NL-18-1384, Fourth 10-Year Interval Inservice Inspection Program ISI Program Update: Notification of Impractical ASME Code Requirements

Fourth 10-Year Interval Inservice Inspection Program ISI Program Update: Notification of Impractical ASME Code Requirements

ML18334A032
Person / Time
Site:	Farley
Issue date:	11/30/2018
From:	Gayheart C Southern Nuclear Operating Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-18-1384 FNP-ISI-RR-02, Ver 1.0, FNP-ISI-RR-03, Ver 1.0
Download: ML18334A032 (44)

Text

~ Southern Nuclear Cheryl A. Gayheart Regulatory Affairs Director 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@ southernco.com NOV 3 0 2018 Docket Nos.: 50-348 NL-18-1384 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Fourth 10-Year lntervallnservice Inspection Program lSI Program Update: Notification of Impractical ASME Code Requirements Ladies and Gentlemen:

In accordance with 10 CFR 50.55a(g)(5)(iii), Southern Nuclear Operating Company (SNC) hereby notifies the U.S. Nuclear Regulatory Commission (NRC) that SNC has determined that conformance with certain ASME Section XI Code (Code) requirements is impractical for the Farley Nuclear Plant, Units 1 and 2 (FNP). SNC submits the enclosed information to support the determinations of impracticality which are based on demonstrated limitations experienced when attempting to comply with the Code requirements during the fourth 10-year lSI program interval. Requests for relief are enclosed.

This letter contains no new NRC Commitments. If you have any questions, please contact Jamie Coleman at 205.992.6611.

Respectfully submitted, Cheryl A. art Regulatory Affairs Director CAG/ndj/sm

Enclosures:

1. FNP-ISI-RR-02, Version 1.0

2. FNP-ISI-RR-03, Version 1.0 cc: Regional Administrator, Region II NRR Project Manager- Farley Nuclear Plant Senior Resident Inspector- Farley Nuclear Plant RTYPE: CFA04.054

Joseph M. Farley Nuclear Plant Fourth 10-Year lntervallnservice Inspection Program lSI Program Update: Notification of Impractical ASME Code Requirements Enclosure 1 FNP-ISI-RR-02, Version 1.0

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

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

Reference:

IWB-2500, Table IWB-2500-1, ASME Code Case N-460, Examination Category: 8-D Item Number: 83.110

Description:

Limited Examination Coverage Component Number: See Tables RR-02.1 and RR-02.2 for a list of Component IDs

2. Applicable Code Edition and Addenda

The Fourth 10-Year Interval of the Farley Nuclear Plant (FNP), Units 1 and 2 lnservice Inspection (lSI) Program was based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

FNP, Units 1 and 2 examinations were performed in accordance with the requirements of ASME,Section XI, Article 4 of Section V. In the case of limited examinations, efforts were made to obtain additional examination coverage.

3. Applicable Code Requirements The extent of examination requirement for Examination Category 8-D, Item Number 83.110, per Table IWB-2500-1, requires a volumetric examination of essentially 100% of the weld length.

4. Impracticality of Compliance Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requirements is impractical since conformance would require extensive structural modifications to the component or surrounding structure.

Due to the original design of these components, it is not feasible to effectively perform the examinations to the extent required for welds and welded attachments (greater than 90% of the volume or area) due to physical obstructions, plant location, and/or component geometry.

FNP is unable to satisfy the ASME Section XI Code requirements to perform a surface or volumetric examination of these components due to the physical component configuration, interference from permanent plant equipment, single-sided access, etc. FNP would incur significant engineering, material, and E1-1

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 installation costs to perform such modifications without a compensating increase in the level of quality and safety. Therefore, relief is requested on the basis that the ASME Section XI Code requirements to examine these components are impractical.

Tables RR-02.1 and RR-02.2 provide a summary of the examination limitations for each component for which relief is requested. The tables also indicate the outage the component was examined, the coverage percentage obtained for each component, and other pertinent design information. These tables are the cumulative lists of the limited ASME Section XI examinations performed during the Fourth lSI Interval. Figures 2-1 through 2-9 provide typical configuration and coverage plots that detail the examination limitations. The shaded areas in the figures show where at least one scan angle is achieved All the welds in the referenced tables, receive an inner radius examination with 100% coverage. No recordable indications have documented on any of these components with the examinations performed during the 4th interval. In reviewing the SNC fleet operating experience, (Vogtle Units 1 and 2 and Farley Units 1 and

2) no leakage or indications that require flaw evaluations or repairs have been found from the Category B-D, Item No. B3.11 0.

Based on the above explanation, SNC requests relief to perform examinations without achieving ASME Section XI Code compliance coverage when the required coverage is impractical.

5. Burden Caused by Compliance Compliance with the applicable ASME Section XI Code volumetric or surface examination requirements can only be accomplished by redesigning and refabricating the subject and/or surrounding components. Based on this, the ASME Section XI Code requirements are deemed impractical in accordance with 10CFR50.55a(g)(5)(iii).

6. Proposed Alternative and Basis for Use FNP has performed the ASME Section XI Code required examinations to the maximum extent practical (Code Coverage), which are documented in Tables RR-02.1 and RR-02.2. Due to the physical interferences causing these limitations, there are no alternative examination techniques currently available to increase coverage.

7. Duration of Proposed Alternative The proposed alternative is applicable for the Fourth lnservice Inspection Interval, extending from December 1, 2007 through November 30, 2017.

E1-2

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02

8. Precedents

• Relief Request 13R-23 was authorized for Limerick Generating Station, Units 1 and 2 by NRC SEdated August 7, 2018 (ADAMS Accession No. ML18192C172).

• Relief Requests LMT-R01, LMT-C02, and LMT-C03 were authorized for Surry Power Station Unit 2 by NRC SEdated February 17, 2017 (ADAMS Accession No. ML16365A118).

9. References NRC Safety Evaluation Report for Third Interval Relief Request RR-6 was approved for the 3rd Interval by NRC TAC numbers M98858, M98859, dated January 12, 1999.

E1-3

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Table RR-02.1 Farley Nuclear Plant, Unit 1 List of Components with Limited Examination Coverage Exam Exam Weld Exam Angle/

Requirements Category/ Outage Diameter/ Actual 3'd Interval Relief Component ID Description Frequency (MHz) I Remarks (Figure No.) Item Examined Thickness Coverage Request (System)1 Mode and Method Number PZR Upper The examination was limited due to Head to IW8*2500*7(b) 8-D o*1 2.251 Long component configuration. An Inner ALA1-2100-9 Safety Nozzle Volumetric 1R22 6"/3.88" 45" /2 .25/ Shear 75% Radius UT and supplemental MT was RR-6 Weld (UT) 83.110 so*12.251 Shear performed with 100% coverage.

