WBL-23-018, Revised Pressure and Temperature Limits Report (PTLR)

Revised Pressure and Temperature Limits Report (PTLR)

ML23100A044
Person / Time
Site:	Watts Bar
Issue date:	04/10/2023
From:	Anthony Williams Tennessee Valley Authority
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
WBL-23-018
Download: ML23100A044 (1)

Text

TV4 TENNESSEE VALLEY AUTHORITY Post Office Box 2000, Spring City, Tennessee 37381 WBL-23-018 April 10, 2023 10 CFR 50.36 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-391

Subject:

Watts Bar Nuclear Plant Unit 2 - Revised Pressure and Temperature Limits Report (PTLR)

The purpose of this letter is to provide the enclosed copy of the Watts Bar Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 7, effective March 29, 2023, in accordance with Technical Specification Section 5.9.6.c.

There are no new regulatory commitments in this letter. Should you have questions regarding this submittal, please contact Jonathan Johnson, Site Licensing Manager, at jtjohnsonO@tva.gov.

Site Vice President Watts Bar Nuclear Plant

U.S. Nuclear Regulatory Commission WBL-23-018 Page 2 April 10, 2023 Enclosure Watts Bar Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR),

Revision 7.

cc: (Enclosure)

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Watts Bar Nuclear Plant NRC Project Manager - Watts Bar Nuclear Plant

Enclosure Watts Bar Nuclear Plant, Unit 2 Pressure and Temperature Limits Report (PTLR),

Revision 7 WBL-23-018 E1 of 22

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 236 of 262 Appendix B (Page 1 of 21)

Watts Bar Unit 2 - RCS Pressure and Temperature Limits Report (PTLR) - Revision 7 APPENDIX B TO RCS SYSTEM DESCRIPTION N3-68-4001 WATTS BAR UNIT 2 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

REVISION 7 Prepared by: J. W. Thompson Checked by: J. F. Fitzsimmons Approved by: C. S. Kerlin

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 237 of 262 Appendix B (Page 2 of 21) 1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This PTLR for Watts Bar Unit 2 has been prepared in accordance with the requirements of Technical Specification 5.9.6. Revisions to the PTLR shall be provided to the NRC within 30 days of issuance.

The Technical Specifications affected by this report are listed below:

LCO 3.4.3, RCS Pressure and Temperature (P/T) Limits LCO 3.4.12, Cold Overpressure Mitigation System (COMS) 2.0 RCS PRESSURE AND TEMPERATURE LIMITS The limits for LCO 3.4.3 are presented in the subsection which follows. These limits have been developed (Ref. 1) using the NRC-approved methodologies specified in Technical Specification 5.9.6.

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.1.1 The minimum boltup temperature is 60!!F.

2.1.2 The RCS temperature rate-of-change limits are:

A. A maximum heatup rate of 100!F per hour.

B. A maximum cooldown rate of 100!F per hour.

C. A maximum temperature change of 10!F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.3 RCS P/T Limits for Heatup, Cooldown, Inservice Hydrostatic and Leak Testing, and Criticality The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 and 2.1-2 (Ref. 1).

3.0 COLD OVERPRESSURE MITIGATION SYSTEM (LCO 3.4.12)

The lift setting limits for the pressurizer Power Operated Relief Valves (PORVs) are presented in the subsection that follows. These lift setting limits have been developed using the NRC-approved methodologies specified in Technical Specification 5.9.6.

3.1 Pressurizer PORV Lift Setting Limits The pressurizer PORV lift setting limits are specified by Figure 3.1-1 and Table 3.1-1 (Ref. 2).

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 238 of 262 Appendix B (Page 3 of 21)

NOTE: These setpoints include allowance for pressure difference between the pressure transmitter and reactor midplane, and also includes a 71.8 psig pressure channel uncertainty, and a 16.3!F temperature uncertainty.

3.2 Arming Temperature COMS shall be armed when any RCS cold leg temperature is 225!F for Unit 2.

4.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The results of these examinations shall be used to update Figures 2.1-1, 2.1-2, and 3.1-1.

The pressure vessel steel surveillance program (Ref. 3) is in compliance with Appendix H to 10 CFR 50 (Ref. 4), entitled Reactor Vessel Material Surveillance Program Requirements.

