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 NOV 3 0 2018 Docket Nos.: 50-348 50-364 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Cheryl A. Gayheart Regulatory Affairs Director Joseph M. Farley Nuclear Plant Fourth 10-Year lntervallnservice Inspection Program 3535 Colonnade Parkway Birmingham, AL 35243 205 992 5316 tel 205 992 7601 fax cagayhea@ southernco.com NL-18-1384 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 1 0-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 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:

Reference:

Examination Category:

Item Number:

==

Description:==

Component Number:

1 IWB-2500, Table IWB-2500-1, ASME Code Case N-460, 8-D 83.110 Limited Examination Coverage See Tables RR-02.1 and RR-02.2 for a list of Component IDs

2.

Applicable Code Edition and Addenda

The Fourth 1 0-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 1 OCFR50.55a(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 1 0-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 1 00% 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 1 OCFR50.55a(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

8.

Precedents 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 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

Weld Exam Requirements Component ID Description (Figure No.)

(System)1 and Method PZR Upper Head to IW8*2500*7(b)

ALA1-2100-9 Safety Nozzle Volumetric Weld (UT)

(831)

PZR Upper ALA1-2100-Head to IW8-2500-7(b) 10 Safety Nozzle Volumetric Weld (UT)

(831)

PZR Upper ALA1-2100-Head to IW8-2500-7(b) 11 Safety Nozzle Volumetric Weld (UT)

(831)

PZR Upper ALA1-2100-Head to IW8-2500-7(b) 12 Spray Nozzle Volumetric Weld (UT)

(831)

PZR Upper ALA1-2100-Head to IW8-2500-7(b) 13 PORV Nozzle Volumetric Weld (UT)

(831)

PZR Lower ALA1-2100-Head to IW8-2500-7(b) 14 Surge Nozzle Volumetric Weld (UT)

(831)

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 Angle/

Category/

Outage Diameter/

Actual Item Examined Thickness Frequency (MHz) I Coverage Remarks Number Mode The examination was limited due to 8-D o* 1 2.251 Long component configuration. An Inner 83.110 1R22 6"/3.88" 45" /2.25/ Shear 75%

Radius UT and supplemental MT was so* 12.251 Shear performed with 100% coverage.

(Figures 2-1 and 2-3)

The examination was limited due to 8-D o* I 2.25/ Long component configuration. An Inner 83.110 1R27 6"13.88" 45" I 2.25 I Shear 78.6%

Radius UT and supplemental MT was 60" 12.251 Shear performed with 100% coverage.

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

The examination was limited due to 8-D o* I 2.25 I Long component configuration. An Inner 83.110 1A24 6"13.88" 45" 12.251 Shear 78.6%

Radius UT and supplemental MT was so* 12.251 Shear performed with 100% coverage.

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

The examination was limited due to 8-D o* I 2.251 Long component configuration. An Inner 83.110 1R24 4"13.88" 45" 12.25/ Shear 78.6%

Radius UT and supplemental MT was so* /2.251 Shear performed with 100% coverage.

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

The examination was limited due to 8-D o* I 2.251 Long component configuration. An Inner 83.110 1R24 6" 13.88" 45" 12.251 Shear 78.6%

Radius UT and supplemental MT was so* 12.251 Shear performed with 1 00% coverage.

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

The examination was limited due to 8-D o* 1 2.251 Long component configuration. An Inner 83.110 1R24 14"/3.88" 45" 12.251 Shear 75%

Radius UT and supplemental MT was 60" I 2.251 Shear performed with 100% coverage.

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

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

E1-4 3'd Interval Relief Request RR-6 RR-6 RR-6 RR-6 RR-6 I

RR-6

Weld Exam Requirements Component 10 Description (Figure No.)

