NL-15-0942, Proposed Lnservice Inspection Alternative FNP-ISI-ALT-18, Version 1.0

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Proposed Lnservice Inspection Alternative FNP-ISI-ALT-18, Version 1.0
ML15198A155
Person / Time
Site: Farley Southern Nuclear icon.png
Issue date: 07/17/2015
From: Pierce C
Southern Co, Southern Nuclear Operating Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NL-15-0942
Download: ML15198A155 (13)
v • d • e


Text

Charles R. Pierce Southern Nuclear Regulatory Affairs Director Operating Company, Inc.

40 Inverness Center Parkway A

Post Office Box 1295 Birmingham, AL 35242 Tel Fax 205.992.7872 205.992.7601 SOUTHERN COMPANY JUL 1 7 2015 Docket Nos.: 50-348 NL-15-0942 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant - Unit 1 Proposed lnservice Inspection Alternative FNP-ISI-ALT-18, Version 1.0 Ladies and Gentlemen:

In accordance with 10CFR50.55a(z)(1 }, Southern Nuclear Operating Company (SNC) hereby requests Nuclear Regulatory Commission (NRC) approval of proposed inservice inspection (lSI) alternative FNP-ISI-ALT-18, Version 1.0. This alternative requests a one-time extension of the time between examinations required by IW8-2412, Inspection Program 8, of Category 8-A and 8-D welds from 10 years to 20 years.

This letter contains no NRC commitments. If you have any questions, please contact Ken McElroy at (205) 992-7369.

Sincerely, t.!l. ~

C.R. Pierce Regulatory Affairs Director CRP/JMC/Iac

Enclosure:

Proposed Alternative FNP-ISI-ALT-18, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

U. S. Nuclear Regulatory Commission NL-15-0942 Page 2 cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President- Farley Mr. M.D. Meier, Vice President- Regulatory Affairs Mr. D. R. Madison, Vice President- Fleet Operations Mr. B. J. Adams, Vice President- Engineering Ms. B. L. Taylor, Regulatory Affairs Manager - Farley RTYPE: CFA04.054 U.S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. S. A. Williams, NRR Project Manager- Farley Mr. P. K. Niebaum, Senior Resident Inspector- Farley

Joseph M. Farley Nuclear Plant- Unit 1 Proposed lnservice Inspection Alternative FNP-ISI-ALT-18, Version 1.0 Enclosure Proposed Alternative FNP-ISI-ALT-18, Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Enclosure to N L-15-0942 Proposed Alternative FNP-ISI-ALT-18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Plant Site-Unit: Joseph M. Farley Nuclear Plant (FNP)- Unit 1.

4th lnservice Inspection (lSI) Interval- December 1, 2007 through November 30, Interval Dates:

2017.

Requested Date for Approval is requested by July 31, 2016.

Approval:

The affected components are Examination Category 8 -A, Items 81 .11 , 8 1.12, ASME Code 81.21 , 81.22 and 81.30 reactor vessel (RV) shell welds, and Examination Category Components 8-D , Items 83.90 and 83.100 RV nozzle welds and nozzle inside radius section .

Affected:

The specific components are provided in Table 4.

Applicable The applicable code Edition and Addenda (for the 4th lSI interval) is ASME Section Code Edition XI, "Rules for lnservice Inspection of Nuclear Power Plant Components," 2001 and Addenda: Edition through the 2003 Addenda (Reference 1).

IW8-2412, Inspection Program 8, requires volumetric examination of essentially Applicable 100% of reactor vessel pressure-retaining welds identified in Table IW8-2500-1 Code once each 10-year interval. The FNP - Unit 1 4th 10-year inservice inspection (lSI)

Requirements:

interval is scheduled to end on November 30, 2017.

An alternative is requested from the requirement of IW8-2412, Inspection Program 8, that volumetric examination of essentially 100% of reactor vessel pressure-retaining Examination Category 8-A and 8-D welds be performed once each 10-year interval. Extension of the interval between examinations of Category 8-A and 8-D welds from 10 years to up to 20 years will result in a reduction in man-rem exposure and examination costs.