(831) (Figures 2-1 and 2-3)

PZR Upper The examination was limited due to ALA1-2100-Head to IW8-2500-7(b) 8-D o* I 2.25/ Long component configuration. An Inner Safety Nozzle Volumetric 1R27 6"13.88" 45" I 2.25 I Shear 78.6% Radius UT and supplemental MT was RR-6 10 83.110 Weld (UT) 60" 12.251 Shear performed with 100% coverage.

(831) (Figures 2-1 and Fig. 2-4)

PZR Upper The examination was limited due to ALA1-2100-Head to IW8-2500-7(b) 8-D o* I 2.25 I Long component configuration. An Inner Safety Nozzle Volumetric 1A24 6"13.88" 45" 12.251 Shear 78.6% Radius UT and supplemental MT was RR-6 11 Weld (UT) 83.110 so* 12.251 Shear performed with 100% coverage.

(831) (Figures 2-1 and Fig. 2-4)

PZR Upper The examination was limited due to ALA1-2100-Head to IW8-2500-7(b) 8-D o* I 2.251 Long component configuration. An Inner Spray Nozzle Volumetric 1R24 4"13.88" 45" 12.25/ Shear 78.6% Radius UT and supplemental MT was RR-6 12 83.110 Weld (UT) so* /2.251 Shear performed with 100% coverage.

(831) (Figures 2-1 and Fig. 2-4)

PZR Upper The examination was limited due to ALA1-2100-Head to IW8-2500-7(b) 8-D o* I 2.251 Long component configuration. An Inner PORV Nozzle Volumetric 1R24 6" 13.88" 45" 12.251 Shear 78.6% Radius UT and supplemental MT was RR-6 13 Weld (UT) 83.110 so* 12.251 Shear performed with 100% coverage.

I (831) (Figures 2-1 and Fig. 2-4)

PZR Lower The examination was limited due to ALA1-2100-Head to IW8-2500-7(b) 8-D o*1 2.251 Long component configuration. An Inner Surge Nozzle Volumetric 1R24 14"/3.88" 45" 12.251 Shear 75% Radius UT and supplemental MT was RR-6 14 83.110 Weld (UT) 60" I 2.251 Shear performed with 100% coverage.

(831) (Figures 2-2 and Fig. 2-5)

1. The following systems and their abbreviations are listed here: Pressurizer (831)

E1-4

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Table RR-02.2 Farley Nuclear Plant, Unit 2 List of Components with Limited Examination Coverage Exam Exam Weld Exam Angle/

Requirements Category/ Outage Diameter/ Actual 3'd Interval Relief Component 10 Description Frequency (MHz) I Remarks (Figure No.) Item Examined Thickness Coverage Request (System)' Mode and Method Number PZR Upper The examination was limited due to Head to IW8*2500-7(b) a* /2.251 Long component configuration. An Inner 8-D APR1-21aa-9 Safety Nozzle Volumetric 2R23 6"13.88" 45" 12.251 Shear 61.1%121 Radius UT and supplemental MT RR-6 83.11a Weld (UT) 60" 12.251 Shear was performed with 100% coverage.

(831) (Figures 2-1 and Fig. 2-6)

PZR Upper The examination was limited due to Head to IW8-2500-7(b) a* 12.251 Long component configuration. An Inner APR1-21aa- 8-D Safety Nozzle Volumetric 2R23 6"13.88" 45" 12.25 I Shear 61.1%121 Radius UT and supplemental MT RR-6 1a 83.11a Weld (UT) 6a* 12.251 Shear was performed with 1aa% coverage.

(831) (Figures 2-1 and Fig. 2-6)

PZR Upper The examination was limited due to Head to IW8-2500-7(b) a* 12.25 I Long component configuration . An Inner APR1-21aa- 8-D so.a%121 Safety Nozzle Volumetric 2R24 6" 13.88" 45" 12.25 I Shear Radius UT was performed with RR-6 11 83.11a Weld (UT) 60" 12.251 Shear 100% coverage.

(831) (Figures 2-1 and Fig. 2*7)

PZR Upper The examination was limited due to Head to IW8-2500-7(b) a* 12.251 Long component configuration. An Inner APR1-21aa- 8-D Spray Nozzle Volumetric 2R24 4" 13.88" 45" 12.25 I Shear 5a.a%121 Radius UT was performed with RR-6 12 83.11a Weld (UT) 60" 12.251 Shear 100% coverage.

(831) (Figures 2-1 and Fig. 2-7)

PZR Upper The examination was limited due to Head to IW8-2500-7(b) a* 12.251 Long component configuration. An Inner APR1-2100- 8-D PORV Nozzle Volumetric 2R19 6" 13.88" 45" 12.251 Shear 75.a% Radius UT and supplemental MT RR-6 13 83.11a Weld (UT) 60" 12.251 Shear was performed with 100% coverage.

(831) (Figures 2-1 and Fig. 2-8)

PZR Lower The examination was limited due to Head to IW8-2500-7(b) a* 12.251 Long component configuration An Inner APR1-2100- 8-D Surge Nozzle Volumetric 2R19 14"13.88" 45" 12.251 Shear 6t.a% Radius UT and supplemental MT RR-6 14 83.11a Weld (UT) 60" 12.251 Shear was performed with 100% coverage.

(831) (Figures 2-2 and Fig. 2-9)

1. The following systems and their abbreviations are listed here: Pressurizer (831 ).

2. Lower percentages due to where technician determined inspection angle changed due to configuration of component.

E1-5

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Figure 2-1: Typical 4" and 6" Nozzle Configuration NOZZLE FORGING (CARBO:-! STEEL~ __....,-+----,rr-~

WELD DEPOSITED VESSEL HEAD CLADDING PRESSURIZER NOZZLE*TO.VESSEL WELD Figure 2-2: Typical 14" Nozzle Configuration E1-6

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISJ-RR-02 Figure 2-3: Nozzle Limitations

[ALA1-21 00-9]

PZR UPPER HEAD Upstream ao*os u-4~* OS Taal IINII.iiiiith Total CW/Val Total CCW/Vol Total UPNol DNNol CWilenalll CCWilenalll 0' 100% 100% 100% - . - - - - - -

45' 26.2% 100% 26.2% 95.2% 100% 9!1.2% 100% 100% 100% 100% 100% 100%

60' 10.5% 100% 10.5% 705% 100% 70.5% 100% 100% 100% 100% 100% 100%

TOTAL COVERAGE - 75%

Figure 2-4: Nozzle Limitations

[ALA1-2100-10, ALA1-2100-11, ALA-2100-12, ALA-2100-13]

PZR UPPER HEAD Upstream UPNOI UP/Unalh Total DN/Val ~ Total CW/Vol CWilenatll Total CCW/Vol CCWIUncdll Total o* 100% 10o% 100%

45' 97% 100% 97% 26.2% 100% 26.2% 100% 100% 100% 100% 100% 100%

60' 95% 100% 95% 10.5% 100% 10.5% 100% 100% 100% 100% 100% 100%

TOTAL COVERAGE- 78.6%

E1-7

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Figure 2-5: Nozzle Limitations