(System)'

and Method PZR Upper Head to IW8*2500-7(b)

APR1-21aa-9 Safety Nozzle Volumetric Weld (UT)

(831)

PZR Upper APR1-21aa-Head to IW8-2500-7(b) 1a Safety Nozzle Volumetric Weld (UT)

(831)

PZR Upper APR1-21aa-Head to IW8-2500-7(b) 11 Safety Nozzle Volumetric Weld (UT)

(831)

PZR Upper APR1-21aa-Head to IW8-2500-7(b) 12 Spray Nozzle Volumetric Weld (UT)

(831)

PZR Upper APR1-2100-Head to IW8-2500-7(b) 13 PORV Nozzle Volumetric Weld (UT)

(831)

PZR Lower APR1-2100-Head to IW8-2500-7(b) 14 Surge Nozzle Volumetric Weld (UT)

(831)

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 Angle/

Category/

Outage Diameter/

Actual Item Examined Thickness Frequency (MHz) I Coverage Remarks Number Mode The examination was limited due to 8-D a* /2.251 Long component configuration. An Inner 83.11a 2R23 6"13.88" 45" 12.251 Shear 61.1%121 Radius UT and supplemental MT 60" 12.251 Shear was performed with 1 00% coverage.

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

The examination was limited due to 8-D a* 12.251 Long component configuration. An Inner 83.11a 2R23 6"13.88" 45" 12.25 I Shear 61.1%121 Radius UT and supplemental MT 6a* 12.251 Shear was performed with 1 aa% coverage.

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

The examination was limited due to 8-D a* 12.25 I Long component configuration. An Inner 83.11a 2R24 6" 13.88" 45" 12.25 I Shear so.a%121 Radius UT was performed with 60" 12.251 Shear 100% coverage.

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

The examination was limited due to 8-D a* 12.251 Long component configuration. An Inner 83.11a 2R24 4" 13.88" 45" 12.25 I Shear 5a.a%121 Radius UT was performed with 60" 12.251 Shear 100% coverage.

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

The examination was limited due to 8-D a* 12.251 Long component configuration. An Inner 83.11a 2R19 6" 13.88" 45" 12.251 Shear 75.a%

Radius UT and supplemental MT 60" 12.251 Shear was performed with 100% coverage.

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

The examination was limited due to 8-D a* 12.251 Long component configuration An Inner 83.11a 2R19 14"13.88" 45" 12.251 Shear 6t.a%

Radius UT and supplemental MT 60" 12.251 Shear was performed with 100% coverage.

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

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

3'd Interval Relief Request RR-6 RR-6 RR-6 RR-6 RR-6 RR-6

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 1 0-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 CLADDING VESSEL HEAD PRESSURIZER NOZZLE* TO. VESSEL WELD Figure 2-2: Typical 14" Nozzle Configuration E1-6

0' 45' 60' UPNol 100%

26.2%

10.5% u-100%

100%

100%

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-Year Interval 10 CFR 50.55a Request Number FNP-ISJ-RR-02 Taal 100%

26.2%

10.5%

Figure 2-3: Nozzle Limitations

[ALA 1-21 00-9]

ao*os 4~* OS DNNol IINII.iiiiith Total CW/Val CWilenalll 95.2%

100%

9!1.2%

100%

100%

705%

100%

70.5%

100%

100%

TOTAL COVERAGE - 75%

Total -

100%

100%

PZR UPPER HEAD Upstream CCW/Vol CCWilenalll 100%

100%

100%

100%

Figure 2-4: Nozzle Limitations

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

PZR UPPER HEAD Upstream Total -

100%

100%

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

UPNol UPII.onath 0'

100%

100%

45' 262%

100%

60' 10.5%

100%

UPNol UPII.onath 0'

69.8%

100%

45' 901%

100%

60' 906%

100%

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02 Tolal 100%

26.2%

10.5%

Tolal 69.8%

90.1%

90.6%

Figure 2-5: Nozzle Limitations

[ALA1-2100-14]

eo* us ONNcil-DNII.enath Tolal CWNol CWn..nath 95.2%

100%

95.2%

100%

100%

70.5%

100%

70.5%

100%

100%

TOTAL COVERAGE -75%

Figure 2-6: Nozzle Limitations

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

DNNol ON/length TOtal CWNcil CWII.onath 31.4%

100%

31.4%

62.9%

100%

161%

100%

16.1%

629%

100%

TOTAL COVERAGE-61.1%

E1-8 TOW -

100%

100%

TOW -

62.9"1.

62.9"1.

Heater sleeve penetrations PZR LOWER HEAD CCWNol ccw~

100%

100%

100%

100%

PZR UPPER HEAD Upstream CCWNol CCWII.onath 629%

100%

62.9%

100%

Total -

100%

100%

Total -

62_9"1.