The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASTM E208 (Ref. 5). The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, Fracture Toughness Criteria for Protection Against Failure, to Section XI of the ASME Boiler and Pressure Vessel Code (Ref. 6). The surveillance capsule removal schedule meets the requirements of ASTM E185-82 (Ref. 7).

The removal schedule is provided in Table 4.0-1.

To date, one surveillance capsule has been removed from Watts Bar Unit 2, Capsule U. The testing results from Capsule U are documented in WCAP-18518-NP (Ref. 10). Regulatory Guide (RG) 1.99, Revision 2, requires at least two credible data sets for use in embrittlement calculations. Since Capsule U is the first capsule to be withdrawn from Watts Bar Unit 2, this criterion is not satisfied for the surveillance forging. However, there is additional surveillance data available from sister plants for the surveillance weld. The Watts Bar Unit 1, Catawba Unit 1, and McGuire Unit 2 surveillance programs contain weld wire Heat # 895075, which was also used in the fabrication of Watts Bar Unit 2 intermediate to lower shell circumferential Weld Seam W05. Thus, the data from these surveillance programs are applicable to Watts Bar Unit 2, and the RG 1.99, Position 2.1 chemistry factor may be used to make embrittlement projections. Table 5-4 contains the sister-plant weld material surveillance data as well as the data from Watts Bar Unit 2, Capsule U, surveillance weld.

The effect of these results on Figures 2.1-1, 2.1-2, and 3.1-1 were evaluated in WCAP-18532-NP (Ref. 11) and LTR-SCS-20-53 (Ref 12).

5.0 SUPPLEMENTAL DATA TABLES

( Table 5-1 contains a Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials.

( Table 5-2 shows a Summary of the Initial RTNDT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange.

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 239 of 262 Appendix B (Page 4 of 21)

( Table 5-3 provides the Summary of the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Material Position 1.1 Chemistry Factors.

( Table 5-4 provides the Watts Bar Unit 2, Catawba Unit 1, Watts Bar Unit 1, and McGuire Unit 2 Surveillance Weld Data for Heat #895075

( Table 5-5 shows the Calculation of the Watts Bar Unit 2 Heat #895075 Position 2.1 Chemistry Factor Using Surveillance Capsule Data.

( Table 5-6 provides Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials.

( Table 5-7 shows Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location.

( Table 5-8 contains Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 3/4T Location.

( Table 5-9 provides a Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves.

( Table 5.10 shows RTPTS calculations for the Watts Bar Unit 2 Beltline and Extended Beltline Materials at 32 EFPY.

6.0 REFERENCES

1. WCAP-18191-NP, Revision 1, Watts Bar Unit 2 Heatup and Cooldown Limit Curves for Normal Operation and Supplemental Reactor Vessel Integrity Evaluations, February 2020.

2. Westinghouse Letter LTR-SCS-17-34, Revision 0, Watts Bar Unit 2 Cold Overpressure Mitigation System (COMS) Setpoint Analysis due to TPBARs dated August 14, 2017.

3. WCAP-9455, Revision 4, Tennessee Valley Authority Watts Bar Unit No. 2 Reactor Vessel Radiation Surveillance Program, August 2019.

4. Code of Federal Regulations, 10 CFR 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, U.S. Nuclear Regulatory Commission, Federal Register, Volume 60, No. 243, December 19, 1995.

5. ASTM E208, Standard Test Method for Conducting Drop-Weight Test to Determine Nil-Ductility Transition Temperature of Ferritic Steels, American Society for Testing and Materials.

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 240 of 262 Appendix B (Page 5 of 21)

6. Appendix G to the 1998 through the 2000 Addenda Edition of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Fracture Toughness Criteria for Protection Against Failure.

7. ASTM E185-82, Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF), ASTM 1982.

8. WCAP-13830, Revision 1, Heat Up and Cool Down Limit Curves for Normal Operation for Watts Bar Unit 2, J. M. Chicots, et al, February 1995.