The Westinghouse pilot plant RV analysis defined in WCAP-16168-NP-A, Revision Reason for 3, utilizes probabilistic fracture mechanics and risk analysis methods to justify Request: extending the lSI interval for reactor vessel welds (Examination Category 8-A),

nozzle-to-vessel welds and nozzle inside radius section (Examination Category 8-D) from 10 years to 20 years.

An analysis has been performed showing that FNP - Unit 1, which is a Westinghouse 3-Loop plant, is bounded by the pilot plant parameters defined in WCAP-16168-NP-A, Revision 3. Therefore, Southern Nuclear Operating Company (SNC) is requesting approval of this alternative to allow the use of the lSI interval extension for the affected FNP - Unit 1 components.

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Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT -18 Version 1.0, in Accordance with 10 CFR 50.55a{z)(1)

SNC is requesting a one-time extension of the lSI interval from 10 years to 20 years for FNP - Unit 1 Examination Category 8-A welds and Examination Category 8-D nozzle-to-vessel welds and nozzle inside radius section.

Proposed Specifically, this proposed alternative would permit the deferral of the ASME Code Alternative: required Examination Category 8-A and 8-D volumetric examinations currently scheduled for the Fall of 2016 (3rd period of 4th interval) until no later than the end of November 2027 (3rd period of 5th interval). The proposed inspection date for FNP - Unit 1 is within one outage of the schedule presented in the latest implementation plan, OG-1 0-238 (Reference 2).

In accordance with 10 CFR 50.55a(z)(1 ), an alternate inspection interval is requested on the basis that the current interval can be revised with negligible change in risk by satisfying the risk criteria specified in Regulatory Guide 1.174 (Reference 3).

The methodology used to demonstrate the acceptability of extending the inspection intervals for Examination Category 8-A and 8-D components is contained in WCAP-16168-NP-A, Revision 3 (Reference 4). This methodology was used to develop a pilot plant risk analysis for Westinghouse (W), Combustion Engineering (CE), and Babcock and Wilcox (B&W) RV designs and is an extension of the work that was performed as part of the Nuclear Regulatory Commission (NRC) Pressurized Thermal Shock {PTS) Risk Re-Evaluation (Reference 5). The WCAP used the estimated through wall cracking frequency (TWCF) as a measure of the risk of RV failure, and it was demonstrated that the inspection interval for the affected components can be extended from 10 years to 20 years while meeting the change in risk guidelines found in Regulatory Guide 1.174 (Reference 3).

Basis for Use:

Reference 4 was subsequently approved by the NRC in a July 26, 2011 revised safety evaluation. Section 3.4 of the safety evaluation provides the requirements for a utility to submit an alternative in accordance with 10 CFR 50.55a(z)(1) to use the WCAP for a plant specific evaluation. These requirements are addressed below:

1. Licensees must demonstrate that the embrittlement of their RV is within the envelope used in the supporting analyses. A plant specific analysis was performed that demonstrated that FNP - Unit 1 RV parameters are bounded by corresponding pilot plant parameters. The critical parameters are identified in Table 1. Table 3 provides detailed information relative to the calculation of the TWCF.
2. Licensees must report whether the frequency of the limiting design basis transients during prior operation are less than the frequency identified in the PWROG (Reference 4) fatigue analysis. As shown in Table 1, the frequency of the FNP - Unit 1 limiting design basis transients are bounded by the frequency identified in the PWROG (Reference 4) fatigue analysis.
3. Licensees must report the results of prior lSI of RV welds and the proposed schedule for the next 20 year lSI interval. The results of the _Qrevious RV E-2

Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT -18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) inspections for FNP- Unit 1 are provided in Table 2. This information confirms that satisfactory examinations have been performed on the FNP-Unit 1 RV.

4. In the request for an alternative, each licensee shall identify the years in which the future inspections will be performed. The FNP - Unit 1 RV examinations currently scheduled for 2016 will be deferred until no later than the end of 2027. The dates provided must be within plus or minus one refueling cycle of the date identified in PWROG letter OG-10-238 , dated July 12, 2010 (Reference 2).