62_9"1.

0' 45' 60' o*

45' 60' NOZZLE Downstream UPI\\Ial 50.0%

100%

80.7%

100%

90.3%

100%

UPNol Pn..nath 100%

100%

262%

100%

10.5%

100%

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-Year Interval 1 0 CFR 50.55a Request Number FNP-ISI-RR-02 Total 50.0%

80.7%

90.3%

Tolal 100%

26.2%

10.5%

Figure 2-7: Nozzle Limitations

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

CL 15" lliiiVol

~

Total CWNol CW~~Anc;~~>

19.3%

100%

19.3%

50.0%

100%

9.7%

100%

9.7%

50.0%

100%

TOTAL COVERAGE-50.0%

Figure 2-8: Nozzle Limitations

[APR1-2100-13]

ONNol OIUL...ath Total CWNol CWIL-952%

100%

95.2%

100%

100%

70.5%

100%

70.5%

100%

100%

TOTAL COVERAGE-75%

E1-9 PZR UPPER HEAD-3.88" (Sketch)

Upstream PZR UPPER HEAD Required Volume 5.6" x UA:lf-per!btg)"- 21.72 hln 0'-10.86mln Tolol.

50.0%

50.0%

Tolol -

100%

100%

45' -17.52 & 4.20 mIn 60' -19.62 & 2.10 mIn CCWNol 50.0%

100%

50.0%

100%

otal 50.0%

50.0%

PZR UPPER HEAD Upstream CCWNol ccwn..nau.

Total 100%

100%

100%

100%

100%

100%

0' 45° so*

NOZZLE Downstream UPNol UP/Lonalh 69.8%

100%

90.1%

100%

906%

100%

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-Year Interval 10 CFR 50.55a Request Number FNP-181-RR-02 Total 69.8%

90.1%

90.6%

Figure 2-9: Nozzle Limitations

[APR1-2100-14]

DNNol OIIIIAaatll Total C:WNol C:W/Lenalll 31.4%

100%

31.4%

62.9°,(,

100%

16.1%

100%

16.1%

62.9%

100%

TOTAL COVERAGE-61%

E1-10 Total -

62.9%

62.9%

Heater sleeve penetrations PZR LOWER HEAD CC:WNol CC:W/Lenalll Tobl 62.9%

100%

62.9%

62.9%

100%

62.9%

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

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Relief Request In Accordance with 1 0 CFR 50.55a(g)(5)(iii)

--lnservice Inspection Impracticality--

1.

ASME Code Component(s) Affected Code Class:

Reference:

Examination Category:

Item Number:

==

Description:==

Component Number:

1&2 IWB-2500, Table IWB-2500-1, IWC-2500, Table IW8-2500-1, ASME Code Case N-460, ASME Code Case N-716, Table 1 8-F, 8-J, C-F-1, R-A 85.70, 89.11, 89.21, C5.11, R1.11, R1.16, R1.20 Limited Examination Coverage 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 1 OCFR50.55a 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 1 00% of the weld length.

E2-1

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 1 00%" 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 1 OCFR50.55a(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 1 0-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 1 OCFR50.55a(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

8.

Precedents 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 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

Weld Component ID Description (System)'

Safe*End to ALA1-4100-SG Nozzle 26RDM Weld (813)

Valve to Pipe ALA 1-4103-4 Weld (813)

Valve to Pipe ALA 1-4104-4 Weld (813)

Pipe to Valve ALA 1-4104-5 Weld (E11)

ALA 1-41 08 Valve to Pipe R8 Weld (E21)

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/

Exam Angle/

Appendix Requirements Item Number Outage Diameter/

Frequency (UHz) I Actual VIII (Figure No.)

credited Examined Thickness Mode Coverage Qualified and Method (current R-A Exam Item#)

IW8-2500-8 45' 12.00 I Long Surface and 8-F, 8570 34' 11.50 I Long Volumetric (R-A, R1.20) 1R23 29"14.75" 40' I 1.00 I Long 52.1%