9 NRC Regulatory Issue Summary (RIS) 2014-11, Information on Licensing Applications for Fracture Toughness Requirements for Ferritic Reactor Coolant Pressure Boundary Components, U.S. Nuclear Regulatory Commission, October 2014. [Agencywide Documents Access and Management System (ADAMS)

Accession Number ML14149A165]

10 WCAP-18518-NP, Revision 0, Analysis of Capsule U from the Watts Bar Unit 2 Reactor Vessel Radiation Surveillance Program, March 2020 11 WCAP-18532-NP, Revision 0, Watts Bar Unit 2 Reactor Vessel Integrity Evaluations, August 2020 12 Westinghouse Letter LTR-SCS-20-53, Revision 0, Watts Bar Unit 2 Cold Overpressure Mitigation System (COMS) Evaluation for Capsule U, dated September 28, 2020.

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 241 of 262 Appendix B (Page 6 of 21) 7.0 FIGURES AND TABLES MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 using Reg. Guide 1.99 Position 1.1 LIMITING ART VALUES AT 32 EFPY: 1/4T, 88!F (Axial Flaw) 3/4T, 71!F (Axial Flaw)

Figure 2.1-1 Watts Bar Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100!F/hr)

Applicable for 32 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

(Plotted data (Ref. 1) provided in Table 2.1-1)

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 242 of 262 Appendix B (Page 7 of 21)

TABLE 2.1-1 Watts Bar Unit 2 Heatup Limits 32 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology Data (Ref. 1) plotted on Figure 2.1-1 LEAK HEATUP CRITICALITY HEATUP CRITICALITY TEST RATE LIMITS RATE LIMITS LIMITS (60!F/HR) (60!F/HR) (100!F/HR) (100!F/HR)

T P T P T P T P T P

(!F) (psig) (!F) (psig) (!F) (psig) (!F) (psig) (!F) (psig) 132 2000 60 0 149 0 60 0 149 0 149 2485 60 621 149 1021 60 621 149 904 65 621 150 1032 65 621 150 909 70 621 155 1076 70 621 155 934 75 621 160 1126 75 621 160 963 80 621 165 1183 80 621 165 996 85 621 170 1246 85 621 170 1035 90 621 175 1317 90 621 175 1079 95 621 180 1395 95 621 180 1129 100 621 185 1483 100 621 185 1185 100 960 190 1580 100 873 190 1248 105 993 195 1687 105 889 195 1318 110 1032 200 1806 110 909 200 1396 115 1076 205 1938 115 934 205 1483 120 1126 210 2083 120 963 210 1580 125 1183 215 2244 125 996 215 1686 130 1246 220 2421 130 1035 220 1805 135 1317 135 1079 225 1936 140 1395 140 1129 230 2080 145 1483 145 1185 235 2240 150 1580 150 1248 240 2417 155 1687 155 1318 160 1806 160 1396 165 1938 165 1483 170 2083 170 1580 175 2244 175 1686 180 2421 180 1805

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 243 of 262 Appendix B (Page 8 of 21)

TABLE 2.1-1 Watts Bar Unit 2 Heatup Limits 32 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology Data (Ref. 1) plotted on Figure 2.1-1 LEAK HEATUP CRITICALITY HEATUP CRITICALITY TEST RATE LIMITS RATE LIMITS LIMITS (60!F/HR) (60!F/HR) (100!F/HR) (100!F/HR)

T P T P T P T P T P

(!F) (psig) (!F) (psig) (!F) (psig) (!F) (psig) (!F) (psig) 185 1936 190 2080 195 2240 200 2417

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 244 of 262 Appendix B (Page 9 of 21)

MATERIAL PROPERTY BASIS LIMITING MATERIAL: Intermediate Shell Forging 05 using Reg. Guide 1.99 Position 1.1 LIMITING ART VALUES AT 32 EFPY: 1/4T, 88!F (Axial Flaw) 3/4T, 71!F (Axial Flaw)

Figure 2.1-2 Watts Bar Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100!F/hr) Applicable for 32 EFPY (with Flange Requirements and without Margins for Instrumentation Errors) using the 1998 through the 2000 Addenda App. G Methodology (w/ KIc)

(Plotted data (Ref. 1) provided in Table 2.1-2)

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 245 of 262 Appendix B (Page 10 of 21)

TABLE 2.1-2 Watts Bar Unit 2 Cooldown Limits 32 EFPY Heatup Curve Data Points Using 1998 through 2000 Addenda App. G Methodology (Data (Ref. 1) plotted on Figure 2.1-2)