[ALA1-2100-14]

Heater sleeve penetrations PZR LOWER HEAD eo* us ONNcil- DNII.enath UPNol UPII.onath Tolal Tolal CWNol CWn..nath TOW CCWNol ccw~ Total 0' 100% 100% 100% - - - - - - -

45' 262% 100% 26.2% 95.2% 100% 95.2% 100% 100% 100% 100% 100% 100%

60' 10.5% 100% 10.5% 70.5% 100% 70.5% 100% 100% 100% 100% 100% 100%

TOTAL COVERAGE -75%

Figure 2-6: Nozzle Limitations

[APR1-2100-9, APR1-2100-10]

PZR UPPER HEAD Upstream UPNol UPII.onath Tolal DNNol ON/length TOtal CWNcil CWII.onath TOW CCWNol CCWII.onath Total 0'

45' 69.8%

901%

100%

100%

69.8%

90.1% 31 .4%

100%

31.4%

62.9%

100%

62.9"1.

629%

100%

62_9"1.

60' 906% 100% 90.6% 161% 100% 16.1% 629% 100% 62.9"1. 62.9% 100% 62_9"1.

TOTAL COVERAGE- 61.1%

E1-8

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Figure 2-7: Nozzle Limitations

[APR1-2100-11, APR1-2100-12]

CL NOZZLE PZR UPPER HEAD- 3.88" (Sketch)

Downstream 15" Upstream PZR UPPER HEAD Required Volume 5.6" x UA:lf-per!btg)"- 21.72 hln 0'-10.86mln 45' -17.52 & 4.20 mIn 60' -19.62 & 2.10 mIn UPI\Ial Total lliiiVol ~ Total CWNol CW~~Anc;~~> Tolol CCWNol otal 0' 50.0% 100% 50.0% . . . . . . . .

45' 80.7% 100% 80.7% 19.3% 100% 19.3% 50.0% 100% 50.0% 50.0% 100% 50.0%

60' 90.3% 100% 90.3% 9.7% 100% 9.7% 50.0% 100% 50.0% 50.0% 100% 50.0%

TOTAL COVERAGE- 50.0%

Figure 2-8: Nozzle Limitations

[APR1-2100-13]

PZR UPPER HEAD Upstream UPNol Pn..nath Tolal ONNol Total CWNol Tolol CCWNol ccwn..nau. Total OIUL...ath CWIL-o* 100% 100% 100% - - . . . - . .

45' 262% 100% 26.2% 952% 100% 95.2% 100% 100% 100% 100% 100% 100%

60' 10.5% 100% 10.5% 70.5% 100% 70.5% 100% 100% 100% 100% 100% 100%

TOTAL COVERAGE- 75%

E1-9

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-181-RR-02 Figure 2-9: Nozzle Limitations

[APR1-2100-14]

Heater sleeve penetrations NOZZLE PZR LOWER HEAD Downstream UPNol UP/Lonalh Total DNNol OIIIIAaatll Total C:WNol C:W/Lenalll Total CC:WNol CC:W/Lenalll Tobl 0' 69.8% 100% 69.8% - - - - - - - - -

45° 90.1% 100% 90.1% 31 .4% 100% 31.4% 62.9°,(, 100% 62.9% 62.9% 100% 62.9%

so* 906% 100% 90.6% 16.1% 100% 16.1% 62.9% 100% 62.9% 62.9% 100% 62.9%

TOTAL COVERAGE- 61%

E1-10

Joseph M. Farley Nuclear Plant Fourth 10-Year lntervallnservice Inspection Program lSI Program Update: Notification of Impractical ASME Code Requirements Enclosure 2 FNP-ISI-RR-03, Version 1.0

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Relief Request In Accordance with 10 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

1. ASME Code Component(s) Affected Code Class: 1&2

Reference:

IWB-2500, Table IWB-2500-1, IWC-2500, Table IW8-2500-1, ASME Code Case N-460, ASME Code Case N-716, Table 1 Examination Category: 8-F, 8-J, C-F-1, R-A Item Number: 85.70, 89.11, 89.21, C5.11, R1.11, R1.16, R1.20

Description:

Limited Examination Coverage Component Number: See Tables RR-03.1 and RR-03.2 for a list of Component IDs

2. Applicable Code Edition and Addenda

The Fourth 10-Year Interval of the Farley Nuclear Plant (FNP), Units 1 and 2 lnservice Inspection (lSI) Program was based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2001 Edition through the 2003 Addenda.

FNP, Units 1 and 2 maintains the responsibility to ensure exams were performed in accordance with the requirements of ASME,Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," as amended and mandated by 10CFR50.55a and as modified by the Performance Demonstration Initiative (PDI) Program description. In the case of limited examinations, efforts were made to obtain additional examination coverage. Tables RR-03.1 and RR-03.2 identify if the examinations were performed in accordance with the requirements of ASME,Section XI, Appendix VIII.

3. Applicable Code Requirements The extent of examination requirement for Examination Category 8-F, Item Number 85.70, per Table IWB-2500-1, requires a volumetric examination of 100% of the weld.

The extent of examination requirement for Examination Category 8-J, Item Numbers 89.11 and 89.21, per Table IWB-2500-1, requires a surface and volumetric examination of essentially 100% of the weld length.

The extent of examination requirement for Examination Category C-F-1, Item Number C5.11, per Table IWC-2500-1, requires a surface and volumetric examination of essentially 100% of the weld length.

E2-1

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 The extent of examination requirement for Examination Category A-A, Item No. R1.11, per Code Case N-716 Table 1, requires a volumetric examination of High Safety Significant (HSS) pressure-retaining welds of Class 1 and 2 welds subject to Thermal Fatigue.

The extent of examination requirement for Examination Category A-A, Item No. R1.16, per Code Case N-716 Table 1, requires a volumetric examination of HHS pressure-retaining welds of Class 1 and 2 welds subject to lntergranular or Transgranular Stress Corrosion Cracking (IGSCC or TGSCC).

The extent of examination requirement for Examination Category A-A, Item No. R1.20, per Code Case N-716 Table 1, requires a volumetric examination of HHS pressure-retaining welds of Class 1 and 2 welds not subject to a degradation method.

During the Fourth lSI Interval, no recordable indications were identified during examination of the Examination Category A-A components listed in Tables RR-03.1 and RR-03.2.

FNP, Units 1 and 2 adopted ASME Code Case N-460 ("Alternative Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division 1"), which defines "essentially 100%" as greater than 90% coverage of the examination volume or surface area, as applicable. The 90% minimum coverage was applied to all surface and volumetric examinations required by ASME Section XI.

4. Impracticality of Compliance Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requirements is impractical since conformance would require extensive structural modifications to the component or surrounding structure.

