NRC 2008-0029, Pressure and Temperature Limits Report

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Pressure and Temperature Limits Report
ML081560718
Person / Time
Site: Point Beach  NextEra Energy icon.png
Issue date: 05/30/2008
From: Mccarthy J
Florida Power & Light Energy Point Beach
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
NRC 2008-0029
Download: ML081560718 (31)
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Text

FPL Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 FPL Energy Poii~tBeach Nuclear Plant May 30,2008 NRC 2008-0029 TS 5.6.5 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Point Beach Nuclear Plant, Units 1 and 2 Dockets 50-266 and 50-301 Renewed License Nos. DPR-24 and DPR-27 Pressure and Temperature Limits Report In accordance with Technical Specification 5.6.5, enclosed are Revision 4 and Revision 5 of the Pressure and Temperature Limits Report for Point Beach Nuclear Plant, Units 1 and 2.

This letter contains no new commitments.

Very truly yours, FPL E m n t Beach, LLC

/ Site Vice President I Enclosures cc: Administrator, Region Ill, USNRC Project Manager, Point Beach Nuclear Plant, USNRC Resident Inspector, Point Beach Nuclear Plant, USNRC PSCW An FPL Group company

ENCLOSURE 1 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 PRESSURE TEMPERATURE LIMITS REPORT REVISION 4, ISSUED APRIL 28,2008 14 pages follow

POINT BEACH NUCLEAR PLANT TRM 2.2 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 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, July 23, 2001, and October 18, 2007.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto. Based upon fluence values in Westinghouse report LTR-REA-04-64 ( ~ e5.15),

f this PTLR is effective for 36.9 EFPY (approximately 2015). (Ref 5.19)

The Technical Specifications addressed in this report are listed below:

1.I 3.4.3 Pressurerremperature (P-T) Limits 1.2 3.4.12 Low Temperature Overpressure Protection (LTOP) 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."

EFPY values listed in this procedure are estimates based on the following past and assumed future reactor power and fuel management strategy for Unit 1 (limiting vessel):

Reactor Fuel Management Strategy Reactor Power (MWt)

Low Leakage Without Hafnium Rods 1518.5 - startup to 02/03/2003 (Hafnium rods inserted from 5/89 1540.0 - 02/03/2003 to 1012008 (BOC 17) to 10108) 1678.0 - 10/2008 to EOLE Power operation outside of these future assumptions is acceptable. However, the effect on EFPY values would need to be evaluated.

2.1 RCS Pressure and Temperature Limits (LC0 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.
b. A maximum cooldown rate of 100°F in any one hour POINT BEACH TRM 2.2 -1 REV. 4 04/28/2008

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

c. An average temperature change of 510°F per hour during inservice leak and hydrostatic testing operations.

2.1.2 The RCS P-T limits for heatup and cooldown are specified by Figures 1 and 2, respectively.

2.1.3 The minimum temperature for pressurization or bolt up, using the methodology, is 60°F, which when corrected for possible instrument uncertainties is a minimum indicated RCS temperature of 78°F (as read on the RCS cold leg meter) or 70°F using the hand-held, digital pyrometer.

2.2 Low Temperature Overpressure Protection System Enable Temperature (LC0 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). (Ref 5.4)

I 2.3 Low Temperature Overpressure Protection System Setpoints (LC0 3.4.12)

Pressurizer Power-Operated Relief Valve Lift Setting Limits The lift setting for the pressurizer power-operated relief valves (PORVs) is 5420 psig (includes instrument uncertainty).

The following operating restrictions ensure continued operability of the LTOP system:

2.3.1 RCP Operating Restriction - No more than one RCP in operation for RCS temperature < I 80°F. (Ref 5.20 to 5.24) 2.3.2 Charging Pumps - Limit the number of operating charging pumps to two when LTOP is in service. (Ref 5.20 to 5.24) 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 I

capsules, shall be approved by the NRC prior to implementation. (Ref 5.16 and 5.17)

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, POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT RTNDT, which is determined in accordance with ASTM E208. The empirical relationship between RTNDT 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 E l 85-82.

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.