Yes

{MT){UT) 40' 11.50 I Shear 34' 11.50 I Shear IW8-2500-8 45' 12.251 Shear Surface and 8-J. 89.11 1R22 6"1 0.75" 70' 12.251 Shear 50.0%

Yes Volumetric (R-A, R1.11) 60' I 2.00 I Long (PT){UT)

IW8-2500-8 Surface and 8-J, 89.11 1R22 6"1 0.75" 45' 12.251 Shear 50.0%

Yes Volumetric (R-A,R1.11) 60' I 2.00 I Long (PT){UT)

IW8-2500-8 Surface and 8-J, 89.11 45' 12.251 Shear Volumetric (R-A, Rl.ll/16) 1R22 6" I 0.75" 60' I 2.00 I Long 50.0%

Yes

{PT)(UT)

IW8-2500-8(c)

IW8-2500-9, o* I 4.00 I Long 10,& 11 R-A, R1.11 1R27 3" I 0.438" 45' 15.00 I Shear 50.0%

Yes Volumetric 70' 12.251 Shear (UT)

Remarks The examination was limited due to component configuration.

(Figure 3-1)

The examination was limited due to component configuration.

Additional inspections performed for MRP-146 during 1 R24, 1 R25, 1R26, & 1R27.

(Figures 3-2 & 3-2a)

The examination was limited due to component configuration.

Additional inspections performed for MRP-146 during 1 R24, 1 R25, 1R26, & 1R27.

(Figure 3-3)

The examination was limited due to component configuration.

(Figure 3-4)

The examination was limited due to component configuration.

(Figure 3-5)

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

HHSI/CVCS (E21)

E2-5

Weld Component 10 Description (System)'

ALA1-4108-Tee to Pipe 14BW-RB Weld (B13)

ALA 1-4202 Pipe to Elbow RB Weld (B13)

ALA 1-4202 Valve to Pipe RB Weld (B13)

Valve to Pipe ALA 1-4204-4 Weld (B13)

ALA 1-4204 Pipe to Valve RB Weld (E21)

ALA1-4209-Flange to 11BW-RB Pump Weld (E21)

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 Requirements Item Number Outage Diameter/

Exam Angle/

Actual VIII (Figure No.)

credited Examined Thickness Frequency (MHz) I Coverage Qualified and Method (current R-A Mode Exam Item#)

IWB-2500-S(c)

IWB-2500-9, o* /4.00 I Long 10, & 11 R-A, R1.20 1R26 2"1 0.344 45" 15.00 I Shear 50.0%

Yes Volumetric 70" I 2.251 Shear (UT)

IWB-2500-S(c)

IWB-2500-9, 45" 12.251 Shear 10,& 11 R-A, R1.11 1R24 6"1 0.75" 60" 12.251 Shear 75.0%

Yes Volumetric 60" I 2.00 I Long (UT)

IWB-2500-S(c)

IWB-2500-9, 45" /2.251 Shear 10, & 11 R-A, R1.11 1R24 6"1 0.75" 60" 12.251 Shear 50.0%

Yes Volumetric 60" 12.00 I Long (UT)

IWB-2500-S(c)

IWB-2500-9, 45" 12.251 Shear 10, & 11 R-A, R1.11 1R24 6"1 0.75" 60" I 2.251 Shear 50.0%

Yes Volumetric 60" 12.00 I Long (UT)

IWB-2500-S(c)

IWB-2500-9, 45" 12.251 Shear 10, & 11 R-A, R1.16 1R24 6"1 0.75" 60" 12.251 Shear 50.0%

Yes Volumetric 60' 12.00 I Long (UT)

IWB-2500-S(c)

IWB-2500-9, 45' I 5.00 I Shear 10,& 11 R-A, R1.20 1R24 2"1 0.196" 70' 12.251 Shear 48.0%

Yes Volumetric (UT)

Remarks The examination was limited on the upstream side due to the tee component configuration.

(Figure 3-6)

The examination was limited due to a welded support and ID Pad obstruction. Additional inspections performed for MRP-146 during 1R25, 1R26, & 1R27.

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

The examination was limited due to component configuration.

Additional inspections performed for MRP-146 during 1R25, 1R26,

& 1R27 (Fi!lures 3-8 & 3-Ba)

The examination was limited due to component configuration.

