NRC 2020-0001, Pressure Temperature Limits Report (PTLR)

Pressure Temperature Limits Report (PTLR)

ML20009E096
Person / Time
Site:	Point Beach
Issue date:	01/09/2020
From:	Schultz E Point Beach
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2020-0001
Download: ML20009E096 (16)

Text

ENER(3Y January 9, 2020 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington DC 20555 Point Beach Nuclear Plant Units 1 and 2 Docket Nos. 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Pressure Temperature Limits Report (PTLR)

In accordance with the requirements of Point Beach Nuclear Plant (PBNP)

POINT BEACH NRC 2020-0001 TS 5.6.5 Technical Specification 5.6.5, enclosed is TRM 2.2, Pressure Temperature Limits Report (PTLR).

The Enclosure to this letter contains the PBNP PTLR that was issued on December 17, 2019.

This letter contains no new regulatory commitments and no revisions to existing commitments.

Sincerely, NextEra Energy Point Beach, LLC Eric Schultz Licensing Manager Enclosure cc:

Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW NextEra Energy Point Beach, LLC

~

~-, --.

6610 Nuclear Road, Two Rivers, WI 54241

ENCLOSURE NEXTERA ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 and 2 PRESSURE TEMPERATURE LIMITS REPORT 14 pages follow

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Note: Applicability limits for pressure temperature limits are discussed in Section 2.0, "Operating Limits."

1.0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

This RCS Pressure and Temperature Limits Report (PTLR) for Point Beach TRM 2.2 Nuclear Plant Units 1 and 2 has been prepared in accordance with the requirements of Technical Specification 5.6.5.

The analytical methods used to determine the RCS pressure and temperature limits shall be those previously reviewed and approved by the NRC; specifically, those described in NRC Safety Evaluations dated October 6, 2000 (Ref5.1), July 23, 2001 (Ref5.2h October 18, 2007 (Ref5.3), and June 30, 2014 (Ref5.4)*

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto (Ref5.16). Based upon fluence values in Westinghouse WCAP-16669-NP (Ref5.5), this PTLR is effective for 50 EFPY (approximately through the end of year 2029).

The Technical Specifications addressed in this report are listed below:

1.1 3.4.3 Pressureff emperature (P-T) Limits 1.2 3.4.12 Low Temperature Overpressure Protection (L TOP) System 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. Changes to these limits must be developed using the NRC approved methodologies specified in Technical Specification 5.6.5.

These limits have been determined such that applicable limits of the safety analysis are met. Items that appear in capitalized type are defined in Technical Specification 1.1, "Definitions."

2.1 RCS Pressure and Temperature Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits are:

a. A maximum heatup rate of 100°F in any one hour. (Ref5.5)

b. A maximum cooldown rate of 100°F in any one hour. (Ref 5.5)

c. An average temperature change of :s;10°F per hour during inservice leak and hydrostatic testing operations. (Ref5.5) 2.1.2 The RCS P-T limits for heatup and cooldown are specified by Figures 1 and 2, respectively. *(Ref 5.5)

POINT BEACH TRM 2.2 -1 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL TRM 2.2 PRESSURE TEMPERATURE LIMITS REPORT 2.1.3 Using the approved methodology of ASME XI Appendix G, the minimum temperature for pressurization of the RCS is 60°F based on the most limiting of the two Point Beach reactor vessels' beltline welds.

The requirements of 10 CFR 50 Appendix G also limit operation with the reactor vessel head tensioned ("bolt-up") to ;:::60°F based on the most limiting of the two Point Beach reactor vessels' flange region materials.

Correcting for possible instrument uncertainties, 60°F actual corresponds to a minimum indicated temperature of 78°F when reading the RCS loop wide range cold leg meter, or 70° when reading the A RHR heat exchanger outlet temperature (TE-627) on PPCS.

Alternatively, the minimum acceptable reactor vessel flange temperature (not the reactor head flange temperature) for tensioned or pressurized operation is 70°F as indicated by hand-held digital pyrometer reading. (Rers.s) 2.2 Low Temperature Overpressure Protection System Enable Temperature (LCO 3.4.6. 3.4.7. 3.4.10 and 3.4.12)

The enable temperature for the Low Temperature Overpressure Protection System is 285°F (includes instrument uncertainty for RCS Tc wide range). (Ref5.6) 2.3 Low Temperature Overpressure Protection System Setpoints (LCO 3.4.12)

Pressurizer Power-Operated Relief Valve Lift Setting Limits The limiting trip setpoint (Ref 5.6) for the pressurizer power-operated relief valves (PORVs) is :::;420 psig (includes instrument uncertainty).