Steady State 20!F/HR 40!F/HR 60!F/HR 100oF/HR T P T P T P T P T P

(!F) (psig) (!F) (psig) (!F) (psig) (!F) (psig) (!F) (psig) 60 0 60 0 60 0 60 0 60 0 60 621 60 621 60 621 60 621 60 621 65 621 65 621 65 621 65 621 65 621 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 100 1080 100 1075 100 1075 100 1075 100 1075 105 1130 105 1130 105 1130 105 1130 105 1130 110 1185 110 1185 110 1185 110 1185 110 1185 115 1247 115 1247 115 1247 115 1247 115 1247 120 1315 120 1315 120 1315 120 1315 120 1315 125 1390 125 1390 125 1390 125 1390 125 1390 130 1473 130 1473 130 1473 130 1473 130 1473 135 1564 135 1564 135 1564 135 1564 135 1564 140 1665 140 1665 140 1665 140 1665 140 1665 145 1777 145 1777 145 1777 145 1777 145 1777 150 1901 150 1901 150 1901 150 1901 150 1901 155 2037 155 2037 155 2037 155 2037 155 2037 160 2188 160 2188 160 2188 160 2188 160 2188 165 2355 165 2355 165 2355 165 2355 165 2355

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 246 of 262 Appendix B (Page 11 of 21)

Setpoint Window Figure 3.1-1 PORV Setpoint vs RCS Temperature (Plotted data (Ref. 2) provided in Table 3.1-1)

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 247 of 262 Appendix B (Page 12 of 21)

TABLE 3.1-1 Watts Bar Unit 2 PORV Setpoints vs Temperature (Data (Ref. 2) Plotted on Figure 3.1-1)

Temperature PCV-334 Setpoint PCV-340A Setpoint

(!F) (psig) (psig) 60 416 409 117 416 409 125 490 483 167 490 483 190 696 590 225 696 590 300 696 590 350 696 590 450 2335 2335

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 248 of 262 Appendix B (Page 13 of 21)

TABLE 4.0-1 Watts Bar Unit 2 Surveillance Capsule Removal Schedule (a)

Capsule Orientation of Lead Removal Expected Capsule Capsule Factor Time Fluence (n/cm2,E > 1.0 MeV)

U Dual 34! 4.70 2.0 EFPY 0.604 x 1019 (EOC 2)

W Single 34! 4.66 7.0 EFPY 1.94 x 1019 (b)

X Dual 34! 4.69 7.0 EFPY to 1.94 x 1019 to 13.7 EFPY 3.88 x 1019 (c)

Z Single 34! 4.69 Note (d) Note (d)

V Dual 31.5! 4.04 Note (d) Note (d)

Y Dual 31.5! 4.04 Note (d) Note (d)

Notes:

(a) This information is taken from the withdrawal schedule contained in WCAP-18518-NP (Ref. 10). EOC = End-of-Cycle (b) Approximate Fluence at vessel inner wall at End-of-Life (32 EFPY). This capsule should be withdrawn at the outage nearest to but following 7.0 EFPY of operation.

(c) Capsule X should be removed between 11.7 EFPY and 13.7 EFPY if possible. Capsule X must be removed between 7.0 EFPY and 13.7 EFPY in order to satisfy the recommendations of the third capsule end-of-license per ASTM E185-82 (Ref. 7). This removal EFPY should be re-visited at a later date, such as after Capsules W is removed.

(d) Capsules Z, V, and Y should remain in the reactor. If additional metallurgical data is needed, withdrawal and testing of these capsules should be considered. In the event that Capsule W cannot be removed, then Capsule Z may serve as a backup and be removed instead during the same outage.

