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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)
	v • d • e

~ 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%

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%

705%

100%

70.5%

100%

TOTAL COVERAGE - 75%

Total -

100%

PZR UPPER HEAD Upstream CCW/Vol CCWilenalll 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%

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%

60' 95%

100%

95%

10.5%

100%

10.5%

100%

TOTAL COVERAGE-78.6%

E1-7

UPNol UPII.onath 0'

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%

70.5%

100%

70.5%

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%

TOW -

62.9"1.

Heater sleeve penetrations PZR LOWER HEAD CCWNol ccw~

100%

PZR UPPER HEAD Upstream CCWNol CCWII.onath 629%

100%

62.9%

100%

Total -

100%

Total -

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%

70.5%

100%

70.5%

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%

Tolol -

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%

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%

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

100%

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

/ *

~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

v t e Letter Sequence Request
EPID:L-2018-LLR-0195, Fourth 10-Year Interval Inservice Inspection Program ISI Program Update: Notification of Impractical ASME Code Requirements (Open)
Initiation Request Acceptance Administration Questions 25 February 2019 Answers 27 March 2019 Results Approval... Other: ML19213A194
MONTHYEAR2019-02-25 [Table View]

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

Latest revision as of 09:58, 5 January 2025

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Enclosures:

Reference:

Applicable Code Edition and Addenda

Reference:

Applicable Code Edition and Addenda

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@@ Line 17: / Line 17: @@
 =Text=
-{{#Wiki_filter:~ 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:
+{{#Wiki_filter:~ 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 10-year lSI program interval. Requests for relief are enclosed.
+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
+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
+: 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
+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--
+--lnservice Inspection Impracticality--
-: 1. ASME Code Component(s) Affected Code Class:                     1
+: 1.
+ASME Code Component(s) Affected Code Class:
 ==Reference:==
-IWB-2500, Table IWB-2500-1, ASME Code Case N-460, Examination Category:           8-D Item Number:                    83.110
+Examination Category:
+Item Number:
-== Description:==
+==
-Limited Examination Coverage Component Number:               See Tables RR-02.1 and RR-02.2 for a list of Component IDs
+Description:==
+Component Number:
+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.
-===2. Applicable Code Edition and Addenda===
+===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.
+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.
+: 3.
-: 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.
+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 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.
+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 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
+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 10CFR50.55a(g)(5)(iii).
+: 5.
-: 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.
+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).
-: 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.
+: 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
-Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-02
+Weld Exam Requirements Component ID Description (Figure No.)
-: 8. Precedents
+(System)1 and Method PZR Upper Head to IW8*2500*7(b)
-* Relief Request 13R-23 was authorized for Limerick Generating Station, Units 1 and 2 by NRC SEdated August 7, 2018 (ADAMS Accession No. ML18192C172).
+ALA1-2100-9 Safety Nozzle Volumetric Weld (UT)
-* 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).
+(831)
-: 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.
+PZR Upper ALA1-2100-Head to IW8-2500-7(b) 10 Safety Nozzle Volumetric Weld (UT)
-E1-3
+(831)
+PZR Upper ALA1-2100-Head to IW8-2500-7(b) 11 Safety Nozzle Volumetric Weld (UT)
-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/
+(831)
-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.
+PZR Upper ALA1-2100-Head to IW8-2500-7(b) 12 Spray Nozzle Volumetric Weld (UT)
-(831)                                                                                     (Figures 2-1 and 2-3)
+(831)
-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.
+PZR Upper ALA1-2100-Head to IW8-2500-7(b) 13 PORV Nozzle Volumetric Weld (UT)
-(831)                                                                                     (Figures 2-1 and Fig. 2-4)
+(831)
-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.
+PZR Lower ALA1-2100-Head to IW8-2500-7(b) 14 Surge Nozzle Volumetric Weld (UT)
-(831)                                                                                     (Figures 2-1 and Fig. 2-4)
+(831)
-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.
+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/
-(831)                                                                                     (Figures 2-1 and Fig. 2-4)
+Category/
-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.
+Outage Diameter/
-I (831)                                                                                     (Figures 2-1 and Fig. 2-4)
+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%
-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.
+Radius UT and supplemental MT was so* 12.251 Shear performed with 100% coverage.
-(831)                                                                                     (Figures 2-2 and Fig. 2-5)
+(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
+E1-4 3'd Interval Relief Request RR-6 RR-6 RR-6 RR-6 RR-6 I
+RR-6
-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/
+Weld Exam Requirements Component 10 Description (Figure No.)
-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.
+(System)'
-(831)                                                                                     (Figures 2-1 and Fig. 2-6)
+and Method PZR Upper Head to IW8*2500-7(b)
-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.
+APR1-21aa-9 Safety Nozzle Volumetric Weld (UT)
-(831)                                                                                     (Figures 2-1 and Fig. 2-6)
+(831)
-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.
+PZR Upper APR1-21aa-Head to IW8-2500-7(b) 1a Safety Nozzle Volumetric Weld (UT)
-(831)                                                                                      (Figures 2-1 and Fig. 2*7)
+(831)
-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.
+PZR Upper APR1-21aa-Head to IW8-2500-7(b) 11 Safety Nozzle Volumetric Weld (UT)
-(831)                                                                                      (Figures 2-1 and Fig. 2-7)