Additional inspections performed for MRP-146 during 1R25, 1R26,

& 1R27 (Figures 3-9 & 3-9a)

The examination was limited due to component configuration.

(Figures 3-10 & 3-10a)

The examination was limited due to component configuration.

(Figure 3-11)

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

E2-6

Weld Component ID Description (System)'

SG Sale-End APR1*4300-to Nozzle 23RDM Weld (913)

SG Safe-End APR1-4300*

to Nozzle 24RDM Weld (913)

Pipe to Valve APR1-4101*B Weld (913)

Pipe to Pipe APR1-4102*

Weld 2-R9 (E21)

Pipe to Valve APR1-4104*

Weld 30 (E21}

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 Appendix Exam Category/

Exam Angle/

Requirements Item Number Outage Diameter/

Frequency (MHz) I Actual VIII (Figure No.)

credited Examined Thickness Mode Coverage Qualified and Method (current R-A Exam Item II}

IW9-2500-B 45' /2.00 I Long 34 * /1.50 I Long Surtace and 9-F, 95.70 2R19 29" /4.75" 40' 11.00 I Long 52.1%

Yes Volumetric (R-A, R1.20) 40' 11.50 I Shear (MT)(UT) 34' 11.50 I Shear IW9-2500-B 45' 12.00 I Long 34' 11.50 I Long Surtace and 9-F, 95.70 2R19 31" 14.75" 40' 11.00 I Long 52.1%

Yes Volumetric (R*A, R1.20) 40' 11.50 I Shear (MT)(UT) 34' 11.50 I Shear IW9-2500-B 45' 12.251 Shear Surtace and 9-J, 99.11 2R20 12" 11.125" 60' 12.0 I Long 50.0%

Yes Volumetric (R-A,R1.11) 70' 12.251 Shear (PT)(UT)

IW9-2500*B(c)

IW9-2500*9, 45' 12.251 Shear 10, & 11 R-A, R1.20 2R24 12" 11.125" 60' I 2.0 I Long 75.0%

Yes Volumetric 60' I 2.25 I Shear (UT)

IW9-2500-B 45' 12.251 Shear Surtace and 9-J, 99.11 2R22 6" I 0.719" 60' 12.251 Shear 50.0%

Yes Volumetric (R-A. R1.16) 60' I 2.00 I Long (PT)(UT)

Remarks The examination was limited due to component configuration.

(Figure 3-1)

The examination was limited due to component configuration.

(Figure 3-1)

The examination was limited due to component configuration.

Additional inspections pertormed for MRP-146 during 2R22.

(Figure 3-12}

The examination was limited due to a box restraint.

(Figure 3-13)

The examination was limited due to component configuration.

(Figure 3-14)

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

HHSI/CVCS {E21)

E2-7 I

J

Weld Component 10 Description (System )I APRl-4106*

Valve to Pipe 8*RB Weld (E21)

Pipe to APR1-4106*

Branch 11-RB Connection Weld (E21)

APR1-4108*

Pipe to Valve 11-RB Weld (813)

APR1-4108-Valve to Pipe 12-RB Weld (E21)

APR1-4108*

Pipe to Valve 13-RB Weld (E21)

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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/

Exam Angle/

Appendix Requirements Item Number Outage Diameter/

Frequency (MHz) I Actual VIII (Figure No.)

credited Examined Thickness Mode Coverage Qualified and Method (current R-A Exam Item#)

IWB*2500*8(c)

IWB-2500*9, 45" 15.00 I Shear 10, & 11 R*A, R1.11 2R23 3" I 0.438" 70" 12.251 Shear 49.4%

Yes Volumetric (UT)

IWB-2500-8(c)

IWB-2500-9, 45" 15.00 I Shear 10, & 11 R-A,R1.11 2R23 3" I 0.438" 70" 12.251 Shear 50.0%

Yes Volumetric (UT)

IWB-2500*8(c)

IWB-2500-9, 45" I 5.00 I Shear 10, & 11 R-A, R1.11 2R24 3" I 0.438" 60" I 5.00 I Shear 50.0%

Yes Volumetric 70" 12.251 Shear (UT)

IWB-2500-8(c)

IWB-2500-9, 45" 15.00 I Shear 10, & 11 R-A, R1.11 2R24 3" I 0.438" 60" 15.00 I Shear 50.0%

Yes Volumetric 70" 12.251 Shear (UT)

IWB-2500-8(c)

IWB-2500-9, 45" I 5.00 I Shear 10, & 11 R-A, R1.11 2R24 3" I 0.438" 60" I 5.00 I Shear 50.0%

Yes Volumetric 70" I 2.25 I Shear (UT)

Remarks The examination was limited due to component configuration and a welded plate.