4.0 SUPPLEMENTAL DATA INFORMATION The RTPTsvalues for the PBNP limiting beltline materials at 36.9 EFPY is:

Unit 1 - Intermediate to Lower Shell Circ Weld = 277OF; Lower Shell Axial Weld = 234OF (Ref. 5.8, Table 13)

Unit 2 - lntermediate to Lower Shell Circ Weld = 292OF; lntermediate Shell Forging = 14g°F (Ref. 5.8, Table 21)

POINT BEACH TRM REV. 4 04/28/2008

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

5.0 REFERENCES

5.1 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision 2, January 1996 5.2 WCAP-15976, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," Revision 1, March 2008 5.3 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," 912211999 5.4 Westinghouse Letter WEP-08-25, "Transmittal of LTOPS Setpoint Evaluation,"

dated March 14, 2008 5.5 Deleted 5.6 BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998 5.7 CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 5.8 WCAP-16274-NP, "Evaluation of Pressurized Thermal Shock for Point Beach Units 1 and 2," Revision 0, June 2004 5.9 ASME B&PVC Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division 1" 5.10 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. 11 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. 12 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments RE: The Conversion to Improved Technical Specifications (TAC Nos. MA7186 and MA7187)", dated August 8,2001 5.13 Deleted 5.14 NRC SE dated October 18, 2007 issuing Amendment Nos. 2291234 to Facility Operating Licenses DPR-24 and DPR-27, (approving use of FERRET Code as approved methodology for determining RCS pressure and temperature limits)

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 5.15 Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluation Point Beach Units 1 and 2," dated June 2004 (Westinghouse Letter WEP-04-107) 5.16 Renewed Facility Operating License DPR-24, Point Beach Nuclear Plant Unit 1 5.17 Renewed Facility Operating License DPR-27, Point Beach Nuclear Plant Unit 2 5.18 Deleted 5.19 Root Cause Evaluation 01092944, "Apparent Non-compliance with TS 5.6.5.q" Corrective Action to Prevent Recurrence (CATPR) 2 Root Cause (RC)2.

5.20 CL 4C, Low Temperature Overpressurization Protection Unit 1 5.21 CL 4C, Low Temperature Overpressurization Protection Unit 2 5.22 OP 3C, Hot Standby to Cold Shutdown 5.23 OP 4B, Reactor Coolant Pump Operation 5.24 OP l A , Cold Shutdown to Hot Standby POINT BEACH TRM REV. 4 0412812008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Figure 1 RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP 100 150 200 250 300 350 400 Moderator Temperature ( O F )

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Figure 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN

- I . ---

nable Temperature = 285°F 50 100 150 200 250 300 350 400 Moderator Temperature ( O F )

POINT BEACH TRM REV. 4 04/28/2008

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 September 1972 (actual)

S December 1975 (actual)

R October 1977 (actual)

T March 1984 (actual)

P April 1994 (actual)

N Standby

  • The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.

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

V November 1974 (actual)

T March 1977 (actual)

R April 1979 (actual)

S October 1990 (actual)

P June 1997 (actual)

N Standby A April 2022**

  • The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.
    • The actual removal date will be adjusted depending on the implementation of a power uprate and operating history of Unit 2. (NRC SEdated 1212005, NUREG 1839)

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 3 POINT BEACH UNIT 1 RPV BELTLINE 36.9 EFPY VALUES'~'

Based on Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2," June 2004 (Ref 5.15).. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 36.9 EFPY, due to changes in core design at certain points in the operating history of the unit.

Vessel Manufacturer: I Babcock & Wilcox Plate and Weld Thickness (without cladding): 1 6.5", without clad 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY Component Description Heat or HeatlLot Inside Surface 114T Fluence 114T Fluence 314T Fluence 314T Fluence Fluence ( E l 9 nlcm2) (El9 nlcm2) fBJ Factor )(' ( E l 9 nlcm2) fBJ Factor ICJ Nozzle Belt Forging 122P237 0.25 0.17 0.53 0.08 0.37 Intermediate Shell Plate A981 1-1 3.38 2.29 1.22 1.05 1.01 Lower Shell Plate C1423-1 3.04 2.06 1.20 0.94 0 98 Nozzle Belt to Intermed. Shell 8T1762 (SA- 1426) 0.25 0.17 0.53 0.08 0.37 Circ Weld (100%)

lntermediate Shell Long 1PO815 (SA-8 12) 2.19 1.48 1.11 0.68 NIA Seam (ID 27%)

lntermediate Shell Long 'A' 1PO661 (SA-775) 2.19 1.48 NIA 0.68 0.89 Seam (OD 73%)

Intermed. to Lower Shell Circ.