The following operating restrictions ensure continued operability of the L TOP system:

2.3.1 RCP Operating Restriction - No more than one RCP in operation for RCS temperature <180°F. (Ref5.6) 2.3.2 Charging Pumps - Limit the number of operating charging pumps to two when L TOP is in service. (Ref 5.6) 2.4 Criticality and Hydrostatic Leak Test Limits 2.4.1 Criticality and hydrostatic leak test limits are shown on the RCS Pressure Temperature Limits for heatup, Figure 1. (Ref 5.5)

POINT BEACH TRM 2.2-2 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedules for Units 1 and 2 are provided in Tables 1 and 2, respectively.

For the period of the renewed facility operating license, all capsules in the reactor vessel that are removed and tested shall meet the test procedures and reporting requirements of ASTM E 185-82. Any changes to the capsule withdrawal schedule, including spare capsules, shall be approved by the NRC prior to implementation. (Ref 5.13 and 5.14)

The pressure vessel surveillance program is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the nil-ductility temperature, RT Nor, which is determined in accordance with ASTM E208. The empirical relationship between RT NDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure/' to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

Surveillance specimens for the limiting materials for the PBNP reactor vessels are not included in the plant specific surveillance program. Therefore, the results of the examinations of these specimens do not meet the credibility criteria of Regulatory Guide 1.99, Revision 2, for PBNP Units 1 and 2.

During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRG-approved Pressurized Water Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref. 5.8)( Ref 5.15) 4.0 SUPPLEMENTAL DATA INFORMATION The limiting RT prs values for the PBNP limiting beltline materials at 50 EFPY are:

Unit 1 - Intermediate to Lower Shell Gire Weld = 245.3 °F; Intermediate Shell Long. Weld= 236.0 °F (Ref. 5.11Table1)

Unit 2 - Intermediate to Lower Shell Circ Weld = 280.6 °F; Intermediate Shell Forging = 155.4 °F (Ref. 5.11Table1)

POINT BEACH TRM 2.2 -3 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT

5.0 REFERENCES

TRM 2.2 5.1 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Exemption from the Requirements of 10CFR50.60 (TAC NOS. MA9680 and MA9681)," dated October 6, 2000 5.2 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Acceptance of Methodology for Referencing Pressure Temperature Limits Report (TAC Nos. MA8459 and MA8460)," dated July 23, 2001 5.3 NRC SE "Amendment Nos. 229/234 to Facility Operating Licenses DPR-24 and DPR-27, (approving use of FERRET Code as approved methodology for determining RCS pressure and temperature limits)," dated October 18, 2007 5.4 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendment (Nos. 250 and 254) Regarding Change to Technical Specification 5.6.5, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) (TAC Nos. MF0532 and MF0533)," and NRC Letter, "Point Beach Nuclear Plant (Point Beach), Units 1 and 2 - Exemption From the Requirements of 10 CFR Section 50.61 and Appendix G to 10 CFR Part 50 (TAC Nos. MF0534 and MF0535),"

dated June 30, 2014 5.5 WCAP-16669, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," Revision 1, January 2009 5.6 Westinghouse calculation CN-SCS-06-68, "L TOPS Setpoint Analysis for Point Beach Units 1 and 2," 10/27/2011

5. 7 WEPCO Calculation Addendum No. 98-0156-00-A, Revision 0, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1," 9/22/1999 5.8 PWR Owner Group Topical Report BAW-1543(NP), Revision 4, Supplement 7-A, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" March 2018(ML18184A520, includes NRC Final SER, ML18038A345) 5.9 BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998 5.10 CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997

5. 11 Areva Calculation 32 9019238 000, "RTPTS Values for Point Beach Unit 1 and Unit 2," Revision 0, May 2006

5. 12 Areva Calculation 32-9019240-000, "ART Values for Point Beach Unit 1 and Unit 2," Revision 0, June 2006 POINT BEACH TRM 2.2 -4 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TRM 2.2 5.103 Renewed Facility Operating License DPR-24, Point Beach Nuclear Plant Unit 1 5.114 Renewed Facility Operating License DPR-27, Point Beach Nuclear Plant Unit 2 5.15 Deleted [NRG 2010-0004]

5.16 Point Beach Technical Specifications, LCO 5.6.5.c 5.17 WCAP-15856, Rev. 0, "Supplemental Reactor Vessel Surveillance Capsule "A" for the Point Beach Units 1 and 2 Reactor Vessel Installed in the Point Beach Unit 2 Reactor Vessel, "May 2002.