+(831)
-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.
+PZR Upper APR1-21aa-Head to IW8-2500-7(b) 12 Spray Nozzle Volumetric Weld (UT)
-(831)                                                                                      (Figures 2-1 and Fig. 2-8)
+(831)
-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.
+PZR Upper APR1-2100-Head to IW8-2500-7(b) 13 PORV Nozzle Volumetric Weld (UT)
-(831)                                                                                      (Figures 2-2 and Fig. 2-9)
+(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 ).
+'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 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-~
+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 VESSEL HEAD CLADDING PRESSURIZER NOZZLE*TO.VESSEL WELD Figure 2-2: Typical 14" Nozzle Configuration E1-6
+WELD DEPOSITED CLADDING VESSEL HEAD 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
+' 45' 60' UPNol 100%
-[ALA1-21 00-9]
+.2%
-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%        -          .                    -           -          -       -          -           -
+.5% u-100%
-' 26.2%  100%     26.2%    95.2%      100%         9!1.2%  100%         100%      100%     100%       100%       100%
+%
-' 10.5%  100%     10.5%    705%       100%        70.5%    100%         100%      100%     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%
+.2%
+.5%
+Figure 2-3: Nozzle Limitations
+[ALA 1-21 00-9]
+ao*os 4~* OS DNNol IINII.iiiiith Total CW/Val CWilenalll 95.2%
+%
+!1.2%
+%
+%
+%
+%
+.5%
+%
+%
 TOTAL COVERAGE - 75%
+Total -
+%
+%
+PZR UPPER HEAD Upstream CCW/Vol CCWilenalll 100%
+%
+%
+%
 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%
+PZR UPPER HEAD Upstream Total -
-'  97%   100%      97%     26.2%      100%        26.2%   100%         100%      100%     100%       100%       100%
+%
-'  95%   100%      95%     10.5%      100%        10.5%   100%         100%      100%     100%       100%       100%
+%
-TOTAL COVERAGE- 78.6%
+UPNOI UP/Unalh Total DN/Val
-E1-7
+~
+Total CW/Vol CWilenatll Total CCW/Vol CCWIUncdll Total o*
+%
+o%
+%
+' 97%
+%
+%
+.2%
+%
+.2%
+%
+%
+%
+%
+%
+%
+' 95%
+%
+%
+.5%
+%
+.5%
+%
+%
+%
+%
+%
+%
+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
+UPNol UPII.onath 0'
+%
+%
+' 262%
+%
+' 10.5%
+%
+UPNol UPII.onath 0'
+.8%
+%
+' 901%
+%
+' 906%
+%
+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%
+.2%
+.5%
+Tolal 69.8%
+.1%
+.6%
+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%         -         -         -             -           -        -                    -
+eo* us ONNcil-DNII.enath Tolal CWNol CWn..nath 95.2%
-' 262%   100%       26.2%    95.2%      100%      95.2%  100%    100%        100%     100%      100%      100%
+%
-' 10.5%  100%       10.5%    70.5%      100%      70.5%  100%    100%        100%     100%      100%      100%
+.2%
+%
+%
+.5%
+%
+.5%
+%
+%
 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'
+DNNol ON/length TOtal CWNcil CWII.onath 31.4%
-' 69.8%
+%
-%
+.4%
+.9%
+%
+%
+%
+.1%
+%
+%
+TOTAL COVERAGE-61.1%
+E1-8 TOW -
+%
+%
+TOW -
+.9"1.
+.9"1.
+Heater sleeve penetrations PZR LOWER HEAD CCWNol ccw~
+%
+%
 %
 %
-.8%
+PZR UPPER HEAD Upstream CCWNol CCWII.onath 629%
-.1%     31 .4%
 %
-.4%
 .9%
 %
-.9"1.
+Total -
 %
 %
+Total -
 _9"1.
-' 906%   100%       90.6%     161%      100%      16.1%  629%    100%        62.9"1. 62.9%     100%      62_9"1.
+_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
+' 45' 60' o*
+' 60' NOZZLE Downstream UPI\\Ial 50.0%
+%
+.7%
+%
+.3%
+%
+UPNol Pn..nath 100%
+%
+%
+%
+.5%
+%
+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%
+.7%
+.3%
+Tolal 100%
+.2%
+.5%
+Figure 2-7: Nozzle Limitations
 [APR1-2100-11, APR1-2100-12]
-CL NOZZLE                                                                   PZR UPPER HEAD- 3.88" (Sketch)''
+CL 15" lliiiVol
-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%          .        .          .                  .          .           .           .           .
+~
-' 80.7%     100%      80.7%     19.3%      100%       19.3%    50.0%     100%       50.0%       50.0%       100%         50.0%
+Total CWNol CW~~Anc;~~>
-' 90.3%     100%      90.3%      9.7%      100%       9.7%     50.0%     100%       50.0%       50.0%       100%         50.0%
+.3%
-TOTAL COVERAGE- 50.0%
+%
+.3%
+.0%
+%
+.7%
+%
+.7%
+.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%           -        -           .       .          .           -          .            .
+ONNol OIUL...ath Total CWNol CWIL-952%
-' 262%       100%      26.2%      952%       100%       95.2%    100%        100%      100%         100%         100%        100%
+%
-'  10.5%     100%      10.5%      70.5%      100%       70.5%    100%        100%      100%         100%         100%        100%
+.2%
-TOTAL COVERAGE- 75%
+%
-E1-9
+%
+.5%
+%
+.5%
+%
+%
+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.
+.0%
+.0%
+Tolol -
+%
+%
+' -17.52 & 4.20 mIn 60' -19.62 & 2.10 mIn CCWNol 50.0%
+%
+.0%
+%
+otal 50.0%
+.0%
+PZR UPPER HEAD Upstream CCWNol ccwn..nau.
+Total 100%
+%
+%
+%
+%
+%
-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
+' 45&deg; so*
+NOZZLE Downstream UPNol UP/Lonalh 69.8%
+%
+.1%
+%
+%
+%
+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%