71249 ~SA-,,lO1) 3.05 2.07 1.20 0.95 0.99 Weld (100%)

Lower Shell Long Seam 'A' 6 1782 (SA-847) 2.08 1.41 1.10 0.65 0.88 (100%)

Footnotes:

' Limiting material I

lB' 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 = fs,# x e-OZ4',where fs,# is expressed in units of E l 9 nlcm2, E>1 MeV, and x is the desired depth in inches into the vessel wall. For example, for the nozzle belt forging, heat no. 122P237, at 36.9 EFPY, at a depth of 114 of the 6.5" vessel wall (1.625"), f = 0.25 x e424(1.625)

- 0.17 E l 9 n/un2.

lC' The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) ~ f ( ~ g $ ~ ~ M ~ ~1.99,

$ ~ i Revision c!e 2: ff = f'0.28-010'ogo

, where f is the fluence in units of E l 9 n/cm2.

For example, the 36.9 EFPY 1/4T fluence factor for nozzle belt forging, heat no. 122P237, ff = 0.17 - 0.53.

lD' Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969 .

lE' EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See Section 2.0, "Operating Limits," for discussion of EFPY values.

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 POINT BEACH UNIT 2 RPV BELTLINE 36.9 EFPY VALUES'~)

Based on Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2," June 2 0 0 4 ( ~ e5.15). f Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 33.9 EFPY, due to changes in core design at certain points in the operating history of the unit.

Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5", without clad 'O) 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY Heat or Inside Surface Component Description Fluence 114T Fluence 314T Fluence 314T Fluence HeatlLot Fluence (El9 nlcm2) IS) Factor IC) (El9 nlcm2) le.' Factor IC)

(El9 nlcm2)

Nozzle Belt Forging 123V352 0.34 0.23 0.60 0.11 0.44 Intermediate Shell Forging 'A' 123V500 3.38 2.29 1.22 1.05 1.01 Lower Shell Forging 122W195 3.30 2.23 1.22 1.02 1.01 Nozzle Belt to Intermed. Shell 21935 Circ Weld (100%) 0.34 0.23 0.60 0.1 1 0.44 Intermed. to Lower Shell Circ 72442 I Weld I 100%) 'A' I ISA-1484) I 3.13 2.12 1.20 0.97 , 0.99 ,

Footnotes:

lAJ Limiting Material

' 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 = fSudx e.OZ4",where fSu6is expressed in units of E l 9 n/cm2,E>1 MeV, and x is the desired depth in inches into the vessel wall. For example, for the nozzle belt forging, heat no. 123V352, at 36.9 EFPY, at a depth of 114 of the 6 . 5 vessel wall (1.625"), f = 0.34 x e-024'1.625)

= 0.23 E l 9 n/cm2.

(

The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99, Revision 2: ff = f" 2 8 - 0 1 0 ' o g n , where f is the fluence in units of E l 9 n/cm2. For example, the 36.9 EFPY 114T fluence factor for nozzle belt forging, heat no. 123V352, ff = 0 . 2 3 ' ~ ~= 0.60. ~ ~ ~ ~ ~ ~ ' ~ ~ ~ ~ ~ ~ ~ )

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

'" EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See Section 2.0, "Operating Limits," for discussion of EFPY values POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 5 POINT BEACH UNIT 1 RPV 1/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY.'~)

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998. (Ref. 5.6)

Vessel Manufacturer: I Babcock & Wilcox Plate and Weld Thickness (without cladding): 1 6 . 5 , without clad")

tootnotes:

See Table 3 Credible Surveillance Data; see BAW-2325 for evaluation.

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measure ARTNDTand predicted ARTNDT based on Table CF is less than 20 (56°F).

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

(Ref.5.3) utilizing latest time-weighted temperature data for Point Beach Unit 1, which supersedes BAW-2325.