Due to the original design of these components, it is not feasible to effectively perform the examinations to the extent required for welds and welded attachments (greater than 90% of the volume or area) due to physical obstructions, plant location, and/or component geometry.

FNP is unable to satisfy the ASME Section XI Code requirements to perform a surface or volumetric examination of these components due to the physical component configuration, interference from permanent plant equipment, single-sided access, etc.

FNP would incur significant engineering, material, and installation costs to perform such modifications without a compensating increase in the level of quality and safety.

Therefore, relief is requested on the basis that the ASME Section XI Code requirements to examine these components are impractical.

For the AI-lSI weld population, Examination Category A-A welds, submitted in this relief request, a case by case review was performed to determine whether additional or alternative welds could have been examined to supplement the reduced volumetric coverage examination. Below summarize the additional examinations performed:

E2-2

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 For item number R1.11, six (6) additional examinations were performed.

For item number R1.20, 147 additional examinations were performed with this degradation method. 138 of these examinations are in the Break Exclusion Region and required by Technical Specifications to inspect every 10 years under an augmented program.

For the Category B-F Item B5.70 welds, Steam Generator Nozzle-to-Safe-End, the steam generator was replaced in the 3rd Interval for Units 1 and 2. The materials of construction for these welds are as follows: Safe End- SA-336 F316LN, Nozzle- SA-508 Class 3, Weld Filler- SFA-5.11 CL ENiCrFe-7. Per the Westinghouse Design Reports, WNEP-9830 (Unit 1} and WCAP-15601(Unit 2), the peak stress ratio in the inspectable section of the exam volume is similar in magnitude to the area that is unable to be inspected. No recordable indications were found with these examinations. This gives reasonable assurance of structural integrity or leak tightness continues to exist Tables RR-03.1 and RR-03.2 provide a summary of the examination limitations for each component for which relief is requested. The tables also indicate the outage the component was examined, the coverage percentage obtained for each component, and other pertinent design information. These tables are the cumulative lists of the limited ASME Section XI examinations performed during the Fourth lSI Interval. Figures 3-1 through 3-30 provide coverage plots which were extracted from the non-destructive examination (NDE) summary sheets that detail the examination limitations.

Based on the above explanation, SNC requests relief to perform examinations without achieving ASME Section XI Code compliance coverage when the required coverage is impractical.

5. Burden Caused by Compliance Compliance with the applicable ASME Section XI Code volumetric or surface examination requirements can only be accomplished by redesigning and refabricating the subject and/or surrounding components. Based on this, the ASME Section XI Code requirements are deemed impractical in accordance with 10CFR50.55a(g)(5)(iii).

6. Proposed Alternative and Basis for Use FNP has performed the ASME Section XI Code required examinations to the maximum extent practical (Code Coverage), which are documented in Tables RR-03.1 and RR-03.2. Due to the physical interferences causing these limitations, there are no alternative examination techniques currently available to increase coverage.

7. Duration of Proposed Alternative Relief is requested for the Fourth lSI Interval for FNP, Units 1 and 2.

E2-3

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03

8. Precedents

• Relief Request 13R-23 was authorized for Limerick Generating Station, Units 1 and 2 by NRC SEdated August 7, 2018 (ADAMS Accession No. ML18192C172).

• Relief Requests LMT-R01, LMT-C02, and LMT-C03 were authorized for Surry Power Station Unit 2 by NRC SEdated February 17, 2017 (ADAMS Accession No. ML16365A118).

E2-4

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Table RR-03.1 Farley Nuclear Plant, Unit 1 List of Components with Limited Examination Coverage Exam Exam Category/ Appendix Weld Exam Angle/

Requirements Item Number Outage Diameter/ Actual VIII Component ID Description Frequency (UHz) I Remarks (Figure No.) credited Examined Thickness Coverage Qualified (System)' Mode and Method (current R-A Exam Item#)

45' 12.00 I Long Safe*End to IW8-2500-8 34' 11.50 I Long The examination was limited due ALA1-4100- SG Nozzle Surface and 8-F, 8570 1R23 29"14.75" 40' I 1.00 I Long 52.1% Yes to component configuration.

26RDM Weld Volumetric (R-A, R1 .20) 40' 11.50 I Shear (Figure 3-1)

(813) {MT){UT) 34' 11.50 I Shear The examination was limited due IW8-2500-8 to component configuration.

Valve to Pipe 45' 12.251 Shear Surface and 8-J. 89.11 Additional inspections performed ALA 1-4103-4 Weld 1R22 6"1 0.75" 70' 12.251 Shear 50.0% Yes Volumetric (R-A, R1 .11) for MRP-146 during 1R24, 1R25, (813) 60' I 2.00 I Long (PT){UT) 1R26, & 1R27.

(Figures 3-2 & 3-2a)

The examination was limited due IW8-2500-8 to component configuration.

Valve to Pipe Surface and 8-J, 89.11 45' 12.251 Shear Additional inspections performed ALA 1-4104-4 Weld 1R22 6"1 0.75" 50.0% Yes Volumetric (R-A,R1 .11) 60' I 2.00 I Long for MRP-146 during 1R24, 1R25, (813)

(PT){UT) 1R26, & 1R27.

(Figure 3-3)

IW8-2500-8 Pipe to Valve The examination was limited due Surface and 8-J, 89.11 45' 12.251 Shear ALA 1-4104-5 Weld 1R22 6" I 0.75" 50.0% Yes to component configuration.

Volumetric (R-A, Rl .ll/16) 60' I 2.00 I Long (E11) (Figure 3-4)

{PT)(UT)

IW8-2500-8(c)

ALA 1-41 08 Valve to Pipe IW8-2500-9, o* I 4.00 I Long The examination was limited due Weld 10,& 11 R-A, R1.11 1R27 3" I 0.438" 45' 15.00 I Shear 50.0% Yes to component configuration.

R8 (E21) Volumetric 70' 12.251 Shear (Figure 3-5)

(UT)

Note: The following systems and their abbreviations are listed here: Reactor Coolant System (813), LHSI/RHR (E11 ),

HHSI/CVCS (E21)

E2-5

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Table RR-03.1 Farley Nuclear Plant, Unit 1 List of Components with Limited Examination Coverage Exam Exam Category I Appendix Weld Exam Angle/

Requirements Item Number Outage Diameter/ Actual VIII Component 10 Description Frequency (MHz) I Remarks (Figure No.) credited Examined Thickness Coverage Qualified (System)' Mode and Method (current R-A Exam Item#)

IWB-2500-S(c)

ALA1-4108-Tee to Pipe IWB-2500-9, o* /4.00 I Long The examination was limited on the upstream side due to the tee Weld 10, & 11 R-A, R1 .20 1R26 2"1 0.344 45" 15.00 I Shear 50.0% Yes 14BW-RB component configuration.