5.18 AR 2309642, "Clarification Required Regarding Implementation of LCO 3.4.3" 5.19 AR 2316053, "Revise TRM 2.2, PTLR Requirement 2.1.3" POINT BEACH TRM 2.2-5 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Figure 1 RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP PSNP 100'Ffhr l-learup Limits Moderator-Temperature r.F)

POINT BEACH TRM 2.2-6 TRM 2.2 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Figure 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN PBNP 100°FlhrCooldown Limits so 100 300 350 POINT BEACH TRM 2.2-7 400 TRM 2.2 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 1*

POINT BEACH NUCLEAR PLANT UNIT 1 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Capsule Identification Letter Approximate Removal Date**

v EOC-1 (Sept 1972) s EOC-3 (Dec 1975)

R EOC-5 (Oct 1977)

T EOC-11 (Mar 1984) p EOC-21 (Apr 1994-Stored in SFP)

N Standby TABLE 2*

POINT BEACH NUCLEAR PLANT UNIT 2 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Capsule Identification Letter Approximate Removal Date**

v EOC-1 (Nov 1974)

T EOC-3 (Mar 1977)

R EOC-5 (Apr 1979) s EOC-16 (Oct 1990) p EOC-22 (Jun 1997-Stored in SFP)

N Standby Supplemental Capsule "A"***

43 EFPY (-Fall 2024)

• During the period of extended operation, reactor vessel surveillance capsules will be removed and tested in accordance with the schedule contained in the most recently NRG approved Pressurized Water Reactor Owners Group (PWROG) Master Integrated Reactor Vessel Surveillance Program (MIRVSP) Document. (Ref.

5.8)

• • For capsules that have not been withdrawn yet, the actual dates will be adjusted to coincide with the closest scheduled plant refueling outage.

• • • Supplemental Capsule "A" (also identified as Supplemental Capsule "W' in some documents) was installed in Cycle 25 and is described in WCAP-15856. (Ref 5.17) The removal date is provided in effective full power years (EFPY) as agreed between the NRG and PWROG. (Ref. 5.8)

POINT BEACH TRM 2.2-8 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 3 POINT BEACH UNIT 1 RPV BEL TUNE 50 EFPY VALUES(EJ Based on Westinghouse Report, WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.5).

Vessel Manufacturer:

Babcock & Wilcox Plate and Weld Thickness (without cladding):

6.5", without clad !DJ Component Description Heat or Heat/Lot Inside Surface 1/4T Fluence 1/4T Fluence 3/4T Fluence 3/4T Fluence Fluence (E19 n/cm2)

(E19 n/cm2) (BJ Factor (CJ (E19 n/cm2) (BJ Factor (CJ Nozzle Belt Forging 122P237 0.36 0.24 0.62 0.11 0.44 Intermediate Shell Plate A9811-1 4.90 3.32 1.31 1.52 1.12 Lower Shell Plate C1423-1 4.55 3.09 1;30 1.41 1.10 Nozzle Belt to lntermed. Shell 8T1762 0.36 0.24 0.62 0.11 0.44 CircWeld (100%)

(SA-1426)

Intermediate Shell Long (A) 1P0815 (SA-812) 3.19 2.16 1.21 N/A N/A Weld (ID 27%)

Intermediate Shell Long (A) 1 P0661 (SA-775) 3.19 N/A N/A 0.99 1.00 Weld (OD 73%)

lntermed. to Lower Shell (A) 71249 (SA-1101) 4.43 3.00 1.29 1.38 1.09 Circ. Weld (100%)

Lower Shell Long Weld 61782 (SA-847) 3.05 2.07 1.20 0.95 0.99 (100%)

Footnotes:

(A) ** Limiting material TRM 2.2 (B)

From an inside surface fluence value (not including cladding), fluence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2: f = fsurt x e-0*24\\ where fsurt is expressed in units of E19 n/cm2, E>1 MeV, and xis the desired depth in inches into the vessel wall~

(C)

• The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99, Revision 2: ff= f 028 0

10109~, where f is the fluence in units of E19 n/cm2.