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT + Margin, where ARTNDT = Chemistry Factor x Fluence Factor, and Margin = 2(oI2+ with 01defined as the standard deviation of the Initial RTNDT and 0, defined as the standard deviation of ARTNDT.For example, for nozzle belt forging, heat n0.122P237, ART = 50 + (77 x 0.53) + 34 = 125°F. Calculated ART values are rounded to the nearest O F in acuxdance w~ththe rounding-off method of ASTM Pradlce E29 Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

Deleted.

EFPY value listed here is based on various reador fuel management strategies and reactor power levels. See Section 2.0, "Operating Limits," for discussion of EFPY values.

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 6 POINT BEACH UNIT 2 RPV 114T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY ,"

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"

May 1998. (Ref. 5.6).

I Vessel Manufacturer: I Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 1 6.5", without cladfF' Footnotes:

"' See Table 4 Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNDT and predicted ARTNDT based on Table CF is less

"'"' than 20 (34°F)

Credible surveillance data; see BAW-2325 for evaluation.

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

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT + Margin, where ARTNOT = Chemistry Factor x Fluence Factor, and Margin = 2(o12+ o:)05, with ol defined as the standard deviation of the Initial RTNDT, and o,,defined as the standard deviation of ART., For example, for nozzle belt forging, heat no. 123V352, ART = 40 + (76 x 0.60) + 34 = 120°F. Calculated ART values are rounded to the nearest OF in accordance with the rounding-off method of ASTM Practice E29.

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

'G' Deleted.

" 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.5.7).

'IJ EFPY value listed here is based on various reador fuel management strategies and reactor power levels. See Section 2.0, "Operating Limits," for discussion of EFPY values.

POINT BEACH TRM REV. 4 0412812008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 7 POINT BEACH UNIT 1 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY" Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"

May 1998. (Ref 5.6)

Vessel Manufacturer: I Babcock & Wilcox Plate and Weld Thickness (without cladding): 1 6.5", without clad")

314T 36.9 EFPY Initial CF ARTNDT Margin ART C o m p o n e n t Description Heat o r HeatlLot XCu %Ni CF Fluence RTNDT(OF) Method actor'^' (OF)

O' OA

("F) OF)'^'

surv.

163.3 14 48.34 187

~ ~ ~ ~ l D l 143.7 Footnotes:

'" See Table 3.

'" Credible Surveillance Data; see BAW-2325 for evaluation.

-Nan-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measured ARTNDTare predicted ARTNDT based on Table CF is less than 20 (56°F).

ID' 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 I , which supersedes BAW-2325.

'E)

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT+ ARTNOT + Margin, where ARTNOT = Chemistry Factor x Fluence Factor, and Margin = 2(aI2+ 0:)05, with 0, defined as the standard deviation of the Initial RTNDT, and ob defined as the standard deviation of ARTNOT.For example, for nozzle belt forging, heat no. 122P237, ART = 50 + (77 x 0.37) + 34 = 113°F. CalculatedART values are rounded to the nearest OF in accordance with the rounding-off method of ASTM Practice E29.

'F)

Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

'G) Deleted.

fH)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See Sedion 2.0, "Operating Limits," for discussion of EFPY values.

POINT BEACH TRM REV. 4 0412812008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 8 POINT BEACH UNIT 2 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY Unless o t h e ~ l i s enoted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"

May 1998. (Ref 5.6)

Vessel Manufacturer: I Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 1 6.5", without clad")

Footnotes:

m, See Table 4.

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNDT and predicted ARTNOT based on Table CF is less than 20 (56'F).

lC' 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.

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT + Margin, where ARTNOT= Chemistry Factor x Fluence Factor, and Margin = 2(oI2 + 0,2), with IS, defined as the standard deviation of the Initial RTNDT, and 0~defined as the standard deviation of ARTNDT.For example, for nozzle belt forging, heat no. 123V352,ART = 40 + (76 x 0.44) + 34 = 107°F. Calculated ART values are rounded to the nearest "F in accordance with the rounding-off method of ASTM Practice E29.

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

fG) Deleted.