(B13) Volumetric 70" I 2.251 Shear (Figure 3-6)

(UT)

The examination was limited due IWB-2500-S(c) to a welded support and ID Pad ALA 1-4202 Pipe to Elbow IWB-2500-9, 45" 12.251 Shear obstruction. Additional inspections RB Weld 10,& 11 R-A, R1 .11 1R24 6"1 0.75" 60" 12.251 Shear 75.0% Yes performed for MRP-146 during (B13) Volumetric 60" I 2.00 I Long 1R25, 1R26, & 1R27.

(UT)

(Fi!lures 3-7 & 3-7a)

The examination was limited due IWB-2500-S(c) to component configuration .

Valve to Pipe IWB-2500-9, 45" /2.251 Shear ALA 1-4202 Additional inspections performed Weld 10, & 11 R-A, R1 .11 1R24 6"1 0.75" 60" 12.251 Shear 50.0% Yes RB for MRP-146 during 1R25, 1R26, (B13) Volumetric 60" 12.00 I Long

& 1R27 (UT)

(Fi!lures 3-8 & 3-Ba)

The examination was limited due IWB-2500-S(c) to component configuration.

Valve to Pipe IWB-2500-9, 45" 12.251 Shear Additional inspections performed ALA 1-4204-4 Weld 10, & 11 R-A, R1.11 1R24 6"1 0.75" 60" I 2.251 Shear 50.0% Yes for MRP-146 during 1R25, 1R26, (B13) Volumetric 60" 12.00 I Long

& 1R27 (UT)

(Figures 3-9 & 3-9a)

IWB-2500-S(c)

Pipe to Valve IWB-2500-9, 45" 12.251 Shear The examination was limited due ALA 1-4204 Weld 10, & 11 R-A, R1.16 1R24 6"1 0.75" 60" 12.251 Shear 50.0% Yes to component configuration.

RB (E21) Volumetric 60' 12.00 I Long (Figures 3-10 & 3-10a)

(UT)

IWB-2500-S(c)

Flange to IWB-2500-9, The examination was limited due ALA1-4209- 45' I 5.00 I Shear Pump Weld 10,& 11 R-A, R1 .20 1R24 2"1 0.196" 48.0% Yes to component configuration.

11BW-RB 70' 12.251 Shear (E21) Volumetric (Figure 3-11)

(UT)

Note: The following systems and their abbreviations are listed here: Reactor Coolant System (813), LHSI/RHR (E11 ), HHSI/CVCS (E21)

E2-6

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Table RR-03.2 Farley Nuclear Plant, Unit 2 List of Components with Limited Examination Coverage Exam Exam Category/ Appendix Weld Exam Angle/

Requirements Item Number Outage Diameter/ Actual VIII Component ID Description Frequency (MHz) I Remarks (Figure No.) credited Examined Thickness Coverage Qualified (System)' Mode and Method (current R-A Exam Item II}

45' /2.00 I Long SG Sale-End IW9-2500-B 34 * /1.50 I Long The examination was limited due to APR1*4300- to Nozzle Surtace and 9 -F, 95.70 2R19 29" /4.75" 40' 11 .00 I Long 52.1% Yes component configuration.

23RDM Weld Volumetric (R-A, R1.20) 40' 11 .50 I Shear (Figure 3-1)

(913) (MT)(UT) 34' 1 1.50 I Shear 45' 12.00 I Long SG Safe-End IW9-2500-B 34 ' 11 .50 I Long The examination was limited due to APR1-4300* to Nozzle Surtace and 9-F, 95.70 2R19 31" 14.75" 40' 11 .00 I Long 52.1% Yes component configuration.

24RDM Weld Volumetric (R*A, R1 .20) 40' 11 .50 I Shear (Figure 3-1)

(913) (MT)(UT) 34' 11.50 I Shear The examination was limited due to IW9-2500-B Pipe to Valve 45' 12.251 Shear component configuration .

Surtace and 9 -J, 99.11 APR1-4101*B Weld 2R20 12" 11 .125" 60' 12.0 I Long 50.0% Yes Additional inspections pertormed Volumetric (R-A,R1 .11)

(913) 70' 12.251 Shear for MRP-146 during 2R22.

(PT)(UT)

(Figure 3-12}

IW9-2500*B(c)

Pipe to Pipe IW9-2500*9, 45' 12.251 Shear The examination was limited due to APR1-4102*

Weld 10, & 11 R-A, R1 .20 2R24 12" 11.125" 60' I 2.0 I Long 75.0% Yes a box restraint.

2-R9 (E21) Volumetric 60' I 2.25 I Shear (Figure 3-13)

(UT)

IW9-2500-B Pipe to Valve 45' 12.251 Shear The examination was limited due to APR1-4104* Surtace and 9-J, 99.11 Weld 2R22 6" I 0.719" 60' 12.251 Shear 50.0% Yes component configuration.

30 Volumetric (R-A. R1.16) I (E21} 60' I 2.00 I Long (Figure 3-14)

J (PT)(UT)

Note: The following systems and their abbreviations are listed here: Reactor Coolant System {813}, LHSI/RHR {E11 ),

HHSI/CVCS {E21)

E2-7

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Table RR-03.2 Farley Nuclear Plant, Unit 2 List of Components with Limited Examination Coverage Exam Exam Category/ Appendix Weld Exam Angle/

Requirements Item Number Outage Diameter/ Actual VIII Component 10 Description Frequency (MHz) I Remarks (Figure No.) credited Examined Thickness Coverage Qualified (System )I Mode and Method (current R-A Exam Item#)

IWB*2500*8(c)

The examination was limited due to Valve to Pipe IWB-2500*9, APRl-4106* 45" 15.00 I Shear component configuration and a Weld 10, & 11 R*A, R1.11 2R23 3" I 0.438" 49.4% Yes 8*RB 70" 12.251 Shear welded plate.

(E21) Volumetric (Figures 3-15 and 3-15a)

(UT)

The examination was limited due to Pipe to IWB-2500-8(c) component configuration.

Branch IWB-2500-9, APR1-4106* 45" 15.00 I Shear Additional inspections performed Connection 10, & 11 R-A,R1.11 2R23 3" I 0.438" 50.0% Yes 11-RB 70" 12.251 Shear for MRP-146 during 2R21, 2R24, &

Weld Volumetric 2R25.

(E21) (UT)

(Figure 3-16)

IWB-2500*8(c)

Pipe to Valve IWB-2500-9, 45" I 5.00 I Shear The examination was limited due to APR1-4108*

Weld 10, & 11 R-A, R1.11 2R24 3" I 0.438" 60" I 5.00 I Shear 50.0% Yes component configuration.

11-RB (813) Volumetric 70" 12.251 Shear (Figure 3-17)

(UT)

IWB-2500-8(c)

Valve to Pipe IWB-2500-9, 45" 15.00 I Shear The examination was limited due to APR1-4108-Weld 10, & 11 R-A, R1.11 2R24 3" I 0.438" 60" 15.00 I Shear 50.0% Yes component configuration.

