in accordance with ASME Section XI, IWB-2411, "Inspection Program."

N-702 stipulates that a VT-1 visual examination may be used in lieu of the volumetric examination for the nozzle inner radii (i.e., Item Number B3.100, "Nozzle Inside Radius Section"). This VT-1 visual examination is outlined in Code Case N-648-2 (N-648-2)

("Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel NozzlesSection XI, Division 1"). CPS will perform either volumetric examination or VT-1

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 4 of 8) visual examination of the inner radius as required by N-702. (Note, however, that CPS is not currently using N-648-2 and is planning to continue to perform volumetric examinations of all required nozzle inner radii.)

The Electric Power Research Institute (EPRI) Technical Report 1003557 (Reference 1),

"BWRVIP-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," (Reference 1) found 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 10-6 for 40 years) with or without any inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified.

EPRI Report BWRVIP-241 received a final NRC Safety Evaluation Report on April 19, 2013 (ML13071A240). In the NRC Safety Evaluation Report, Section 5.0, "Conditions and Limitations," indicates that each licensee who plans to request relief from ASME 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 N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-241 report to their units in the relief request by demonstrating that the following general and nozzle-specific criteria are satisfied:

In the case of CPS, the single set of values (e.g., nozzle radii, nozzle thicknesses, etc.)

used in the following equations are correct and applicable to CPS. These values are minimum design values.

Responses to NRC Plant Specific Applicability

1. The maximum RPV heatup/cooldown rate is limited to less than 115°F/hour.

This criterion is met by adherence to CPS Technical Specification (TS) 3.4.11, "Reactor Coolant System Pressure/Temperature Limits," Surveillance Requirement 3.4.11.1 which requires verification that the Reactor Coolant System (RCS) heatup and cooldown rates are limited to less than or equal to 100°F in any one hour period and, less than or equal to 20°F in any one hour period during RPV pressure testing.

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 5 of 8)

2. For the Reactor Recirculation Inlet Nozzles (N2), the following criteria must be met:
a. (pr/t)/CRPV 1.15; p=RPV Normal Operating Pressure 1025 psig r=RPV inner radius 110.19 inches t=RPV wall thickness 6.1 inches CRPV= 19332 Result: (pr/t)/CRPV = 0.96 1.15 The calculation for the CPS N2 Nozzle results in a maximum value of 0.96, which is less than 1.15 and satisfies this criteria.
b. [p(ro2 +ri2)/(ro2-ri2)]/CNOZZLE 1.15; p=RPV Normal Operating Pressure 1025 psig ro=nozzle outer radius 11.69 inches ri=nozzle inner radius 5.81 inches CNOZZLE 1637 Result: [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE = 1.04 1.15 The calculation for the CPS N2 Nozzle results in a maximum value of 1.04, which is less than 1.15 and satisfies this criteria.
3. For the Reactor Recirculation Outlet Nozzles (N1), the following criteria must be met:
a. (pr/t)/CRPV 1.15; p=RPV Normal Operating Pressure 1025 psig r=RPV inner radius 110.19 inches t=RPV wall thickness 6.1 inches CRPV= 16171 Result: (pr/t)/CRPV = 1.14 1.15 The calculation for the CPS N1 Nozzle results in a value of 1.14, which is less than 1.15 and satisfies the criteria.
b. [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE 1.15; p=RPV Normal Operating Pressure 1025 psig ro=nozzle outer radius 16.3125 inches ri=nozzle inner radius 9.0 inches CNOZZLE 1977 Result: [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE = 0.97 1.15

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 6 of 8)

The calculation for the CPS N1 Nozzle results in 0.97, which is less than 1.15 and satisfies this criteria.

Based upon the above information, all CPS RPV nozzle-to-vessel shell full penetration welds and nozzle inner radii sections meet the general and nozzle-specific criteria in BWRVIP-241.

The analyses for the nozzles in BWRVIP-108NP and BWRVIP-241 are based on the assumption that fluence at the nozzles is negligible because the analysis is for the initial 40 years of plant operation and do not address an extended operating period. Pressure-Temperature Limits reports applicable to CPS, concluded that peak fluence over the period of extended operation (54 effective full power years) is expected to be less than the fluence criteria of 1.0E17 n/cm2, as contained in 10 CFR 50, Appendix H for all nozzles and welds for which this relief request is applied. Therefore, the fluence criteria is satisfied and use of BWRVIP-108 and BWRVIP-241 remain applicable to the CPS nozzles contained in this relief request.

Therefore, use of N-702 provides an acceptable level of quality and safety pursuant to 10 CFR 50.55a(z)(1) for all applicable full penetration RPV nozzle-to-vessel shell welds and nozzle inner radii sections for the Fourth ISI Interval.