(0) Instruction Manual, 132-lnch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969 (E)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. ~

POINT BEACH TRM 2.2-9 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 POINT BEACH UNIT 2 RPV BEL TUNE 50 EFPY VALUES<EJ TRM 2.2 Based on Westinghouse Report WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.5).

Vessel Manufacturer:

Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding):

6.5", without clad (DJ Heat or Inside Surface 1/4T Fluence 1/4T Fluence 3/4T Fluence 3/4T Fluence Component Description Heat/Lot Fluence (E19 n/cm2) rs; Factor re; (E19 n/cm2) rs; Factor re; (E19 n/cm2)

Nozzle Belt Forging 123V352 0.50 0.34 0.70 0.16 0.51 Intermediate Shell Forging (Al 123V500 5.05 3.42 1.32 1.57 1.12 Lower Shell Forging 122W195 4.90 3.32 1.31 1.52 1.12 Nozzle Belt to lntermed. Shell 21935 0.50 0.34 0.70 0.16 0.51 Circ Weld (100%)

lntermed. to Lower Shell Circ 72442 4.65 3.15 1.30 1.44 1.10 Weld (100%) (Al (SA-1484)

Footnotes:

(A)

• Limiting Material (8) From an inside surface fluence value (not including cladding), fluence is attenuated to a desired thickness using equation (3) of Regulatory Guide 1.99, Revision 2: f = fsur1 x e.o*24X, where fsur1 is expressed in units of E19 n/cm2, E>1 MeV, and x is the desired depth in inches into the vessel wall.

(C) The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99, Revision 2: ff= f 0*28

• 0*10 tog n, where f is the fluence in units of E19 n/cm2*

(D) ' Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970.

(E)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.

POINT BEACH TRM 2.2 -10 REV. 11 12/2019

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 5 POINT BEACH UNIT 1 RPV 1/4T BEL TUNE MATERIAL ADJUSTED REFERENCE TEMPERA TURES AT 50 EFPY.(HJ TRM 2.2 Unless otherwise noted, all ART input data obtained from WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.5).

Vessel Manufacturer:

I Babcock & Wilcox I

Plate and Weld Thickness (without cladding):

I 6.5", without clad!FJ I

Component Description Heat or Initial

%Cu

%Ni CF CF 1/4T Fluence

.b..RTNDT O"l(G)

O"A(G)

Margin Heat/Lot RTNDT (°F)

Method Factor!AJ (oF)

(oF)

Nozzle Belt Forging 122P237

+50 0.11 0.82 77.0 Table 0.62 47.4 0

17.0 34.0 Intermediate Shell Plate A9811-1

+1 0.20 0.06 79.3 Surv.

1.31 104.1 26.9 8.5 56.4 Data(8)

Lower Shell Plate C1423-1

+1 0.12 0.07 35.8 Surv.

1.30 46.4 26.9 8.5 56.4 Data(8)

Nozzle Belt to lntermed. Shell 8T1762

-47.6 0.19 0.57 167.0 Table 0.62 102.9 17.2 28 65.7 CircWeld (100%)

(SA-1426)

Intermediate Shell Long Weld (ID 27%)

1P0815

-47.6 0.17 0.52 167.0 Table 1.21 201.9 17.2 28 65.7 (SA-812)

Intermediate Shell Long Weld 1P0661

-47.6 0.17 0.64 167.0 Table NIA NIA 17.2 28 NIA (0073%)

(SA-775) lntermed. to Lower Shell Circ. Weld 71249

-47.4 0.23 0.59 167.6 Table(Cl 1.29 216.4 12.9 28 61.7 (100%)

(SA-1101)

Lower Shell Long Weld (100%)

61782

-47.6 0.23 0.52 167.0 Table 1.20 199.9 17.2 28 65.7 (SA-847)

Footnotes:

(AJ ' See Table 3 (BJ

• Credible Surveillance Data; see BAW-2325 for evaluation.

(CJ Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measure t.RT NoT and predicted LiRT NOT based on Table CF is less than 2cr (56°F).