The intent of the schedule identified in PWROG letter OG-1 0-238 is to provide for a sampling of vessel weld inspections in the PWR fleet over the 20 year interval such that any emerging degradation mechanisms are detected in a timely manner. The dates that are proposed for FNP - Unit 1 in this request for the alternative are consistent with the dates identified in Basis for Use PWROG letter OG-1 0-238. These dates will result in one fewer examination (Cont.): being performed in 2016. This change in dates will still provide for at least one inspection each year and will have a negligible impact on the ability of the schedule to provide for early detection of emerging degradation mechanisms.

FNP - Unit 1 is bounded by the pilot plant application because the total TWCF for FNP - Unit 1 was calculated as 4.81 E-11; therefore, the use of this proposed alternative will provide an acceptable level of quality and safety. Therefore, it is requested that the NRC authorize this proposed alternative in accordance with 10 CFR 50.55a(z)(1 ).

The reactor vessel neutron fluence values used for FNP- Unit 1 are based on the pressurized thermal shock evaluation in WCAP-17506-NP (Reference 9). The fluence values in Reference 9 are based on more recent fluence evaluations than the values reported in the Farley License Renewal Application (Reference 10). The updated fluence values are also summarized in the Farley request to revise Technical Specifications associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report (Reference 11 ),

which was accepted by the NRC in Reference 12.

Duration of The fourth 10 Year lSI Interval is scheduled to end in November of 2017. Granting Proposed approval of this proposed alternative will allow an extension of the fourth interval for Alternative: these exams to be performed no later than the end of November 2027.

  • "Joseph M. Farley Nuclear Plant, Unit 2 (Farley Unit 2) - Relief Request for Extension of the Reactor Vessellnse rvice Inspection Date to the Year 2020 (Plus or Minus One Outage) (TAC No. ME3010)," dated July 12, 2010 (ADAMS Accession Number ML101750402).

Precedents:

  • "Surry Power Station Units 1 and 2 - Relief Implementing Extended Reactor Vessel Inspection Interval (TAC Nos. ME8573 and ME8574)," dated April 30, E-3

Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT-18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1) 2013 (ADAMS Accession Number ML13106A140).

  • "Vogtle Electric Generating Plant, Units 1 and 2- Request for Alternatives VEGP-ISI-ALT-05 and VEGP-ISI-ALT-06 (TAG Nos. MF2596 and MF2597),"

dated March 20,2014 (ADAMS Accession Number ML14030A570) .

  • "Catawba Nuclear Station Units 1 and 2: Proposed Relief Request 13-CN-003, Request for Alternative to the Requirement of IWB-2500, Table IWB-2500-1, Category B-A and Category B-D for Reactor Pressure Vessel Welds (TAG Nos.

MF1922 and MF1923)," dated March 26, 2014 (ADAMS Accession Number ML14079A546).

  • "Sequoyah Nuclear Plant, Units 1 and 2- Requests for Alternatives 13-ISI-1 and 13-ISI-2 to Extend the Reactor Vessel Weld lnservice Inspection Interval (TAG Precedents Nos. MF2900 and MF2901)," dated August 1, 2014 (ADAMS Accession Number (Cont): ML14188B920).
  • "Byron Station, Unit No. 1 - Relief from Requirements of the ASME Code to Extend the Reactor Vessellnserv ice Inspection Interval (TAG No. MF3596),"

dated December 10, 2014 (ADAMS Accession Number ML14303A506).