(Figures 3-15 and 3-15a)

The examination was limited due to component configuration.

Additional inspections performed for MRP-146 during 2R21, 2R24, &

2R25.

(Figure 3-16)

The examination was limited due to component configuration.

(Figure 3-17)

The examination was limited due to component configuration.

(Figure 3-18)

The examination was limited due to component configuration.

(Figure 3-19)

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

HHSI/CVCS (E21)

E2-8

Weld Component ID Description (System)'

Pipe to Valve APR1-4301-8 Weld (813)

APR1-4302-Elbow to Pipe 11-R8 Weld (813)

APR1-4302-Pipe to Elbow 12-R8 Weld (813)

Pipe to Valve APR1-4304-Weld 18 (EH)

APR1-4304-Valve to Pipe 19-R8 Weld (EH)

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 Requirements Item Number Outage Diameter/

Exam Angle/

Actual VIII (Figure No.)

credited Examined Thickness Frequency (MHz) I Coverage Qualified and Method (current R-A Mode Exam ltemltl IW8-2500-8 Surface and B-J, 89.11 2R20 45' 12.251 Shear Volumetric (R-A.RU1) 12" I 1.125" eo* 12.00 I Long 50.0%

Yes (PT)(UT)

IW8-2500-8(c)

IW8-2500-9, 10, & 11 R-A, R1.20 2R25 12" I 1.125" 45" 12.251 Shear 75.0%

Yes Volumetric (UT)

IW8-2500-8(c)

IW8-2500-9, 10, & 11 R-A, R1.20 2R25 12"1 1.125" 45" 12.251 Shear 87.0%

Yes Volumetric (UT)

IW8-2500-8 Surface and 8-J, 89.11 45" 11.50 I Shear 2R21 6" I 0.719" Volumetric (R-A, R1.11/16) so* 12.00 I Long 50.0%

Yes (PT)(UT)

IW8-2500-8(c)

IW8-2500-9, 45" 12.251 Shear 10, & 11 R-A,RU1 2R23 6" I 0.719" eo* 12.0 I Long 50.0%

Yes Volumetric eo* 12.25 I Shear (UT)

Remarks The examination was limited due to component configuration.

(Figure 3-20)

The examination was limited from 18" to 28" due to a whip restraint.

(Figures 3-12, 3-21a, 3-21b & 3-21c)

The examination was limited from 18" to 28" due to a whip restraint.

(Figuras 3-22, 3-22a, 3-22b, & 3-22c)

The examination was limited due to component configuration.

(Figure 3-23)

The examination was limited due to component configuration.

Additional inspections performed for MRP-146 during 2R21, 2A22, &

2R24.

(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

Weld Component 10 Description (System)'

Branch APA1-4307-Connection 21BC-AB Weld (E21)

Branch APA1-4500-1 Connection to Pipe Weld (613)

Flange to APA2-4511-10 Pipe Weld (E11)

Pipe to Valve APA2-4511-11 Weld (E11)

Tee to Elbow APA2-4511-Weld 12 (E11)

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 Appendix Exam Category I Exam Angle/

Requirements Item Number Outage Diameter/

Frequency (MHz) I Actual VIII (Figure No.)

credited Examined Thickness Coverage Qualified Mode Exam and Method (current R-A Item#)

IWB-2500-B(c)

IWB-2500*9, 30' /2.25/ Shear 10, & 11 A-A, A1.11 2A25 4"1 0.719" 45' /2.251 Shear 22.3%

Yes Volumetric 60' I 2.251 Shear (UT)

IWB-2500*8 45' 11.50 I Shear Surface and 6-J, 69.11 2A19 14"11.4" 60' 11.50 I Shear 45.5%