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

' EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See Section 2.0, "Operating Limits," for discussion of applicability dates.

POINT BEACH TRM REV. 4 0412812008

ENCLOSURE 2 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS 1 AND 2 PRESSURE TEMPERATURE LIMITS REPORT REVISION 5, ISSUED MAY 30,2008 14 pages follow

POINT BEACH NUCLEAR PLANT TRM 2.2 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 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, July 23, 2001, and October 18, 2007.

The PTLR shall be provided to the NRC upon issuance for each reactor vessel fluence period and for any revision or supplement thereto (Ref 5.19). Based upon fluence values in I Westinghouse report LTR-REA-04-64 (Ref 5.15),this PTLR is effective for 36.9 EFPY (approximately 2015). (Ref 5.2) I The Technical Specifications addressed in this report are listed below:

1.1 3.4.3 Pressurerremperature (P-T) Limits 1.2 3.4.12 Low Temperature Overpressure Protection (LTOP) 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. I ,

"Definitions."

I 2.1 RCS Pressure and Temperature Limits (LC0 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
b. A maximum cooldown rate of 100°F in any one hour.
c. An average temperature change of 110°F per hour during inservice leak and hydrostatic testing operations.

2.1.2 The RCS P-T limits for heatup and cooldown are specified by Figures 1 and 2, respectively. (Ref 5.2)

POINT BEACH TRM REV. 5 05/30/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 2.1.3 The minimum temperature for pressurization or bolt up, using the methodology, is 60°F, which when corrected for possible instrument uncertainties is a minimum indicated RCS temperature of 78°F (as read on the RCS cold leg meter) or 70°F using the hand-held, digital pyrometer.

2.2 Low Temperature Overpressure Protection Svstem Enable Temperature (LC0 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 T, wide range). (Ref 5.4) 2.3 Low Temperature Overpressure Protection Svstem Setpoints (LC0 3.4.12)

Pressurizer Power-Operated Relief Valve Lift Setting Limits The lift setting for the pressurizer power-operated relief valves (PORVs) is 1420 psig (includes instrument uncertainty).

The following operating restrictions ensure continued operability of the LTOP system:

2.3.1 RCP Operating Restriction - No more than one RCP in operation for RCS temperature < I 80°F. (Ref 5.20to 5.24) 2.3.2 Charging Pumps - Limit the number of operating charging pumps to two when LTOP is in service. (Ref 5.20 to 5.24) 2.4 Criticalitv and Hvdrostatic 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 5.2) 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 5.16 and 5.17)

POINT BEACH TRM REV. 5 05/30/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 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, RTNDT, which is determined in accordance with ASTM E208. The empirical relationship between R T N Dand~ 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 E l 85-82.

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.

4.0 SUPPLEMENTAL DATA INFORMATION The RTPTsvalues for the PBNP limiting beltline materials at 36.9 EFPY is:

Unit 1 - lntermediate to Lower Shell Circ Weld = 277OF; Lower Shell Axial Weld = 234OF (Ref. 5.8, Table 13)

Unit 2 - lntermediate to Lower Shell Circ Weld = 292OF; lntermediate Shell Forging = 14g°F (Ref. 5.8, Table 21)

POINT BEACH TRM REV. 5 05/30/2008

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

5.0 REFERENCES

5.1 WCAP-14040-NP-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision 2, January 1996 5.2 WCAP-15976, "Point Beach Units 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," Revision 1, March 2008 5.3 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," 912211999 5.4 Westinghouse Letter WEP-08-25, "Transmittal of LTOPS Setpoint Evaluation,"

dated March 14, 2008 5.5 PWR Owner Group Topical Report BAW-1543(NP), Revision 4, Supplement 6-A, "Supplement to the Master Integrated Reactor Vessel Surveillance Program" (TAC No. MC9608), June 2007 5.6 BAW-2325, "Response to Request for Additional Information ( M I ) Regarding Reactor Pressure Vessel Integrity," May 1998 5.7 CEOG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 5.8 WCAP-16274-NP, "Evaluation of Pressurized Thermal Shock for Point Beach Units 1 and 2," Revision 0, June 2004 5.9 ASME B&PVC Code Case N-641, "Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements,Section XI, Division 1" 5.10 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. 11 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. 12 NRC Letter, "Point Beach Nuclear Plant, Units 1 and 2 - Issuance of Amendments RE: The Conversion to Improved Technical Specifications (TAC Nos. MA7186 and MA7187)", dated August 8,2001 5.13 Deleted 5.14 NRC SE dated October 18, 2007 issuing Amendment Nos. 2291234 to Facility Operating Licenses DPR-24 and DPR-27, (approving use of FERRET Code as approved methodology for determining RCS pressure and temperature limits)