  • 'Wolf Creek Generating Station - Request for Relief Nos. 13R-08 and I3R-09 for the Third 10-Year lnservice Inspection Program Interval (TAG Nos. MF3321 and MF3322)," dated December 10, 2014 (ADAMS Accession Number ML14321A864).
  • "Callaway Plant, Unit 1 - Request for Relief 13R-17, Alternative to ASME Code Requirements Which Extends the Reactor Vessel Inspection Interval from 10 to 20 Years (TAG No. MF3876)," dated February 10,2015 (ADAMS Accession Number ML15035A148).
1. ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with the 1989 Addenda up to and including the 2001 Edition with the 2003 Addenda, American Society of Mechanical Engineers, New York.
2. PWROG Letter OG-1 0-238, "Revision to the Revised Plan for Plant Specific Implementation of Extended lnservice Inspection Interval perWCAP-16 168-NP, Revision 1, "Risk-Informed Extension of the Reactor Vessel In-Service Inspection Interval." PA-MSC-0120," July 12, 2010 (ADAMS Accession Number ML11153A033).

References:

3. NRC Regulatory Guide 1.174, Revision 1, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," November 2002.
4. Westinghouse Report WCAP-16168-NP-A, Revision 3, "Risk-Informed Extension of Reactor Vessel In-Service Inspection Interval," October 2011 (ADAMS Accession Number ML113060207).
5. NUREG-1874, "Recommended Screening Limits for Pressurized Thermal Shock (PTS)," U. S. Nuclear Regulatory Commission, March, 2010.

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Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT -18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

6. NRC Letter Report, "Generalization of Plant-Specific Pressurized Thermal Shock (PTS) Risk Results to Additional Plants," U.S. Nuclear Regulatory Commission ,

December 14, 2004 (ADAMS Accession Number ML042880482) .

7. NRC Regulatory Guide 1.99, Revision 2, "Radiation Embrittlement of Reactor Vessel Materials," U. S. Nuclear Regulatory Commission, May 1988.
8. CE Report MISC-PENG-ER-011, Revision 0, The Reactor Vessel Group Record Evaluation Program Phase II Final Report for the Farley 1 Reactor Pressure Vessel Plates, Forgings, Welds and Cladding," October 1995.
9. Westinghouse Report WCAP-17506-NP , Revision 0, "Farley Units 1 and 2 Pressurized Thermal Shock Evaluations," December 2011 .
10. "Joseph M. Farley Nuclear Plant Application for License Renewal ," Facility Operating License No. NPF-2.
11. Southern Company Letter NL-12-0868, "Joseph M. Farley Nuclear Plant - Units 1 & 2 Request to Revise Technical Specifications Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report," Docket Nos. 50-348 and 50-364, dated August 15, 2012 (ADAMS Accession Number ML12229A521).

References (Cont): 12. NRC Safety Evaluation Report, "Joseph M. Farley Nuclear Plant, Units 1 and 2, Issuance of Amendments Regarding Technical Specifications Revisions Associated with the Low Temperature Overpressure Protection System and the Pressure and Temperature Limits Report (TAG Nos. ME9256 and ME9257) (NL-12-0868)," dated October 2, 2013 (ADAMS Accession Number ML13249A386).

13. Code of Federal Regulations, 10 CFR Part 50.61 a, "Alternate Fracture Toughness Requirements for Protection against Pressurized Thermal Shock Events," U.S. Nuclear Regulatory Commission, Washington D. C., Federal Register, Volume 75, No. 1, dated January 4, 2010 and No. 22 with corrections to part (g) dated February 3, 2010, March 8, 2010, and November 26, 2010.
14. Constellation Energy Letter, Attachment 1, "Calvert Cliffs Nuclear Power Plant Unit No. 2; Docket No. 50-318 Revised Request to Extend the lnservice Inspection Interval for Reactor Vessel Weld Examination s- Relief Requests (lSI-020 and ISI-021 ),"dated October 1, 2008. (ADAMS Accession Numbers ML082760282 and ML082760283)
15. NRC Letter, "Safety Evaluation for Relief Requests ISI-020 and 021 Reactor Vessel Weld Examination Extension- Calvert Cliffs Nuclear Power Plant, Unit No.2 (TAG Nos. MD9773 and MD9774)," dated AprilS, 2009. (ADAMS Accession Number ML090920077)

Status: Awaiting NRC approval.