+.1%
+.6%
+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%       -         -         -      -         -          -      -         -           -
+DNNol OIIIIAaatll Total C:WNol C:W/Lenalll 31.4%
-&deg; 90.1%    100%      90.1%    31 .4%   100%       31.4%  62.9&deg;,(,   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%
+.4%
-TOTAL COVERAGE- 61%
+.9&deg;,(,
-E1-10
+%
+.1%
+%
+.1%
+.9%
+%
+TOTAL COVERAGE-61%
+E1-10 Total -
+.9%
+.9%
+Heater sleeve penetrations PZR LOWER HEAD CC:WNol CC:W/Lenalll Tobl 62.9%
+%
+.9%
+.9%
+%
+.9%
-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
+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 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)
+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--
+--lnservice Inspection Impracticality--
-: 1. ASME Code Component(s) Affected Code Class:                     1&2
+: 1.
+ASME Code Component(s) Affected Code Class:
 ==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
+Examination Category:
+Item Number:
-== Description:==
+==
-Limited Examination Coverage Component Number:               See Tables RR-03.1 and RR-03.2 for a list of Component IDs
+Description:==
+Component Number:
+&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.
-===2. Applicable Code Edition and Addenda===
+===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.
+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.
+: 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.
+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 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.
+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 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.
+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 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.
+: 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.
@@ Line 193: / Line 426: @@
 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.
+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.
+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).
+: 5.
-: 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.
+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).
-: 7. Duration of Proposed Alternative Relief is requested for the Fourth lSI Interval for FNP, Units 1 and 2.
+: 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
-Southern Nuclear Operating Company Farley Nuclear Plant, Units 1 & 2 Fourth 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03
+Weld Component ID Description (System)'
-: 8. Precedents
+Safe*End to ALA1-4100-SG Nozzle 26RDM Weld (813)
-* Relief Request 13R-23 was authorized for Limerick Generating Station, Units 1 and 2 by NRC SEdated August 7, 2018 (ADAMS Accession No. ML18192C172).
+Valve to Pipe ALA 1-4103-4 Weld (813)
-* 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).
+Valve to Pipe ALA 1-4104-4 Weld (813)
-E2-4
+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/                                                          Appendix Weld                                                                   Exam Angle/
+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/
-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#)
+Exam Angle/
-' 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.
+Appendix Requirements Item Number Outage Diameter/
-RDM          Weld         Volumetric      (R-A, R1 .20) 40' 11.50 I Shear                    (Figure 3-1)
+Frequency (UHz) I Actual VIII (Figure No.)
-(813)         {MT){UT) 34' 11.50 I Shear The examination was limited due IW8-2500-8                                                                                    to component configuration.
+credited Examined Thickness Mode Coverage Qualified and Method (current R-A Exam Item#)
-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.
+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 IW8-2500-8                                                                                    to component configuration.
+The examination was limited due 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)
+Additional inspections performed for MRP-146 during 1 R24, 1 R25, 1R26, & 1R27.
-(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.
+The examination was limited due to component configuration.
-Volumetric   (R-A, Rl .ll/16)                           60' I 2.00 I Long (E11)                                                                                                     (Figure 3-4)
+(Figure 3-4)
-{PT)(UT)
+The examination was limited due to component configuration.
-IW8-2500-8(c)
+(Figure 3-5)
-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/
+Weld Component 10 Description (System)'
-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#)
+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)
-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.
+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%
-(B13)       Volumetric                                          70" I 2.251 Shear (Figure 3-6)
+Yes Volumetric 60" 12.00 I Long (UT)
-(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.
+IWB-2500-9, 45" 12.251 Shear 10, & 11 R-A, R1.16 1R24 6"1 0.75" 60" 12.251 Shear 50.0%
-RB (E21)       Volumetric                                          60' 12.00 I Long                      (Figures 3-10 & 3-10a)