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 249 of 262 Appendix B (Page 14 of 21)

Table 5-1 Summary of the Best Estimate Cu and Ni Weight Percent and Initial RTNDT Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials(a)

Material Description Chemical Composition Initial Cu Ni RTNDT wt. % wt. %

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 0.05 0.78 14!F Lower Shell Forging 04 0.05 0.81 5!F Intermediate to Lower Shell Circumferential 0.05 0.70 -50!F Weld Seam W05 (Heat #895075)

Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 0.07 0.91 -14!F Upper to Intermediate Shell Circumferential 0.03 0.75 10!F Weld Seam W06 (Heat #899680)

Lower Shell to Bottom Head Ring 0.03 0.75 10!F Circumferential Weld Seam W04 (Heat

1. 899680)

Bottom Head Ring 03 0.06 0.86 -40!F Note:

(a) Values taken from WCAP-18191-NP (Ref. 1). The initial RTNDT values are measured values. The reactor vessel nozzle materials are not considered part of the beltline or extended beltline since the nozzle material fluence values fall below the 1 x 1017 n/cm2 (E > 1.0 n/cm2) threshold defined by NRC RIS 2014-11 (Ref. 9). Nozzle forging material properties and ART values are detailed in Appendix B of WCAP-18191-NP (Ref. 1).

Table 5-2 Summary of the Initial RTNDT Values for the Watts Bar Unit 2 Closure Head and Vessel Flange Material Identification Initial RTNDT(a)

Closure Head Flange -40!F Vessel Flange -22!F Note:

(a) The initial RTNDT values are measured values, taken from WCAP-18191-NP (Ref. 1) and consistent with WCAP-13830, Revision 1 (Ref. 8)

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 250 of 262 Appendix B (Page 15 of 21)

Table 5-3 Summary of the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Material Position 1.1 Chemistry Factors Material Description Position 1.1 Chemistry Factor Reactor Vessel Beltline Materials Intermediate Shell Forging 05 31.0!F Lower Shell Forging 04 31.0!F Intermediate to Lower Shell 68.0!F Circumferential Weld Seam W05 Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 44.0!F Upper to Intermediate Shell Circumferential Weld 41.0!F Seam W06 Lower Shell to Bottom Head Ring Circumferential 41.0!F Weld Seam W04 Bottom Head Ring 03 37.0!F Table 5-4 Watts Bar Unit 2, Catawba Unit 1, Watts Bar Unit 1, and McGuire Unit 2 Surveillance Weld Data for Heat #895075(a)

Material Capsule Cu Ni Irradiation Capsule #RTNDT (wt. %) (wt. %) Temperature fluence (!F)

(!!F) (x 1019 n/cm2, E > 1.0 MeV)

Watts Bar U 0.033 0.70 559 0.604 32.6 Unit 2 Catawba Z 0.05 0.73 562 0.286 1.91 Unit 1 Y 0.05 0.73 562 1.29 17.79 V 0.05 0.73 562 2.27 26.5 Watts Bar U 0.03 0.75 560 0.447 0.0 Unit 1 W 0.03 0.75 560 1.08 30.5 X 0.03 0.75 560 1.71 25.8 Z 0.03 0.75 560 2.40 13.9 McGuire V 0.04 0.74 557 0.302 38.51 Unit 2 X 0.04 0.74 557 1.38 35.93 U 0.04 0.74 557 1.90 23.81 W 0.04 0.74 557 2.82 43.76 Note:

(a) Surveillance data taken from WCAP-18532-NP (Ref. 11).

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 251 of 262 Appendix B (Page 16 of 21)

Table 5-5 Calculation of the Watts Bar Unit 2 Heat # 895075 Position 2.1 Chemistry Factor Using Surveillance Capsule Data Material Capsule Capsule f(a) FF(b) #RTNDT(c) FF* #RTNDT FF2 (x 1019 n/cm 2, (!F)

E > 1.0 MeV)

Watts Bar U 0.604 0.859 52.2 (32.6) 44.87 0.738 Unit 2 Catawba Z 0.286 0.658 6.9 (1.91) 4.55 0.433 Unit 1 Y 1.29 1.071 22.8 (17.79) 24.41 1.147 V 2.27 1.222 31.5 (26.5) 38.49 1.493 Watts Bar U 0.447 0.776 5.0 (0.0) 3.86 0.602 Unit 1 W 1.08 1.022 55.6 (30.5) 56.81 1.044 X 1.71 1.148 47.8 (25.8) 54.87 1.317 Z 2.40 1.236 28.1 (13.9) 34.67 1.528 McGuire V 0.302 0.672 48.5 (38.51) 32.61 0.452 Unit 2 X 1.38 1.089 45.3 (35.93) 49.32 1.187 U 1.90 1.176 30.0 (23.81) 35.27 1.382 W 2.82 1.276 55.1 (43.76) 70.35 1.628 SUM: 450.07 12.949 CFWeld Heat #895075 = (FF * #RTNDT) (FF2) = (450.07) (12.949) = 34.8!F Notes:

(a) f = fluence (b) FF = fluence factor = f(0.28-0.10*log(f))

(c) #RTNDT values are the measured 30 ft-lb shift values. The #RTNDT values are adjusted first by the difference between the capsule irradiation temperature and the Watts Bar Unit 2 operating temperature, then using the ratio procedure to account for differences in the surveillance weld chemistry and the reactor vessel weld chemistry. Pre-adjusted values are listed in parentheses. The temperature adjustments for each capsule were calculated from the data in Table 5-4 and the plant irradiation temperature for Watts Bar Unit 2 is 557!F. The #RTNDT values for Watts Bar Unit 2, Catawba Unit 1, Watts Bar Unit 1, and McGuire Unit 2 were adjusted by ratios of 1.51, 1.00, 1.66, and 1.26 respectively.

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 252 of 262 Appendix B (Page 17 of 21)

Table 5-6 Fluence Values for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials 32 EFPY Fluence Material Description (n/cm2, E > 1.0 MeV)

Clad/Base 1/4T Location 3/4T Location Metal Interface (x=2.116 in.) (x=6.349 in.)

(Inner Surface)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 1.86E+19 1.12E+19 4.05E+18 Lower Shell Forging 04 1.94E+19 1.17E+19 4.23E+18 Intermediate to Lower Shell 1.83E+19 1.10E+19 3.99E+18 Circumferential Weld Seam W05 Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 5.12E+17 3.08E+17 1.12E+17 Upper to Intermediate Shell 5.12E+17 3.08E+17 1.12E+17 Circumferential Weld Seam W06 Lower Shell to Bottom Head Ring 2.47E+18 1.49E+18 5.38E+17 Circumferential Weld Seam W04 Bottom Head Ring 03 2.47E+18 1.49E+18 5.38E+17

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 253 of 262 Appendix B (Page 18 of 21)

Table 5-7 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 1/4T Location Reactor Vessel CF 1/4T f 1/4T #RTNDT IRTNDT(a) l(a) # M ART Location (!!F) (n/cm2, FF (!F) (!F) (!F) (!F) (!F) (!F)

E>1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell 31.0 1.12E+19 1.031 32.0 14 0.0 16.0 32.0 78.0 Forging 05 Lower Shell Forging 31.0 1.17E+19 1.043 32.3 5 0.0 16.2 32.3 69.7 04 Intermediate to 68.0 1.10E+19 1.027 69.8 -50 0.0 28.0 56.0 75.8 Lower Shell Circumferential Weld Seam W05 Using Surveillance 34.8 1.10E+19 1.027 35.7 -50 0.0 14.0 28.0 13.7 Capsule Data(b)

Reactor Vessel Extended Beltline Materials Upper Shell Forging 44.0 3.08E+17 0.223 9.8 -14 0.0 4.9 9.8 5.6 06 Upper to 41.0 3.08E+17 0.223 9.1 10 0.0 4.6 9.1 28.3 Intermediate Shell Circumferential Weld Seam W06 Lower Shell to 41.0 1.49E+18 0.501 20.5 10 0.0 10.3 20.5 51.1 Bottom Head Ring Weld Seam W04 Bottom Head Ring 37.0 1.49E+18 0.501 18.5 -40 0.0 9.3 18.5 -2.9 03 Notes:

(a) The initial RTNDT values are measured values; therefore, l = 0!F (b) The Heat #895075 surveillance data is deemed credible per WCAP-18518-NP (Ref. 10).