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Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT -18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Table 1 Critical Parameters for the Application of the Bounding Analysis as Applied to FNP - Unit 1 Additional Evaluation Parameter Pilot Plant Basis FNP- Unit 1 Basis Required?

Dominant Pressurized Thermal Shock NRC PTS Risk PTS Generalization No (PTS) Transients in the NRC PTS Risk Study (Reference 5) Study (Reference 6)

Study are applicable Through Wall Cracking Frequency 1.76E-08 Events per 4.81 E-11 Events per No (TWCF) year (Reference 4) year (Calculated using Reference 4)

Frequency and Severity of Design Basis 7 heatup/cooldowns Bounded by 7 No Transients per year (Reference heatuR/cooldowns per

4) year <l Cladding Layers (Single/Multiple) Single Layer Single Layer No (Reference 4)

(1) Per the J. M. Farley Apphcatron for Lrcense Renewal (Reference 10), after 60 years of operatron, the projected number of design basis transients is below the number specified in the 40-year design bases. As a result, FNP- Unit 1 is conservatively bounded by 7 heatup/cooldown events per year.

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Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT -18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Table 2 Additional Information Pertaining to Reactor Vessel Inspections for FNP - Unit 1 Inspection The latest lSI was conducted in accordance with the ASME Code,Section XI methodology: 1989 Edition, with no Addenda. Examinations of Category 8-A and B-0 welds were performed to ASME Section XI Appendix VIII, 2001 Edition with the 2003 Addenda, as modified by 10 CFR 50.55a(b)(2)(xiv, xv and xvi) . Future inservice inspections will be performed to ASME Section XI Appendix VIII requirements.

Number of past Three 10-Year inservice inspections have been performed.

inspections:

Number of There were three indications identified in the beltline region during the most indications found: recent inservice inspection. These subsurface indications are located in the lower shell plate (Item 5 in Table 3). All indications are acceptable per Table IWB-3510-1 of Section XI of the ASME Code. All three indications are within the inner 1/1 Oth or 1" of the reactor vessel thickness. Two of these indications are acceptable per the requirements of the Alternate PTS Rule, 10 CFR 50.61 a (Reference 13), since the number of flaws is less than the allowable number of flaws for each flaw size increment. One indication, in plate material with a through-wall extent of 0.38", does not meet the requirements in the Alternate PTS Rule, 10 CFR 50.61a (Reference 13).

While one flaw is outside the limits in 10 CFR 50.61 a, it is not expected that this flaw would increase the FNP - Unit 1 TWCF value above that of the pilot plant for the following reasons:

  • The plate which the flaw is located has a maximum RT NDT + !::..T 3o of 145.50°F and is not the limiting material in the beltline region.

Furthermore, the peak fluence was used to calculate the t::..T30 shift for this plate; however, the fluence at the position of this flaw is approximately 30% of the peak fluence. This was determined using the fluence values reported in Table 5.1-1 of WCAP-17506-NP (Reference 9) since the location of the flaw is just slightly offset from the location of the lower shell longitudinal welds. Therefore, the actual RTNDT + !::.. T30 at the specific location of this flaw is less than 145.50°F.

  • The total number of flaws detected in the FNP - Unit 1 beltline is far less than those allowed in the Alternate PTS Rule, 10 CFR 50.61 a.
  • The TWCF for the FNP - Unit 1 reactor vessel is more than 2 orders of magnitude below that for the bounding pilot plant reactor vessel in WCAP-16168-NP-A, Revision 3 (Reference 4). Furthermore, the TWCF for FNP- Unit 1 was conservatively determined for 54 EFPY, corresponding to the end of license. This is conservative because FNP-Unit 1 is conservatively projected to have operated to approximately 42.5 EFPY in 2027, which is the proposed date for the next inspection.

No indications are located within the weld or forging material of the reactor vessel beltline. The following indications are located within the plate material of the reactor vessel beltline.