+Yes Volumetric 60' 12.00 I Long (UT)
-(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.
+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%
-BW-RB                                                                        70' 12.251 Shear (E21)       Volumetric                                                                                (Figure 3-11)
+Yes Volumetric (UT)
-(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
-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/
+Weld Component ID Description (System)'
-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}
+SG Sale-End APR1*4300-to Nozzle 23RDM Weld (913)
-' /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.
+SG Safe-End APR1-4300*
-RDM          Weld        Volumetric     (R-A, R1.20) 40' 11 .50 I Shear                     (Figure 3-1)
+to Nozzle 24RDM Weld (913)
-(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.
+Pipe to Valve APR1-4101*B Weld (913)
-RDM         Weld         Volumetric     (R*A, R1 .20) 40' 11 .50 I Shear                     (Figure 3-1)
+Pipe to Pipe APR1-4102*
-(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 .
+Weld 2-R9 (E21)
-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)
+Pipe to Valve APR1-4104*
-(913)                                                                70' 12.251 Shear                       for MRP-146 during 2R22.
+Weld 30 (E21}
-(PT)(UT)
+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}
-IW9-2500*B(c)
+The examination was limited due to a box restraint.
-Pipe to Pipe   IW9-2500*9,                                               45' 12.251 Shear                       The examination was limited due to APR1-4102*
+(Figure 3-13)
-Weld           10, & 11      R-A, R1 .20      2R24     12" 11.125"    60' I 2.0 I Long    75.0%     Yes     a box restraint.
+The examination was limited due to component configuration.
--R9 (E21)        Volumetric                                              60' I 2.25 I Shear                     (Figure 3-13)
+(Figure 3-14)
-(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.
-                       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
+E2-7 I
+J
-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/
+Weld Component 10 Description (System )I APRl-4106*
-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#)
+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)
-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.
+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%
-(E21)       Volumetric (Figures 3-15 and 3-15a)
+Yes Volumetric (UT)
-(UT)
+IWB-2500-8(c)
-The examination was limited due to Pipe to    IWB-2500-8(c) component configuration.
+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%
-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, &
+Yes Volumetric (UT)
-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*
+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%
-Weld         10, & 11      R-A, R1.11      2R24     3" I 0.438" 60" I 5.00 I Shear   50.0%     Yes     component configuration.
+Yes Volumetric 70" 12.251 Shear (UT)
--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.
+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%
--RB (E21)       Volumetric                                          70" 12.251 Shear                       (Figure 3-18)
+Yes Volumetric 70" 12.251 Shear (UT)
-(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*
+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%
-Weld         10, & 11      R-A, R1.11      2R24     3" I 0.438" 60" I 5.00 I Shear   50.0%     Yes     component configuration.
+Yes Volumetric 70" I 2.25 I Shear (UT)
--RB (E21)       Volumetric                                          70" I 2.25 I Shear                     (Figure 3-19)
+Remarks The examination was limited due to component configuration and a welded plate.
-(UT)
+(Figures 3-15 and 3-15a)
+The examination was limited due to component configuration.
+Additional inspections performed for MRP-146 during 2R21, 2R24, &
+R25.
+(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
-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/
+Weld Component ID Description (System)'
-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)
+Pipe to Valve APR1-4301-8 Weld (813)
-Volumetric     (R-A.RU1) 2R20     12" I 1.125" eo* 12.00 I Long 50.0%     Yes     component configuration.
+APR1-4302-Elbow to Pipe 11-R8 Weld (813)
-(Figure 3-20)
+APR1-4302-Pipe to Elbow 12-R8 Weld (813)
-(PT)(UT)
+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)
-The examination was limited from Elbow to Pipe  IW8-2500-9, APR1-4302-                                                                                                               18" to 28" due to a whip restraint.
+IW8-2500-9, 10, & 11 R-A, R1.20 2R25 12"1 1.125" 45" 12.251 Shear 87.0%
-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)
+Yes Volumetric (UT)
-(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)
-The examination was limited from Pipe to Elbow  IW8-2500-9, APR1-4302-                                                                                                               18" to 28" due to a whip restraint.
+IW8-2500-9, 45" 12.251 Shear 10, & 11 R-A,RU1 2R23 6" I 0.719" eo* 12.0 I Long 50.0%
-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)