12-RB (E21) Volumetric 70" 12.251 Shear (Figure 3-18)

(UT)

IWB-2500-8(c)

Pipe to Valve IWB-2500-9, 45" I 5.00 I Shear The examination was limited due to APR1-4108*

Weld 10, & 11 R-A, R1.11 2R24 3" I 0.438" 60" I 5.00 I Shear 50.0% Yes component configuration.

13-RB (E21) Volumetric 70" I 2.25 I Shear (Figure 3-19)

(UT)

Note: The following systems and their abbreviations are listed here: Reactor Coolant System (813), LHSI/RHR (E11 ),

HHSI/CVCS (E21)

E2-8

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Table RR-03.2 Farley Nuclear Plant, Unit 2 List of Components with Limited Examination Coverage Exam Exam Category I Appendix Weld Exam Angle/

Requirements Item Number Outage Diameter/ Actual VIII Component ID Description Frequency (MHz) I Remarks (Figure No.) credited Examined Thickness Coverage Qualified (System)' Mode and Method (current R-A Exam ltemltl IW8-2500-8 Pipe to Valve The examination was limited due to Surface and B-J, 89.11 45' 12.251 Shear APR1-4301-8 Weld (813)

Volumetric (R-A.RU1) 2R20 12" I 1.125" eo* 12.00 I Long 50.0% Yes component configuration.

(Figure 3-20)

(PT)(UT)

IW8-2500-8(c)

The examination was limited from Elbow to Pipe IW8-2500-9, APR1-4302- 18" to 28" due to a whip restraint.

Weld 10, & 11 R-A, R1.20 2R25 12" I 1.125" 45" 12.251 Shear 75.0% Yes 11-R8 (Figures 3-12, 3-21a, 3-21b & 3-(813) Volumetric 21c)

(UT)

IW8-2500-8(c)

The examination was limited from Pipe to Elbow IW8-2500-9, APR1-4302- 18" to 28" due to a whip restraint.

Weld 10, & 11 R-A, R1.20 2R25 12"1 1.125" 45" 12.251 Shear 87.0% Yes 12-R8 (Figuras 3-22, 3-22a, 3-22b, & 3-(813) Volumetric 22c)

(UT)

IW8-2500-8 Pipe to Valve The examination was limited due to APR1-4304- Surface and 8-J, 89.11 45" 11.50 I Shear 6" I 0.719" 18 Weld (EH)

Volumetric (R-A, R1.11/16) 2R21 so* 12.00 I Long 50.0% Yes component configuration.

(Figure 3-23)

(PT)(UT)

The examination was limited due to IW8-2500-8(c) component configuration.

Valve to Pipe IW8-2500-9, 45" 12.251 Shear APR1-4304-Weld 10, & 11 R-A,RU1 2R23 6" I 0.719" eo* 12.0 I Long 50.0% Yes Additional inspections performed 19-R8 (EH) Volumetric eo* 12.25 I Shear for MRP-146 during 2R21, 2A22, &

2R24.

(UT)

(Figure 3-24)

Note: The following systems and their abbreviations are listed here: Reactor Coolant System (813), LHSI/RHR (E11 ),

HHSI/CVCS (E21)

E2-9

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Table RR-03.2 Farley Nuclear Plant, Unit 2 List of Components with Limited Examination Coverage Exam Exam Category I Appendix Weld Exam Angle/

Requirements Item Number Outage Diameter/ Actual VIII Component 10 Description Frequency (MHz) I Remarks (Figure No.) credited Examined Thickness Coverage Qualified (System)' Mode and Method (current R-A Exam Item#)

The examination was limited due to component configuration and IWB-2500-B(c)

Branch thickness changes. During outage IWB-2500*9, 30' /2 .25/ Shear APA1-4307- Connection 2A25, component ID APA1-4208*

10, & 11 A-A, A1 .11 2A25 4"1 0.719" 45' /2.251 Shear 22.3% Yes 21BC-AB Weld 23BC-AB was inspected to 100%

Volumetric 60' I 2.251 Shear (E21) coverage with similar degradation (UT) method as added assurance.

fFiaure 3-251 Branch IWB-2500*8 45' 11 .50 I Shear The examination was limited due to Connection to Surface and 6-J, 69.11 APA1-4500-1 2A19 14"11 .4" 60' 11 .50 I Shear 45.5% Yes component configuration.

Pipe Weld Volumetric (A*A. A1 .11) 60' 12.00 I Shear (Figure 3-26)

(613) (PT)(UT)

IWC-2500-7 Flange to 45' 12.251 Shear The examination was limited due to APA2-4511- Surface and C-F-1, C5.11 Pipe Weld 2A20 10"/ 0.719" 60' 12.251 Shear 50.0% Yes component configuration.

10 Volumetric (A-A, A1 .11)

(E11) 60' I 2.00 I Long (Figure 3-27)

(PT)(UT)

IWC-2500-7 Pipe to Valve 45' 12.251 Shear The examination was limited due to APA2-4511- Surface and C-F-1, C5.11 Weld 2A20 10"/ 0.719" 60' I 2.251 Shear 50.0% Yes component configuration .

11 Volumetric (A-A, A1 .11)

(E11) 60' 12.00 I Long (Figure 3-28)

(PT)(UT)

IWC-2500-7 Tee to Elbow The examination was limited due to I APA2-4511- Surface and C-F-1, C5.11 45' 12.251 Shear Weld 2A20 10"11.2" 50.0% Yes component configuration.

12 Volumetric (A*A, A1 .11) 60' 12.00 I AL (E11) (Figure 3-29)

(PT)(UT)

Note: The following systems and their abbreviations are listed here: Reactor Coolant System (813), LHSI/RHR (E11 ), HHSI/CVCS (E21)

E2-10

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-1: Safe-End to Nozzle [ALA1-4300-26RDM, APR1-4300-23RDM, &

APR1-4300-24RDM]

3.63 square inches required for ~~ .16 square inches achieved complete coverage for circumference ~ wilh 45 degree exam angle* 4.4% for !he of the weld. circumference ofthe weld 1.90 square inches achieved 3.63 square inches achieved in the with 34 degree exam angle- 52% for the circumferential¥:811 directions- all elllUTI angles I00% for the circumference of the weld.

circumference of the weld.

Total coverage obtained"" 52.1% lor the drrumlerence of the weld.