POINT BEACH TRM REV. 5 0513012008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 5.15 Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluation Point Beach Units 1 and 2," dated June 2004 (Westinghouse Letter WEP-04-107) 5.16 Renewed Facility Operating License DPR-24, Point Beach Nuclear Plant Unit 1 5.17 Renewed Facility Operating License DPR-27, Point Beach Nuclear Plant Unit 2 5.18 Deleted 5.19 Root Cause Evaluation 01092944, "Apparent Non-compliance with TS 5.6.5.c,"

Corrective Action to Prevent Recurrence (CATPR) 2 Root Cause (RC)2.

5.20 CL 4C, Low Temperature Overpressurization Protection Unit 1 5.21 CL 4C, Low Temperature Overpressurization Protection Unit 2 5.22 OP 3C, Hot Standby to Cold Shutdown 5.23 OP 4B, Reactor Coolant Pump Operation 5.24 OP I A , Cold Shutdown to Hot Standby POINT BEACH TRM REV. 5 05/30/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Figure 1 RCS PRESSURE-TEMPERATURE LIMITS FOR HEATUP Maximum LTOP Setpoint = 420 psig with RCP restriction Moderator Temperature (OF)

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT Figure 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN Moderator Temperature ( O F )

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 1 POINT BEACH NUCLEAR PLANT UNlT 1 REACTOR VESSEL SURVEILLANCE CAPSULE REMOVAL SCHEDULE Capsule Identification Letter Approximate Removal Date*

V September 1972 (actual)

S December 1975 (actual)

R October 1977 (actual)

T March 1984 (actual)

P April 1994 (actual)

N Standby

  • The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.

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

V November 1974 (actual)

T March 1977 (actual)

R April 1979 (actual)

S October 1990 (actual)

P June 1997 (actual)

N Standby A April 2022**

  • The actual removal dates will be adjusted to coincide with the closest scheduled plant refueling outage or major reactor plant shutdown.
    • The withdraw schedule for Capsule A is shown in PWROG BAW-1543 (NP) (Ref. 5.5). The I

actual removal date will be adjusted depending on the implementation of a power uprate and operating history of Unit 2. (NRC SE dated 1212005, NUREG 1839)

POINT BEACH TRM REV. 4 0412812008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 3 POINT BEACH UNIT 1 RPV BELTLINE 36.9 EFPY VALUES(~)

Based on Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2," June 2004 (Ref 5.15).. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 36.9 EFPY, due to changes in core design at certain points in the operating I Babcock & Wilcox Vessel Manufacturer:

Plate and Weld Thickness (without claddina): 1 6.5", without clad I 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY Component Description Heat o r HeaffLot Inside Surface 114T Fluence 114T Fluence 314T Fluence 314T Fluence Fluence (El9 nlcm2) (El9 nlcm2) )(' Factor )(' (El9 nlcm2) )(' Factor )('

Nozzle Belt Forging 122P237 0.25 0.17 0.53 0.08 0.37 Intermediate Shell Plate A981 1-1 3.38 2.29 1.22 1.05 1.01 Lower Shell Plate C1423-1 3.04 2.06 1.20 0.94 0.98 Nozzle Belt to Intermed. Shell 8T1762 0.25 0.17 0.53 0.08 0.37 Circ Weld (100%) (SA-1426)

Intermediate Shell Long 1.48 1.I 1 1P0815 (SA-812) 2.19 0.68 N/A Seam (ID 27%)

Intermediate Shell Long (A) 1PO661 (SA-775) 2.19 1.48 NIA 0.68 0.89 Seam (OD 73%)

Intermed. to Lower Shell Circ.