Yes Volumetric (A*A. A1.11) 60' 12.00 I Shear (PT)(UT)

IWC-2500-7 45' 12.251 Shear Surface and C-F-1, C5.11 2A20 10"/ 0.719" 60' 12.251 Shear 50.0%

Yes Volumetric (A-A, A1.11) 60' I 2.00 I Long (PT)(UT)

IWC-2500-7 45' 12.251 Shear Surface and C-F-1, C5.11 2A20 10"/ 0.719" 60' I 2.251 Shear 50.0%

Yes Volumetric (A-A, A1.11) 60' 12.00 I Long (PT)(UT)

IWC-2500-7 Surface and C-F-1, C5.11 45' 12.251 Shear 50.0%

Yes 2A20 10"11.2" Volumetric (A*A, A1.11) 60' 12.00 I AL (PT)(UT)

Remarks The examination was limited due to component configuration and thickness changes. During outage 2A25, component ID APA1-4208*

23BC-AB was inspected to 1 00%

coverage with similar degradation method as added assurance.

fFiaure 3-251 The examination was limited due to component configuration.

(Figure 3-26)

The examination was limited due to component configuration.

(Figure 3-27)

The examination was limited due to component configuration.

(Figure 3-28)

The examination was limited due to I component configuration.

(Figure 3-29)

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 1 0-Year Interval 1 0 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]

Valve 3.63 square inches required for complete coverage for circumference of the weld.

1.90 square inches achieved with 34 degree exam angle-52% for the circumference of the weld.

~~.16 square inches achieved

~

wilh 45 degree exam angle* 4.4% for !he circumference ofthe weld 3.63 square inches achieved in the circumferential¥:811 directions-all elllUTI angles I 00% for the 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 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

Valve Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 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

45 Dee:

Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 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

TEE Pipe 11 Elbow I Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-6: 2" Tee to Pipe [ALA1-4108-14BW-RB]

FLOIJ

~

~

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

CODE COVERAGE PWT (AREA LIMITATION ONLY) Not to Seale

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

l 21" 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 J

Valve 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 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

Valve 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 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

Valve 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 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

Flange Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 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 1 0-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 f:LP£

-FLOW~

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

VALV£ ELOW...._

L rElfEIH Of COVEP.A6E AlllAINfO 7.5'"'

E2-19

Valve Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 1 0-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 I+-- 1.5"

... 1 pipe 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 Welded plate

.43r pipe FLOW WeldCL APRl-4106-8 I+-- 1.05"

... 1

.146" 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 1 0-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 1 0-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-17: 3" Pipe to Valve [APR1-4108-11-RB]

r--~-~~-

---

_.;a..----v.._ -----

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

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

I

(

---'v' --- ___. ________ _

/*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 1 0-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::

\\.o~

l.~'il I. 111.

/.lOb PtPE.

/.oCJ 16 ElCAM1t-1A-r1fl~

1/o~un£.:

ufSTAEA~ Ax - 15%

UPSTP.EP.I\\1 C..lll..C..- t5*to HLI(,HT- 0.31" uJ'bTH ~ '2..1..

LE~bTH* 'H)*

Dow!'o\\STIU.AM Ax- /5°/Q 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:

E2-24 I

I I

I 1:::.====1 I

I WELb II

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

1.101 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...

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

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~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 1 0 CFR 50.55a Request Number FNP-181-RR-03 Limited on back side of pipe E2-27

Valve 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 I+- 1.25" pipe 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)

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

CODE COVERAGE PLOT FLOW WeldCL APR1-4304-19 1+- 1.25"

• I

.239" pipe 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 1 0-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

• • • -- 60° 45° 30" CODE COVERAGE ACHIEVE 17.6% FOR 380 DEGREE AROUND THE PIPE 11.15"'

TOTAL CODE VOLUME a 0.7525 SQ IN.

45" SCAN MAX OF 0.8 ON SUFACE OBT AININO CODE COVERAGE o 27 SQ 1H TOTAL COOS COVERAGE ACHIEVE

• 17.9% FOR 110 DEORESS OF THE PIPE.

600 ACHIEVED q% COVERAGE 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%

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 1 0-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..

II GBJ

I.l..O..

<t.

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

PIPf 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]

/. b 21_"

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