6. Duration of Proposed Alternative

Relief is requested for the Fourth ISI Interval for CPS, or until the NRC approves N-702, or a later revision, in Regulatory Guide 1.147 or other document during the interval.

7. Precedents

  • Clinton Power Station, Unit 1, Third ISI Interval Relief Request I3R-02 was authorized by NRC SE dated December 22, 2010 (ADAMS Accession No. ML103360335) (Reference 7). This relief request for the Clinton Power Station, Unit 1, Fourth ISI Interval, utilizes a similar approach to the previously approved relief request.
  • Peach Bottom Atomic Power Station, Units 2 and 3 Relief Request I5R-04 was authorized by NRC SE dated December 21, 2018 (ADAMS Accession No. ML18331A216) (Reference 8).
  • James A. FitzPatrick Nuclear Power Plant Relief Request I5R-05 was authorized by NRC SE dated September 10, 2018 (ADAMS Accession No. ML18239A010)

(Reference 9).

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 7 of 8)

8. References
1. EPRI Technical Report 1003557, "BWRVIP-108: BWR 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,"

dated October 2002

2. ASME Boiler and Pressure Vessel Code, Code Case N-648-2, "Alternative Requirements for Inner Radius Examinations of Class 1 Reactor Vessel Nozzles,Section XI, Division 1," September 4, 2014
3. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," dated February 20, 2004
4. NRC Regulatory Guide RG 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 19
5. Letter from Matthew A. Mitchell (NRC) to Rick Libra (BWRVIP Chairman), "Safety Evaluation of Proprietary EPRI Report, 'BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108),'" dated December 19, 2007 (ADAMS Accession No. ML073600374)
6. Letter from Kevin Hsueh (NRC) to Tim Hanley (BWRVIP Chairman), "Revised Final Safety Evaluation for the License Renewal Appendix A for "BWRVIP-241-A:

BWR Vessel and Internals Project, Probabilistic Fracture Mechanics Evaluation for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," and "BWRVIP-108NP-A: BWR 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" (TAC NO. MF4638), April 26, 2017(ADAMS Accession No. ML17114A096)

7. Letter from R. D. Carlson (NRC) to M. J. Pacilio (EGC), "Clinton Power Station, Unit No. 1 -Relief Requests I3R-01, I3R-02, I3R-03, I3R-04, and I3R-05 Associated with the Third Inservice Inspection Interval (TAC Nos. ME2987, ME2988, ME2989, ME2990, and ME2991)," dated December 22, 2010 (ADAMS Accession No. ML103360335)
8. Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "Peach Bottom Atomic Power Station, Units 2 and 3 - Issuance of Alternative Requests Related to the Fifth Inservice Inspection Interval (EPID L-2018-LLR-0055, EPID L-2018-LLR-0057, EPID L-2018-LLR-0058, and EPID L-2018-LLR-0059)," dated December 21, 2018 (ADAMS Accession No. ML18331A216)

10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Nozzle-to-Vessel Welds and Nozzle Inner Radius Section in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Page 8 of 8)

9. Letter from J. G. Danna (NRC) to B. C. Hanson (EGC), "James A. FitzPatrick Nuclear Power Plant - Issuance of Relief from the Requirements of the ASME Code N-702 for Plant Nozzle-to-Vessel Welds and Inner Radii Examinations (EPID L-2017-LLR-0093)," dated September 10, 2018 (ADAMS Accession No. ML18239A010)
9. Enclosure Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1

Enclosure 10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Weld Inspection (Nozzle to Shell and Inner Radius) in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Enclosure Page 1 of 3)

Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N1A 20" Recirculation Outlet Nozzle N1A to Vessel B-D B3.90 Weld N1A-IRS 20" Recirculation Outlet Nozzle N1A Inner B-D B3.100 Radius N1B 20" Recirculation Outlet Nozzle N1B to Vessel B-D B3.90 Weld N1B-IRS 20" Recirculation Outlet Nozzle N1B Inner B-D B3.100 Radius N2A 10" Recirculation Inlet Nozzle N2A to Vessel B-D B3.90 Weld N2A-IRS 10" Recirculation Inlet Nozzle N2A Inner Radius B-D B3.100 N2B 10" Recirculation Inlet Nozzle N2B to Vessel B-D B3.90 Weld N2B-IRS 10" Recirculation Inlet Nozzle N2B Inner Radius B-D B3.100 N2C 10" Recirculation Inlet Nozzle N2C to Vessel B-D B3.90 Weld N2C-IRS 10" Recirculation Inlet Nozzle N2C Inner B-D B3.100 Radius N2D 10" Recirculation Inlet Nozzle N2D to Vessel B-D B3.90 Weld N2D-IRS 10" Recirculation Inlet Nozzle N2D Inner B-D B3.100 Radius N2E 10" Recirculation Inlet Nozzle N2E to Vessel B-D B3.90 Weld N2E-IRS 10" Recirculation Inlet Nozzle N2E Inner Radius B-D B3.100 N2F 10" Recirculation Inlet Nozzle N2F to Vessel B-D B3.90 Weld N2F-IRS 10" Recirculation Inlet Nozzle N2F Inner Radius B-D B3.100 N2G 10" Recirculation Inlet Nozzle N2G to Vessel B-D B3.90 Weld N2G-IRS 10" Recirculation Inlet Nozzle N2G Inner B-D B3.100 Radius N2H 10" Recirculation Inlet Nozzle N2H to Vessel B-D B3.90 Weld N2H-IRS 10" Recirculation Inlet Nozzle N2H Inner B-D B3.100 Radius N2J 10" Recirculation Inlet Nozzle N2J to Vessel B-D B3.90 Weld N2J-IRS 10" Recirculation Inlet Nozzle N2J Inner Radius B-D B3.100

Enclosure 10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Weld Inspection (Nozzle to Shell and Inner Radius) in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Enclosure Page 2 of 3)

Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N2K 10" Recirculation Inlet Nozzle N2K to Vessel B-D B3.90 Weld N2K-IRS 10" Recirculation Inlet Nozzle N2K Inner Radius B-D B3.100 N3A 24" Main Steam Nozzle N3A to Vessel Weld B-D B3.90 N3A-IRS 24" Main Steam Nozzle N3A Inner Radius B-D B3.100 N3B 24" Main Steam Nozzle N3B to Vessel Weld B-D B3.90 N3B-IRS 24" Main Steam Nozzle N3B Inner Radius B-D B3.100 N3C 24" Main Steam Nozzle N3C to Vessel Weld B-D B3.90 N3C-IRS 24" Main Steam Nozzle N3C Inner Radius B-D B3.100 N3D 24" Main Steam Nozzle N3D to Vessel Weld B-D B3.90 N3D-IRS 24" Main Steam Nozzle N3D Inner Radius B-D B3.100 N5A 12" Core Spray Nozzle N5A to Vessel Weld B-D B3.90 N5A-IRS 12" Core Spray Nozzle N5A Inner Radius B-D B3.100 N5B 12" Core Spray Nozzle N5B to Vessel Weld B-D B3.90 N5B-IRS 12" Core Spray Nozzle N5B Inner Radius B-D B3.100 N6A 10" Low Pressure Core Injection Nozzle N6A to B-D B3.90 Vessel Weld N6A-IRS 10" Low Pressure Core Injection Nozzle N6A B-D B3.100 Inner Radius N6B 10" Low Pressure Core Injection Nozzle N6B to B-D B3.90 Vessel Weld N6B-IRS 10" Low Pressure Core Injection Nozzle N6B B-D B3.100 Inner Radius N6C 10" Low Pressure Core Injection Nozzle N6C to B-D B3.90 Vessel Weld N6C-IRS 10" Low Pressure Core Injection Nozzle N6C B-D B3.100 Inner Radius N7 6" Top Head Spray Nozzle N7 to Vessel Weld B-D B3.90 N7-IRS 6" Top Head Spray Nozzle N7 Inner Radius B-D B3.100 N8 6" Top Head Spare Nozzle N8 to Vessel Weld B-D B3.90 N8-IRS 6" Top Head Spare Nozzle N8 Inner Radius B-D B3.100 N9A 4" Jet Pump Instrumentation Nozzle N9A to B-D B3.90 Vessel Weld N9A-IRS 4" Jet Pump Instrumentation Nozzle N9A Inner B-D B3.100 Radius N9B 4" Jet Pump Instrumentation Nozzle N9B to B-D B3.90 Vessel Weld N9B-IRS 4" Jet Pump Instrumentation Nozzle N9B Inner B-D B3.100 Radius

Enclosure 10 CFR 50.55a Relief Request I4R-02 Alternative Examination Requirements for Reactor Pressure Vessel Weld Inspection (Nozzle to Shell and Inner Radius) in Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality or Safety--

Revision 1 (Enclosure Page 3 of 3)

Table of ASME Section XI Components Affected, Clinton Power Station, Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N16 Vibration Instrumentation Nozzle to Vessel B-D B3.90 Weld N16-IRS Vibration Instrumentation Nozzle Inner Radius B-D B3.100

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