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Enclosure to NL-15-0942 Proposed Alternative FNP-ISI-ALT-18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Table 2 Additional Information Pertaining to Reactor Vessel Inspections for FNP- Unit 1 Through-Wall Extent, TWE (in.) Scaled maximum Number of plate flaws number of plate flaws (Axiai/Circ.)

TWEMIN TWEMAX 0.075 0.375 74 2 (210) 0 .125 0.375 29 2 {2/0) 0.175 0.375 8 0 0.225 0.375 3 0 0.275 0.375 1 0 0.325 0.375 1 0 0.375 Infinite 0 1 (1 /0)

Note that a reactor vessel beltline flaw with a through-wall extent of 0.60 inch was identified at Calvert Cliffs Unit 2 during the last 10-year inservice inspection.

This flaw also exceeded the acceptance criteria, and was included in the Calvert Cliffs Unit 2 relief request (Reference 14) to extend the reactor vessel inservice inspection interval from 10 to 20 years for the 8-A and 8-D welds. The flaw in the Calvert Cliffs Unit 2 reactor vessel, which was also larger than the flaw found in the Farley Unit 1 reactor vessel , was determined to be acceptable by the NRC in Reference 15.

Proposed The fourth inservice inspection originally scheduled for 2016 will be performed inspection no later than the end of November 2027. These RPV examinations will be schedule for performed to the ASME Code in effect for the ten-year lSI interval they are balance of plant performed in, which if this alternative is approved would be the fourth inservice life: inspection interval. The proposed inspection date is consistent with the latest revised implementation plan, OG-1 0-238 (Reference 2).

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Enclosure to NL-15-0942 Proposed Alternative FNP-181-ALT-18 Version 1.0, in Accordance with 10 CFR 50.55a(z)(1)

Table 3 Details of the Through Wall Cracking Frequency Calculation for FNP - Unit 1 @ 54 EFPY Inputs Inter. & Lower Shell T wall [inches]: 8.031 Reactor Coolant System Temperature, T c[°F]: N/A Nozzle Shell T wall [inches]: 9.156 Un- Flue nee Region/Component Description Material /Flux Material Heat Cu Ni R.G. [1019 No. CF [0 F] Irradiated (Ref. 9) Type (Ref. 8) No. (Ref. 8)

(wt%] [wt%] 1.99 Neutron/cm2 (Ref. 9) (Ref. 9) (Ref. 9) RTNDT(u) ( 0 F]

Pos. E>1 MeV]

(Ref. 9)

(Ref. 9) 1 Upper Shell Forging 86914 A508 Cl. 2 123W209VA1 0.16 0.684 1.1 120.1 30 1.02 2 Intermediate Shell Plate 86903-2 A533 Gr. 8 Cl. 1 C6294-1 0.13 0.60 1.1 91.0 0 5.93 3 Intermediate Shell Plate 86903-3 A533 Gr. 8 Cl. 1 C6308-2 0.12 0.56 1.1 82.2 10 5.93 4 Lower Shell Plate 86919-1 A533 Gr. 8 Cl. 1 C6940-1 0.14 0.55 2.1 106.7 15 5.81 5 Lower Shell Plate 86919-2 A533 Gr. 8 Cl. 1 C6897-2 0.14 0.56 1.1 98.2 5 5.81 6 Inter. Shell Long. Weld 19-894 A Linde 1092 33A277 0.258 0.165 1.1 126.3 -56 1.83 7 Inter. Shell Long. Weld 19-894 8 Linde 1092 33A277 0.258 0.165 1.1 126.3 -56 1.83 8 Lower Shell Long. Weld 20-894 A Linde 0091 90099 0.197 0.06 1.1 91.4 -56 1.79 9 Lower Shell Long. Weld 20-894 8 Linde 0091 90099 0.197 0.06 1.1 91.4 -56 1.79 10 Upper to Inter. Shell Circ. Weld 10-894 Linde 0091 90099 0.197 0.06 1.1 91.4 -56 1.02 11 Inter. to Lower Shell Circ. Weld 11-894 Linde 0091 6329637 0.205 0.105 1.1 98.4 -56 5.81 Outputs Methodology Used to Calculate 11T 30 : Regulatory Guide 1.99, Revision 2 (Reference 7)