(DJ Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1,"

(Ref.5.7) utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.

(EJ Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART= Initial RT NOT+ LiRT NOT+ Margin, where LiRT NOT= Chemistry Factor x Fluence Factor, and Margin= 2(cr12 + cr~

2) 0 5

, with cr1 defined as the standard deviation of the Initial RTNoT and cr~ defined as the standard deviation of LiRTNoT-(FJ Instruction Manual, 132-lnch l.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

(GJ cr1 and cr~ values obtained from Areva Calculation 32-9019240-000, Table 41 (Ref. 5.12)

(HJ EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.

POINT BEACH TRM 2.2 - 11 REV. 11 12/2019 ART (oF)IEJ 131.4 161.5 103.8 121.0 220.0 NIA 230.7 218.1

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 6 POINT BEACH UNIT 2 RPV 1/4T BEL TUNE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 50 EFPY.('J TRM 2.2 Unless otherwise noted, all ART input data obtained from WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.5).

Vessel Manufacturer:

Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding):

6.5", without cladlFJ Component Description Heat or Initial

%Cu

%Ni CF CF 1/4T-Fluence h.RTNDT cn!Gl cr,i.(G)

Margin Heat/Lot RTNDT (°F)

Method FactorlAJ (oF)

(oF)

Noz2:1e Belt Forging 123V352

+40 0.11 0.73 76.0 Table 0.70 53.5 0

17.0 34.0 Intermediate Shell Forging 123V500

+40 0.09 0.70 58.0 Table18l 1.32 76.6 0

17.0 34.0 Lower Shell Forging 122W195

+40 0.05 0.72 42.8 Surv.

1.31 56.2 0

8.5 17.0 Data!Cl Nozzle Belt to lntermed. Shell Circ Weld 21935

-56 0.18 0.70 170.5 Table!Hl 0.70 120.0 17.0 28 65.5 (100%)

lntermed. to Lower Shell Circ. Weld 72442 (SA-1484)

-30 0.26 0.60 180.0 Tablel0l 1.30 234.4 11.9 28 60.8 (100%)

Footnotes:

(A)

• See Table 4 (B). Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured LI.RT Nor and predicted LI.RT Nor based on Table CF is Jess than 2cr (34°F)

(C) i Credible surveillance data; see BAW-2325 for evaluation.

(D) : Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore, conservative.

(E) Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART= Initial RT Nor + LI.RT Nor + Margin, where LI.RT Nor= Chemistry Factor x Fluence Factor, and Margin = 2(cr12 + c;,,.2)0*5, with cr1 defined as the standard deviation of the Initial RT NDT, and"" defined as the standard deviation of LI.RT NDT*

(F)

Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant Unit 2, Combustion Engineering, CE Book #4869, October 1970.

(G) 1 cr1and ""values obtained from Areva Calculation 32-9019240-000, Table 41 (Ref. 5.12)

(H) Table CF value based on best-estimate chemistry data from CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 (Ref.5.10).

(/), EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.

POINT BEACH TRM 2.2 - 12 REV. 11 12/2019 ART (oF)IEJ 127.5 150.6 113.2 129.5 265.2

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 7 POINT BEACH UNIT 1 RPV 3/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 50 EFPY(HJ TRM 2.2 Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAJ) Regarding Reactor Pressure Vessel Integrity," May 1998 (Ref. 5.9) and..WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.5).

Vessel Manufacturer:

I Babcock & Wilcox I

Plate and Weld Thickness (without cladding):

I 6.5", without cladlFJ I

Component Description Heat or Heat/Lot Initial

%Cu

%Ni CF CF 3/4T-Fluence LiRTNDT O"l(G) cr,i.(G)

Margin RTNDT (°F)

Method FactorlAJ (oF)

(oF)

Nozzle Belt Forging 122P237

+50 0.11 0.82 77.0 Table 0.44 33.7 0

17.0 33.7 Intermediate Shell Plate A9811-1

+1 0.20 0.06 79.3 Surv.

1.12 88.5 26.9 8.5 56.4 Data<8l Lower Shell Plate C1423-1

+1 0.12 0.07 35.8 Surv.