+Yes Volumetric eo* 12.25 I Shear (UT)
-(UT)
+Remarks The examination was limited due to component configuration.
-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)
+(Figure 3-20)
-Volumetric   (R-A, R1.11/16) 2R21 so*  12.00 I Long 50.0%     Yes     component configuration.
+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)
-(PT)(UT)
+The examination was limited due to component configuration.
-The examination was limited due to IW8-2500-8(c) component configuration.
+Additional inspections performed for MRP-146 during 2R21, 2A22, &
-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, &
 R24.
-(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/
+Weld Component 10 Description (System)'
-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#)
+Branch APA1-4307-Connection 21BC-AB Weld (E21)
-The examination was limited due to component configuration and IWB-2500-B(c)
+Branch APA1-4500-1 Connection to Pipe Weld (613)
-Branch                                                                                                 thickness changes. During outage IWB-2500*9,                                           30' /2 .25/ Shear APA1-4307-   Connection                                                                                                2A25, component ID APA1-4208*
+Flange to APA2-4511-10 Pipe Weld (E11)
-, & 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%
+Pipe to Valve APA2-4511-11 Weld (E11)
-Volumetric                                           60' I 2.251 Shear (E21)                                                                                                 coverage with similar degradation (UT) method as added assurance.
+Tee to Elbow APA2-4511-Weld 12 (E11)
-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.
+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/
-Pipe Weld     Volumetric     (A*A. A1 .11) 60' 12.00 I Shear                     (Figure 3-26)
+Requirements Item Number Outage Diameter/
-(613)       (PT)(UT)
+Frequency (MHz) I Actual VIII (Figure No.)
-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.
+credited Examined Thickness Coverage Qualified Mode Exam and Method (current R-A Item#)
-                     Volumetric     (A-A, A1 .11)
+IWB-2500-B(c)
-(E11)                                                           60' I 2.00 I Long                     (Figure 3-27)
+IWB-2500*9, 30' /2.25/ Shear 10, & 11 A-A, A1.11 2A25 4"1 0.719" 45' /2.251 Shear 22.3%
-(PT)(UT)
+Yes Volumetric 60' I 2.251 Shear (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 .
+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%
-                     Volumetric     (A-A, A1 .11)
+Yes Volumetric (A*A. A1.11) 60' 12.00 I Shear (PT)(UT)
-(E11)                                                           60' 12.00 I Long                      (Figure 3-28)
+IWC-2500-7 45' 12.251 Shear Surface and C-F-1, C5.11 2A20 10"/ 0.719" 60' 12.251 Shear 50.0%
-(PT)(UT)
+Yes Volumetric (A-A, A1.11) 60' I 2.00 I Long (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.
+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%
-                     Volumetric     (A*A, A1 .11)                          60' 12.00 I AL (E11)                                                                                                 (Figure 3-29)
+Yes Volumetric (A-A, A1.11) 60' 12.00 I Long (PT)(UT)
-(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*
+BC-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 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, &
+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]
-.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&#xa5;:811 directions- all elllUTI angles I00% for the circumference of the weld.
+Valve 3.63 square inches required for complete coverage for circumference of the weld.
-circumference of the weld.
+.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&#xa5;: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 Valve pipe TOTAL CODE REQUIRED COVERAGE= 0.4 SQUARE INCHES 100%
+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&deg; coverage, Green crosshatch- 45&deg; & 70&deg; coverage)
+(Red cross hatch-no coverage, Blue + Green cross hatch-60&deg; coverage, Green crosshatch-45&deg; & 70&deg; 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]
+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 Valve pipe TOTAL CODE REQUIRED COVERAGE= 0.4 SQUARE INCHES 100%
+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&deg; coverage, Green crosshatch- 45&deg; coverage)
+(Red cross hatch-no coverage, Blue+ Green cross hatch-60&deg; coverage, Green crosshatch-45&deg; 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]
+Dee:
-CODE COVERAGE PLOT FLOW WeldCl
+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
-Dee:               60 Dee:          ,
+~
-:                            * *..........~** _':J\wr&:t~ a~~ u~~ ~ili _
+WeldCl 60 Dee:
-Required Exam Area= .56 SQ INCHES+ 100%
+****.......... ~** _
-Coverage Obtained in the Axial Scans = .28 SQ. INCHES = 50%
+':J\\wr&:t~~a~~~u~~~~ili _
-Coverage Obtained in the Circ Scans = .28 SQ. INCHES = 50%
+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:&#xa3;.
-C. LAmP I~~ .'i3J ,43~
+C. LAmP I~~.'i3J,43~
-Examined 50% of Required Volume Height- .15", Width- 1.75" Length-0.5" Upstream Ax- 100% Downstream Ax 0%