Figure 3-2: 6" Pipe to Valve [ALA1-4103-4-RB]

CODE COVERAGE PLOT FLOW WeldCL ALAl-4103-4 Valve pipe TOTAL CODE REQUIRED COVERAGE= 0.4 SQUARE INCHES 100%

TOTAL CODE REQUIRED COVERAGE ON OS SIDE= 0.2 SQUARE INCHES 50%

BEST EFFORT EXAM ON FAR SIDE OF WELD= 0.09 SQUARE INCHES NO COVERAGE ACHIEVED ON FAR SIDE OF WELD= 0.11 SQUARE INCHES TOTAL CODE COVERAGE 50%

(Red cross hatch- no coverage, Blue + Green cross hatch- 60° coverage, Green crosshatch- 45° & 70° coverage)

E2-11

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-3: 6" Pipe to Valve [ALAl-4104-4]

CODE COVERAGE PLOT FLOW WeldCL ALAl-4104-4 Valve pipe TOTAL CODE REQUIRED COVERAGE= 0.4 SQUARE INCHES 100%

TOTAL CODE REQUIRED COVERAGE ON OS SIDE= 0.2 SQUARE INCHES 50%

BEST EFFORT EXAM ON FAR SIDE OF WELD= 0.09 SQUARE INCHES NO COVERAGE ACHIEVED ON FAR SIDE OF WELD= 0.11 SQUARE INCHES TOTAL CODE COVERAGE 50%

(Red cross hatch- no coverage, Blue+ Green cross hatch- 60° coverage, Green crosshatch- 45° coverage)

E2-12

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-181-RR-03 Figure 3-4: 6" Pipe to Valve [ALA1-4104-5]

CODE COVERAGE PLOT FLOW WeldCl

~

45 Dee: 60 Dee: ,

* *..........~** _':J\wr&:t~ a~~ u~~ ~ili _

Required Exam Area= .56 SQ INCHES+ 100%

Coverage Obtained in the Axial Scans = .28 SQ. INCHES = 50%

Coverage Obtained in the Circ Scans = .28 SQ. INCHES = 50%

No coverage obtained in shaded area Combined Coverage = 50%

Figure 3-5: 3" Pipe to Valve [ALA1-4108-8-RB]

fLDvJ vALVf:£.

C. LAmP I~~ .'i3J ,43~

Examined 50% of Required Volume Height- .15", Width- 1.75" Length-0.5" Upstream Ax- 100% Downstream Ax 0%

Upstream Circ- 100% Downstream Circ 0%

E2-13

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-6: 2" Tee to Pipe [ALA1-4108-14BW-RB]

FLOIJ TEE

~* ~*

Figure 3-7: 6" Pipe to Elbow [ALAl-4202-3-RB]

CODE COVERAGE PWT (AREA LIMITATION ONLY) Not to Seale Pipe 1

1 Elbow I l

~------------------~~~~

21" J

Up San (Pipe) -ll" Req'd I 13.40" Esamlaed- 64%

Do ScaD (Elbow) -21" Req'd I 16.45" Eumlaed -78%

CW Scan- 21" Req'd I 16..45" Ena1aed- 78%

CCW Sen -21" Req'd I 16.45" Eumlaed- 78'Ye TOTAL CALCULATED COVERAGE *75%

l&pestfor RelkfReqlllntl E2-14

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-8: 6" Pipe to Valve [ALAl-4202-4-RB]

CODE COVERAGE PLOT FLOW WeldCL ALAl-4202-4 Valve pipe TOTAL CODE REQUIRED EXAM COVERAGE= 0.4 SQUARE INCHES TOTAL BEST EFFORT COVERAGE= 0.12 SQUARE INCHES NO EXAM ON UPSTREAM SIDE = 0.08 SQUARE INCHES TOTAL CODE EXAM COVERAGE= 0.2 SQUARE INCHES= 50%

(Red cross hatch- no coverage, Blue+ Green cross hatch- 60° coverage, Green crosshatch- 45° coverage)

E2-15

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-9: 6" Pipe to Valve [ALAl-4204-4]

CODECOVERAGEPLOT FLOW WeldCL ALAl-4202-4 Valve pipe TOTAL CODE REQUIRED COVERAGE= 0.45 SQ INCHES TOTAL COVERAGE ACHIEVED= 0.225 sa INCHES BEST EFFORT COVERAGE 0.09 sa INCHES TOTAL CODE COVERAGE= 50%

SCANNED ALL ACCESSABLE AREAS (ID PLATE) LOCATED NEAR 90° (Red cross hatch- no coverage, Blue + Green cross hatch- 60° coverage, Green crosshatch- 45° coverage)

E2-16

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-10: 6" Pipe to Valve [ALAl-4204-5-RB]

CODE COVERAGE PLOT FLOW WeldCL ALAl-4204-5 Valve pipe TOTAL CODE REQUIRED COVERAGE= 0.45 SQ INCHES TOTAL COVERAGE ACHIEVED= 0.225 SQ INCHES BEST EFFORT COVERAGE 0.09 SQ INCHES TOTAL CODE COVERAGE= 50%

(Red cross hatch- no coverage, Blue+ Green cross hatch- 60° coverage, Green crosshatch- 45° coverage)

E2-17

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-11: 2" Flange to Pump [ALAl-4209-llBW-RB]

CODE COVERAGE PLOT FLOW WeldCL Flange pipe TOTAL CODE COVERAGE= 0.06 SQ. INCHES TOTAL COVERAGE= 0.029 SQ INCHES=48%

(Red cross hatch- no coverage, Green crosshatch- 70° and 45° coverage)

E2-18

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-12: 12" Pipe to Valve [APR1-4101-8]

51/tluL.'.. .Sibfb tVrM - 5o'/. !'UrM C!.OVffZ.M&. CJB.'f71r.IN~

1.1'1 J.t'i 1.oss t.J'f I I I I VALV£ f:LP£

-FLOW~

Figure 3-13: 12" Pipe to Pipe [APR1-4102-2-RB]

ELOW...._

L*

rElfEIH Of COVEP.A6E AlllAINfO 7.5'"'

E2-19

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-14: 6" Pipe to Valve [APR1-4104-30]

CODE COVERAGE PLOT FLOW WeldCL APRl-4104-30 Valve pipe I+-- 1.5" ... 1 TOTAL CODE REQUIRED COVERAGE= 0.358 SQ INCHES TOTAL CODE COVERAGE ACHIEVED 0.179 SQ INCHES CODE COVERAGE ACHIEVED= 50%

(Red cross hatch- no coverage, Blue + Green cross hatch- 60° coverage, Green crosshatch- 45° coverage)

Figure 3-15: 3" Pipe to Valve [APR1-4106-8-RB]

CODE COVERAGE PLOT FLOW WeldCL APRl-4106-8 Welded plate

.43r pipe

.146" I+-- 1.05" ... 1 TOTAL CODE REQUIRED COVERAGE= 0.15 SO INCHES TOTAL CODE COVERAGE ACHIEVED 0.0741 SO INCHES CODE COVERAGE ACHIEVED= 49.4%

(Red cross hatch- no coverage, Green crosshatch- 45° and 70° coverage)

E2-20

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-15a: 3" Pipe to Valve [APR1-4106-8-RB]