1.10 39.2 26.9 8.5 56.4 Data18l Nozzle Belt to lntermed. Shell Circ Weld 8T1762 (SA-1426)

-47.6 0.19 0.57 167.0 Table 0.44 73.2 17.2 28 65.7 (100%)

Intermediate Shell Long Weld (ID 27%)

1 P0815 (SA-812)

-47.6 0.17 0.52 167.0 Table NIA NIA 17.2 28 NIA Intermediate Shell Long Weld (OD 73%)

1 P0661 (SA-775)

-47.6 0.17 0.64 167.0 Table 1.00 166.5 17.2 28 65.7 lntermed. To Lower Shell Circ. Weld (100%)

71249 (SA-1101)

-47.4 0.23 0.59 167.6 Table!Cl 1.09 182.3 12.9 28 61.7 Lower Shell Long Weld (100%)

61782 (SA-847)

-47.6 0.23 0.52 167.0 Table 0.99 164.5 17.2 28 65.7 Footnotes:

(A)

See Table 3.

(8)

Credible Surveillance Data; see BAW-2325 for evaluation.

(C)

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measured Li.RTNoT are predicted Li.RTNoT based on Table CF is less than 2cr (56°F).

(D)

Credible Surveillance Data; see WE Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistry Factor for Weld Wire Heat 61782, Point Beach Unit 1,"

utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.

(E)

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART= Initial RT NDT + Li.RT NOT + Margin, where Li.RT NoT = Chemistry Factor x Fluence Factor, and Margin= 2(cr1 2 + cr"2) 0*5, with cr1 defined as the standard deviation of the Initial RTNoT. and cro defined as the standard deviation of t.RTNDT*

(F)

Instruction Manual, 132-lnch 1.0. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

(G) cr1and er" values obtained from Areva Calculation 32-9019240-000, Table 41 (Ref. 5.12)

(H)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.

POINT BEACH TRM 2.2 -13 REV.11 12/2019 ART (oF)IEJ 117.4 145.9 96.6 91.3 NIA 184.6 196.6 182.7

POINT BEACH NUCLEAR PLANT TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 8 POINT BEACH UNIT 2 RPV 3/4T BEL TUNE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 50 EFPY<1J TRM 2.2 Unless otherwise noted, all ART input data obtained from WCAP-16669, Revision 1, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," (Ref 5.5).

Vessel Manufacturer:

Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding):

6.5", without clad!FJ Component Description Heat or Initial

%Cu

%Ni CF CF 3/4T11l Fluence ARTNDT O"l(G)

CY&(G)

Margin Heat/Lot RTNDT(°F)

Method Factor!AJ (OF\\

(oF)

No:Zzle Belt Forging 123V352

+40 0.11 0.73 76.0 Table 0.51 38.9 0

17.0 34.0 Intermediate Shell Forging 123V500

+40 0.09 0.70 58.0 Table!BJ 1.12 65.2 0

17.0 34.0 Lower Shell Forging 122W195

+40 0.05 0.72 42.8 Surv.

1.12 47.8 0

8.5 17.0 Data!CJ Nozzle Belt to lntermed. Shell Circ 21935

-56 0.18 0.70 170.5 Table!HJ 0.51 87.3 17.0 28 65.5 Weld (100%)

lntermed. to Lower Shell Circ. Weld 72442

-30 0.26 0.60 180.0 Table !DJ 1.10 198.4 11.9 28 60.8 (100%)

(SA-1484)

Footnotes:

(A)

See Table 4.

(B)

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ~RTNoT and predicted ~RTNoT based on Table CF is less than 2cr (56°F).

(C)

Credible surveillance data; see BAW-2325 for evaluation.

(D)

Non-credible surveillance data; Table CF value based on best-estimate chemistry is higher than best fit calculated using surveillance data, and therefore, conservative.

(E) Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART= Initial RT NOT + ~RT NoT + Margin, where ~RT NOT = Chemistry Factor x Fluence Factor, and Margin= 2(cri2 + crA2)0*5, with cr1 defined as the standard deviation of the Initial RTNoT. and cr" defined as the standard deviation of ~RTNoT (F)

Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970.

(G) cr1and crA values obtained from Areva Calculation 32-9019240-000, Table 41 (Ref. 5.12)

(H) Table CF value based on best-estimate chemistry data from GEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997

(!)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels.

POINT BEACH TRM 2.2 - 14 REV. 11 12/2019 ART (oF)!EJ 112.9 139.2 104.8 96.8 229.2