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 254 of 262 Appendix B (Page 19 of 21)

Table 5-8 Adjusted Reference Temperature Evaluation for the Watts Bar Unit 2 Reactor Vessel Beltline and Extended Beltline Materials through 32 EFPY at the 3/4T Location Reactor Vessel CF 3/4T f 3/4T #RTNDT IRTNDT(a) l(a) # M ART Location (!!F) (n/cm2, FF (!F) (!F) (!F) (!F) (!F) (!F)

E>1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell 31.0 4.05E+18 0.750 23.2 14 0.0 11.6 23.2 60.5 Forging 05 Lower Shell Forging 31.0 4.23E+18 0.761 23.6 5 0.0 11.8 23.6 52.2 04 Intermediate to 68.0 3.99E+18 0.745 50.7 -50 0.0 25.3 50.7 51.3 Lower Shell Circumferential Weld Seam W05 Using Surveillance 34.8 3.99E+18 0.745 25.9 -50 0.0 13.0 25.9 1.9 Capsule Data(b)

Reactor Vessel Extended Beltline Materials Upper Shell Forging 44.0 1.12E+17 0.118 5.2 -14 0.0 2.6 5.2 -3.6 06 Upper to 41.0 1.12E+17 0.118 4.8 10 0.0 2.4 4.8 19.7 Intermediate Shell Circumferential Weld Seam W06 Lower Shell to 41.0 5.38E+17 0.305 12.5 10 0.0 6.2 12.5 35.0 Bottom Head Ring Weld Seam W04 Bottom Head Ring 37.0 5.38E+17 0.305 11.3 -40 0.0 5.6 11.3 -17.5 03 Notes:

(a) The initial RTNDT values are measured values; therefore, l = 0!F (b) The Heat #895075 surveillance data is deemed credible per WCAP-18518-NP (Ref. 10).

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 255 of 262 Appendix B (Page 20 of 21)

Table 5-9 Summary of the Limiting ART Values Used in the Generation of the Watts Bar Unit 2 Heatup/Cooldown Curves EFPY Limiting ART(a) (!!F) 1/4T 3/4T 32 88 71 Note:

(a) The ART Values used for heatup and cooldown limit curve development are the limiting ART values calculated in Tables 5-7 and 5-8 (corresponding to Intermediate Shell Forging 05) rounded and increased by 10!F to add additional margin; this approach is conservative.

System REACTOR COOLANT SYSTEM SDD-N3-68-4001 Description Unit 1 / Unit 2 Rev. 0048 Document QA Record Page 256 of 262 Appendix B (Page 21 of 21)

Table 5-10 RTPTS Calculations for the Watts Bar Unit 2 Beltline and Extended Beltline Materials at 32 EFPY Material CF 32 EFPY Fluence FF(a) IRTNDT #RTNDT u(b) #(c) M (d) RTPTS(e)

(!!F) (n/cm2, (!F) (!F) (!F) (!F) (!F)

(!F)

E>1.0 MeV)

Reactor Vessel Beltline Materials Intermediate Shell Forging 05 31.0 1.86E+19 1.170 14 36.3 0.0 17.0 34.0 84.3 Lower Shell Forging 04 31.0 1.94E+19 1.181 5 36.6 0.0 17.0 34.0 75.6 Intermediate to Lower Shell 68.0 1.83E+19 1.166 -50 79.3 0.0 28.0 56.0 85.3 Circumferential Weld Seam W05 Using Surveillance Capsule 34.8 1.83E+19 1.166 -50 40.6 0.0 14.0 28.0 18.6 Data Reactor Vessel Extended Beltline Materials Upper Shell Forging 06 44.0 5.12E+17 0.296 -14 13.0 0.0 6.5 13.0 12.1 Upper to Intermediate Shell 41.0 5.12E+17 0.296 10 12.2 0.0 6.1 12.2 34.3 Circumferential Weld Seam W06 Lower Shell to Bottom Head 41.0 2.47E+18 0.621 10 25.5 0.0 12.7 25.5 60.9 Ring Weld Seam W04 Bottom Head Ring 03 37.0 2.47E+18 0.621 -40 23.0 0.0 11.5 23.0 6.0 Notes:

(a) FF = fluence factor = f(0.28-0.1log(f))

(b) As indicated in Table 5-1 of this report, the IRTNDT values are measured; hence, according to 10 CFR 50.61, u = 0!F (c) Per the guidance of 10 CFR 50.61, the base metal # = 17!F and the weld metal # = 28!F when surveillance data is not utilized. Also per 10 CFR 50.61, # = 14!F for weld metal with credible surveillance data. However, # need not exceed 0.5*#RTNDT (d) M = Margin = 2 * (u2 + #2)1/2 (e) RTPTS = IRTNDT + #RTPTS + Margin