Controlling 19 Material Region Fluence [10 FF (Fiuence No. (From RT MAX*XX [ 0 R] Neutron/cm 2 , /1T3o [°F] TWCF95-XX E > 1.0 MeV] Factor)

Above)

Limiting Axial Weld - AW 4 598.43 1.79 1.160 138.76 O.OOE+OO Limiting Plate - PL 4 627.33 5.81 1.431 167.66 1.51E-11 Forging -FO 1 610.44 1.02 1.006 150.77 4.28E-12 Circumferential Weld- CW 4 627.33 5.81 1.431 167.66 2.37E-19 TWCF95*TOTAL(aAwTWCF9s-Aw + OpLTWCF95*PL + OpLTWCF9s-Fo + acw TWCF9s-cw): 4.81 E-11 E-9

Enclosure to N L-15-0942 Proposed Alternative FNP-ISI-ALT-18 Version 1.0 in Accordance with 10 CFR 50.55a(z)(1)

Table 4 List of Affected Components for FNP- Unit 1 ASME ASME Category Component ID Description Item Number 8-A 81.11 ALA1-1100-2 UPPER TO MIDDLE SHELL 8-A 81 .11 ALA1-1100-5 MIDDLE TO LOWER SHELL 8-A 81 .11 ALA 1-1100-8 LOWER SHELUBOTT OM HEAD 8-A 81.12 ALA1-1100-3 MIDDLE SHELL LONG. SEAM 8-A 81.12 ALA1-1100-4 MIDDLE SHELL LONG. SEAM 8-A 81.12 ALA1-1100-6 LOWER SHELL LONG. SEAM 8-A 81.12 ALA 1-1100-7 LOWER SHELL LONG. SEAM 8-A 81 .21 ALA1 -1100-15 LOWER HEAD/MERID IONAL CIR.

8-A 81.21 ALA1 -1100-16 BOTTOM HD.RING/LOWER HD. SHELL 8-A 81.22 ALA 1-11 00-1 0 LOWER HEAD MERIDIONAL SEAM 8-A 81.22 ALA 1-11 00-11 LOWER HEAD MERIDIONAL SEAM 8-A 81 .22 ALA 1-11 00-12 LOWER HEAD MERIDIONAL SEAM 8-A 81.22 ALA 1-11 00-13 LOWER HEAD MERIDIONAL SEAM 8-A 81.22 ALA1-1100-14 LOWER HEAD MERIDIONAL SEAM 8-A 81.22 ALA1-1100-9 LOWER HEAD MERIDIONAL SEAM 8-A 81.30 ALA1-1100-1 FLANGE TO UPPER SHELL 8 -D 83.100 ALA1-1100-171R NOZZLE-INSIDE RADIUS (OUTLET) 8-D 83.100 ALA 1-11 00-181R NOZZLE-INSIDE RADIUS (INLET) 8-D 83.100 ALA1-1100-191R NOZZLE-INSIDE RADIUS (OUTLET) 8-D 83.100 ALA 1-11 00-201R NOZZLE-INSIDE RADIUS (INLET) 8-D 83.100 ALA 1-11 00-211 R NOZZLE-INSIDE RADIUS (OUTLET) 8-D 83.100 ALA1 -1100-221R NOZZLE-INSIDE RADIUS (INLET) 8-D 83.90 ALA 1-11 00-17 NOZZLE/VESSEL WELD (OUTLET) 8-D 83.90 ALA1-1100-18 NOZZLE/VESSEL WELD (INLET) 8-D 83.90 ALA1 -1100-19 NOZZLE/VESSEL WELD (OUTLET) 8-D 83.90 ALA 1-11 00-20 NOZZLE/VESSEL WELD (INLET) 8-D 83.90 ALA 1-1100-21 NOZZLE/VESSEL WELD .(OUTLET) 8-D 83.90 ALA1-1100-22 NOZZLE/VESSEL WELD (INLET)

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