+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%
+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]
+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 TEE
+FLOIJ
-                                   ~*             ~*
+~
+~
 Figure 3-7: 6" Pipe to Elbow [ALAl-4202-3-RB]
-CODE COVERAGE PWT (AREA LIMITATION ONLY) Not to Seale Pipe 1
+CODE COVERAGE PWT (AREA LIMITATION ONLY) Not to Seale
-Elbow I l
+~------------------~~~~
-           ~------------------~~~~
+l 21" Up San (Pipe) -ll" Req'd I 13.40" Esamlaed-64%
-" 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%
+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%
+CCW Sen -21" Req'd I 16.45" Eumlaed-78'Ye TOTAL CALCULATED COVERAGE *75%
-l&pestfor RelkfReqlllntl E2-14
+l&pestfor RelkfReqlllntl E2-14 J
-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]
+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 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%
+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&deg; coverage, Green crosshatch- 45&deg; coverage)
+(Red cross hatch-no coverage, Blue+ Green cross hatch-60&deg; coverage, Green crosshatch-45&deg; 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]
+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 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%
+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&deg; (Red cross hatch- no coverage, Blue + Green cross hatch- 60&deg; coverage, Green crosshatch- 45&deg; coverage)
+SCANNED ALL ACCESSABLE AREAS (ID PLATE) LOCATED NEAR 90&deg; (Red cross hatch-no coverage, Blue + Green cross hatch-60&deg; coverage, Green crosshatch-45&deg; 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]
+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 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%
+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&deg; coverage, Green crosshatch- 45&deg; coverage)
+(Red cross hatch-no coverage, Blue+ Green cross hatch-60&deg; coverage, Green crosshatch-45&deg; 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]
+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 Flange pipe TOTAL CODE COVERAGE= 0.06 SQ. INCHES TOTAL COVERAGE= 0.029 SQ INCHES=48%
+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&deg; and 45&deg; coverage)
+(Red cross hatch-no coverage, Green crosshatch-70&deg; and 45&deg; 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]
+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]
-/tluL.'.. .Sibfb tVrM -    5o'/. !'UrM C!.OVffZ.M&. CJB.'f71r.IN~
+/tluL.'...Sibfb tVrM -
-.1'1     J.t'i    1.oss     t.J'f I         I        I         I VALV&#xa3; f:LP&#xa3;
+o'/. !'UrM C!.OVffZ.M&.
-                                       -FLOW~
+CJB.'f71r.IN~
+.1'1 J.t'i 1.oss t.J'f I
+I I
+I f:LP&#xa3;
+-FLOW~
 Figure 3-13: 12" Pipe to Pipe [APR1-4102-2-RB]
-ELOW...._
+VALV&#xa3; ELOW...._
-L*
+L rElfEIH Of COVEP.A6E AlllAINfO 7.5'"'
-rElfEIH Of COVEP.A6E AlllAINfO  7.5'"'
+E2-19
-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]
+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 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%
+CODE COVERAGE PLOT FLOW WeldCL APRl-4104-30 I+-- 1.5"
-(Red cross hatch- no coverage, Blue + Green cross hatch- 60&deg; coverage, Green crosshatch- 45&deg; coverage)
+... 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&deg; coverage, Green crosshatch-45&deg; coverage)
 Figure 3-15: 3" Pipe to Valve [APR1-4106-8-RB]
-CODE COVERAGE PLOT FLOW WeldCL APRl-4106-8 Welded plate
+CODE COVERAGE PLOT Welded plate
-        .43r   pipe
+.43r pipe FLOW WeldCL APRl-4106-8 I+-- 1.05"
-                                                     .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%
+... 1
-(Red cross hatch- no coverage, Green crosshatch- 45&deg; and 70&deg; coverage)
+.146" TOTAL CODE REQUIRED COVERAGE= 0.15 SO INCHES TOTAL CODE COVERAGE ACHIEVED 0.0741 SO INCHES CODE COVERAGE ACHIEVED= 49.4%
-E2-20
+(Red cross hatch-no coverage, Green crosshatch-45&deg; and 70&deg; 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]
+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
+FLOW WELD#8 APR1-4100-8
-            ---,-1\
+\\
-Figure 3-16: 3" Pipe to Branch Connection [APR1-4106-11-RB]
+---,-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
+* 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]
+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]
-         --     _,. ...- ----_.;a..--        --
 r--~-~~-
-v .._ -----
+----_.;a..----v.._ -----
-                                                            /
 Figure 3-18: 3" Pipe to Valve [APR1-4108-12-RB]
-          - -- 'v' ---
 _...--~-L---
-I                   ,,
+I
-(           /*II          .   >
+(
-E2-22
+---'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 10-Year Interval 10 CFR 50.55a Request Number FNP-ISI-RR-03 Figure 3-21: 12" Elbow to Valve [APR1-4302-11-RB]
+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:*
+Wt-\\1f
+~E.S.T~AII\\IT LtMtTS St.At.l FR~M ll~"-lf:*
 FLoW h-
-                                                                      &#xa2;_
+&#xa2;_
-o.o*L::                                                                                                       PtPE.
+o.o*L::
-               \.o~
+\\.o~