FLOW WELD#8 APR1-4100-8

---,-1\

Figure 3-16: 3" Pipe to Branch Connection [APR1-4106-11-RB]

Ft.. OW f

TOTAL CODE COVERAGE REQUIRED 1.55"

• 0.15" = 0.2325." SQUARE INCHES 1.55* 0.2325 /2 =0.11625" sq. in. =50% CODE COVERAGE E2-21

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-17: 3" Pipe to Valve [APR1-4108-11-RB]

-- _,. ...- ----_.;a..-- --

r--~-~~-

v .._ -----

/

Figure 3-18: 3" Pipe to Valve [APR1-4108-12-RB]

- -- 'v' ---

_...--~-L---

I ,,

( /*II . >

E2-22

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-19: 3" Pipe to Valve [APR1-4108-13-RB]

< )

Figure 3-20: 12" Pipe to Valve [APR1-4301-8]

E2-23

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-21: 12" Elbow to Valve [APR1-4302-11-RB]

Wt-\1f ~E.S.T~AII\IT LtMtTS St.At.l FR~M ll~"-lf:*

FLoW h-

¢_

o.o*L:: PtPE.

\.o~

I. 111. /.lOb / .oCJ 16 l.~'il ElCAM1t-1A-r1fl~ 1/o~un£.: HLI(,HT- 0.31" uJ'bTH ~ '2..1.. LE~bTH* 'H)*

ufSTAEA~ Ax - 15% Dow!'o\STIU.AM Ax- /5°/Q UPSTP.EP.I\1 C..lll..C..- t5*to D~~NST~f-A m C.l!l.c..- 75°/o Figure 3-21a: 12" Elbow to Valve [APR1-4302-11-RB]

Comments: WI-IlP RE.STRAIIJT LIMITS WEl.CU II ,..,1..

FP.o P'\ 18"- 'Z.i

Slcoll:h or Phola:

I I I I 1:::.====1 - WELb II I

I E2-24

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-181-RR-03 Umlted on back side of pipe E2-25

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-22: 12" Pipe to Elbow [APR1-4302-12-RB]

FLM.J ...

1.101 f.t.o1. I.ISK 1.1.51" E)(F\1""\tNC::..~ g1,5./o DF ({f.quta..£~ \IOC....l.JI"'t!.

ElC'A IY\tNRnot.\ VoL..ul't"'E. biMF-rlSu:.N5: ~Erc,~r- o ..,** wiDif.-'*1.1" u:.JGrr..t '"lo" Uf'5T~E.AI""\ A'I<- ls*/,. t~Dwf\lC::Tit£A~ Ax- I b~ "/a uPST"Il..E.AP1 C.lll.C.-15 8 / * ~NSTitfAfl"' CtPL- IOo"/o Figure 3-22a: 12" Pipe to Elbow [APR1-4302-12-RB]

Comll*lll: Wl-ltP RE.STRf\1"--T LtmiT~ WELDS 11..-1t.

F""-o M I&"- 2.i

Sl<etl:ll at Pllolo:

-lJELl:l II l'l

- WELb 13 E2-26

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-181-RR-03 Limited on back side of pipe E2-27

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-23: 6" Pipe to Valve [APR1-4304-18]

CODE COVERAGE PLOT FLOW WeldCL APR1-4304-18 Valve pipe I+- 1.25" .. ,

IGSCC TECHNIQUES AND EQUIPMENT WERE UTILIZED TO PREFORM THIS EXAM CODE VOLUME REQUIRED = 0.3 SQ INCHES CODE VOLUME ACHIEVED= 0.15 SQ INCHES= 50%

TOTAL CODE COVERAGE= 50%

(Red cross hatch- no coverage, Blue+ Green cross hatch- 60° coverage, Green crosshatch- 45° coverage)

Figure 3-24: 6" Valve to Pipe [APR1-4304-19-RB]

CODE COVERAGE PLOT FLOW WeldCL APR1-4304-19 Valve pipe

.239" 1+- 1.25" *I TOTAL CODE REQUIRED COVERAGE= 0.299 SQ INCHES CODE COVERAGE ACHIEVED= 0.149 SQ INCHES= 50%

(Red cross hatch- no coverage, Blue+ Green cross hatch- 60° coverage, Green crosshatch - 45° coverage)

E2-28

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-25: 4" Branch Connection to Pipe Weld [APR1-4307-21BC-RB]

• • • • • 60" 45' JO* CIRC SCAN I.T'

• TOTAL CODE VOlUME= 0 425 SQ IN 45 COVERAGE ACHIEVED s 2!-S~.C. FOR ~80 DEGREES OF THE PIPE 60" CODE COVCRAGE ACHIEVED

• 7*3% FOR 180 DEGREES OFT HE PIPE

,. AND 80 CODE COVERAGE OBTAINED" 36.1% FOR 180 DEGREES OF THE PIPE 30" CODE COVERAGE ACHIEVE 17.6% FOR 380 DEGREE AROUND THE PIPE TOTAL CODE VOLUME a 0.7525 SQ IN.

45" SCAN MAX OF 0.8 ON SUFACE OBTAININO CODE COVERAGE o27 SQ 1H TOTAL COOS COVERAGE ACHIEVE

• 17.9% FOR 110 DEORESS OF THE PIPE .

• • • -- 60° 600 ACHIEVED q% COVERAGE 45° 17.9% + 36.1%

• 27'\ CODE COVERAGE ACHIEVED FOR THE AX SCANS 11.6% CODE COVERAGE ACHIEVEDCIRC SCANS TOTAL CODE COVERAGEACIUEVEO*l7Yo+ 17.6%/2 *22l%

11.15"'

Figure 3-26: 14" Branch Connection to Pipe Weld [~PRl-4500-1]

FLOW as ""'-

PIPE t 4 ,.1 NOZZLE REQUIRED EXAMAAEA= 1.21 SQ. INCHES COVERPGE OBTAINED IN THE AXIAL SCAN DIRECTION= 50% (.605 SQ INCHES)

COVERAGE OBTAINED IN lHE CIRCUMFERENTIAL SCAN DIRECTION= 41% (.50 SQ. INCHES)

REQUIRED EXNJIINATION AAEA = 45.5%

E2-29

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-27: 10" Flange to Pipe [APR2-4511-10]

.77:J.". 775~. 7'fa**.75] ...15J:

~

FLANG£ PIP£ FLOW ,....,.

.fO% R£aUIR£D EXAM VOLUME OBTAINED Figure 3-28: 10" Pipe to Valve [APR1-4511-11]

. 7J.7... 7J.l.... 130 .. GBJ

II I.l..O ..

<t.

PIPE VALV£ FLOW SO!. REQUIRED EXAM VOLVM£ OBTAINf.D E2-30

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-29: 10" Tee to Elbow [APR1-4511-12]

/. b21_"

PIPf so~\o COVERAbE 08TAINED OF REQUIRED EXAM VDLVME E2-31