-I. 111.         /.lOb       / .oCJ 16 l.~'il ElCAM1t-1A-r1fl~       1/o~un&#xa3;.:           HLI(,HT- 0.31"      uJ'bTH ~ '2..1.. LE~bTH* 'H)*
+l.~'il I. 111.
-ufSTAEA~          Ax - 15%                      Dow!'o\STIU.AM Ax- /5&deg;/Q UPSTP.EP.I\1     C..lll..C..- t5*to             D~~NST~f-A m C.l!l.c..- 75&deg;/o Figure 3-21a: 12" Elbow to Valve [APR1-4302-11-RB]
+/.lOb PtPE.
-Comments:    WI-IlP RE.STRAIIJT    LIMITS    WEl.CU II  ,..,1..
+/.oCJ 16 ElCAM1t-1A-r1fl~
-FP.o P'\ 18"- 'Z.i'''
+/o~un&#xa3;.:
+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&deg;/Q D~~NST~f-A m C.l!l.c..- 75&deg;/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'\\
+"- 'Z.i'''
 Slcoll:h or Phola:
-I         I I         I 1:::.====1      -   WELb II I
+E2-24 I
-I E2-24
+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
-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]
+.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 ...
+FLM.J...
-.101                                    f.t.o1.       I.ISK          1.1.51" E)(F\1""\tNC::..~      g1,5./o DF ({f.quta..&#xa3;~ \IOC....l.JI"'t!.
+f.t.o1.
-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&#xa3;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]
+I.ISK 1.1.51" E)(F\\1""\\tNC::..~
-Comll*lll:     Wl-ltP RE.STRf\1"--T         LtmiT~   WELDS 11..-1t.
+g1,5./o DF
-F""-o M    I&"- 2.i'''
+({f.quta..&#xa3;~ \\IOC....l.JI"'t!.
-Sl<etl:ll at Pllolo:
+ElC'A IY\\tNRnot.\\ VoL..ul't"'E. biMF-rlSu:.N5:
-                                                                                                   -lJELl:l II l'l
+~Erc,~r-o..,** wiDif.-'*1.1" u:.JGrr..t '"lo" Uf'5T~E.AI""\\ A 'I<- ls*/,.
-                                                                                          -   WELb 13 E2-26
+t~Dwf\\lC::Tit&#xa3;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 1 0 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]
+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 Valve pipe I+- 1.25"         .. ,
+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%
-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&deg; coverage, Green crosshatch- 45&deg; coverage)
+(Red cross hatch-no coverage, Blue+ Green cross hatch-60&deg; coverage, Green crosshatch-45&deg; coverage)
-Figure 3-24: 6" Valve to Pipe [APR1-4304-19-RB]
+Valve Figure 3-24: 6" Valve to Pipe [APR1-4304-19-RB]
-CODE COVERAGE PLOT FLOW WeldCL APR1-4304-19 Valve pipe
+CODE COVERAGE PLOT FLOW WeldCL APR1-4304-19 1+- 1.25"
-                                             .239" 1+- 1.25"         *I TOTAL CODE REQUIRED COVERAGE= 0.299 SQ INCHES CODE COVERAGE ACHIEVED= 0.149 SQ INCHES= 50%
+*I
-(Red cross hatch- no coverage, Blue+ Green cross hatch- 60&deg; coverage, Green crosshatch - 45&deg; coverage)
+.239" pipe TOTAL CODE REQUIRED COVERAGE= 0.299 SQ INCHES CODE COVERAGE ACHIEVED= 0.149 SQ INCHES= 50%
-E2-28
+(Red cross hatch-no coverage, Blue+ Green cross hatch-60&deg; coverage, Green crosshatch - 45&deg; 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]
+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'
+***** 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
-* 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
-* 7*3% FOR 180 DEGREES OFT HE PIPE
+***-- 60&deg; 45&deg; 30" CODE COVERAGE ACHIEVE 17.6% FOR 380 DEGREE AROUND THE PIPE 11.15"'
-                         ,. 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.
+TOTAL CODE VOLUME a 0.7525 SQ IN.
-" SCAN MAX OF 0.8 ON SUFACE OBTAININO CODE COVERAGE o27 SQ 1H TOTAL COOS COVERAGE ACHIEVE
+" 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 .
+* 17.9% FOR 110 DEORESS OF THE PIPE.
-              ***-- 60&deg;                                          600 ACHIEVED q% COVERAGE 45&deg; 17.9% + 36.1%
+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%
+* 27'\\ CODE COVERAGE ACHIEVED FOR THE AX SCANS 11.6% CODE COVERAGE ACHIEVEDCIRC SCANS TOTAL CODE COVERAGEACIUEVEO*l7Yo+ 17.6%/2 *22l%
-.15"'
+Figure 3-26: 14" Branch Connection to Pipe Weld [~PRl-4500-1]
-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)
-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]
+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:
+.77:J.". 775~. 7'fa**. 75]... 15J:
-                                         ~
+~
-FLANG&#xa3;                                                 PIP&#xa3; FLOW           ,....,.
+FLANG&#xa3; PIP&#xa3; FLOW
 .fO% R&#xa3;aUIR&#xa3;D EXAM VOLUME OBTAINED Figure 3-28: 10" Pipe to Valve [APR1-4511-11]
-            . 7J.7... 7J.l.... 130 .. GBJ''
+. 7J.7... 7J.l.... 130..
-II                I.l..O ..
+II GBJ''
-                                           <t.
+I.l..O..
-PIPE                                                       VALV&#xa3; FLOW SO!. REQUIRED EXAM VOLVM&#xa3; OBTAINf.D E2-30
+<t.
+PIPE VALV&#xa3; FLOW SO!. REQUIRED EXAM VOLVM&#xa3; 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]
+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]
-                                                   /. b21_"
+/. b 21_"
-PIPf so~\o   COVERAbE 08TAINED OF REQUIRED EXAM VDLVME E2-31}}
+so~\\o COVERAbE 08TAINED OF REQUIRED EXAM VDLVME E2-31}}

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

Latest revision as of 09:58, 5 January 2025

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Applicable Code Edition and Addenda

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