71249 (SA-l 01) 3.05 2.07 1.20 0.95 0.99 Weld (100%)

Lower Shell Long Seam (A' 1.10 61782 (SA-847) 2.08 141 0.65 0.88 (100%)

Footnotes:

(*' Limiting material

)(' 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 = f,* x e-0.24x, where d ,f is expressed in units of E l 9 nlcm2, E>1 MeV, and x is the desired depth in inches into the vessel wall. For example, for the nozzle belt forging, heat no. 122P237, at 36.9 EFPY, at a depth of 114 of the 6.5" vessel wall (1.625"), f = 0.25 x e-024'1.625)

= 0.17 E l 9 n/cmZ.

)('

The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatoy Guide 1.99, Revision 2: ff = f'028-010'og1) , where f is the fluence in units of E l 9 n/cm2.

For example, the 36.9 EFPY 114T fluence factor for nozzle belt forging, heat no. 122P237, ff = 0.17'028~0.10'og 3724) = 0.53.

fD)

Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969 .

fE)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP-15976 (Ref .52)11for discussion of EFPY values. I POlNT BEACH TRM REV. 4 0412812008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 POINT BEACH UNIT 2 RPV BELTLINE 36.9 EFPY VALUES'~)

Based on Westinghouse Report LTR-REA-04-64, "Pressure Vessel Neutron Exposure Evaluations Point Beach Units 1 and 2," June 2004 (Ref 5.15). Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 36.9 EFPY, due to changes in core design at certain points in the operating history of the unit.

Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5", without clad 'D) 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY 36.9 EFPY Heat or Inside Surface Component Description Fluence 114T Fluence 314T Fluence 314T Fluence HeatILot Fluence

( E l 9 nlcm2) )(' Factor ") (El9 nlcm2))(' Factor )('

(El9 nlcm2)

Nozzle Belt Forging 123V352 0.34 0.23 0.60 0.1 1 0.44 Intermediate Shell Forging 'A) 123V500 3.38 2.29 1.22 1.05 1.O1 Lower Shell Forging 122W195 3.30 2.23 1.22 1.02 1.O1 Nozzle Belt to Intermed. Shell 21935 0.34 0.23 0.60 0.1 1 0.44 Circ Weld (100%)

Intermed. to Lower Shell Circ 72442 3.13 2.12 1.20 0.97 0.99 Weld (100%) (A' (SA-1484)

Footnotes:

'*) Limiting Material

("

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 = fSufix e-0.24x, where f,,~ is expressed in units of E l 9 n/cm2, E>1 MeV, and x is the desired depth in inches into the vessel wall. For example, for the nozzle belt forging, heat no. 123V352, at 36.9 EFPY, at a depth of 114 of the 6 . 5 vessel wall (1.625"), f = 0.34 x e-024'1.625)

= 0.23 E l 9 n/cm2.

("

The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1.99 Revision 2: ff = f'028-010'q)f) , where f is the fluence in units of E l 9 n/crn2. For example, the 36.9 EFPY 1/4T fluence factor for noule belt forging, heat no. 123V352, ff = 0.23'~'~-O10'q)05910) = 0.60.

(D)

Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book M869, October 1970

'E)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP-15976 (Ref 5.2) for discussion of EFPY values.

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 5 POINT BEACH UNIT 1 RPV 1/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY."

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity," May 1998. (Ref. 5.6)

Vessel Manufacturer: I Babcock & Wilcox Plate and Weld Thickness (without cladding): 1 6.5",without clad")

Initial Heat or Component Description HeatlLot R T (OF)

~ %Cu

~ ~ %INi CF CF Method 114T 36.9 EFPY Fluence ~ ~ d (OF)

ARTNDT

~ ~ ("~ )

OA Margin (OF)

ART (OF)"

163.3 Surv.

14 48.34 223 179.6 Footnotes:

(A' See Table 3

('j Credible Surveillance Data; see BAW-2325 for evaluation.

("

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adjusted measure ARTNOT and predicted ARTNOT based on Table

'" CF is less than 20 (56°F).

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

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNOT + Margin, where ARTNOT= Chemistry Factor x Fluence Factor, and Margin = 2(0? + with 0, defined as the standard deviation of the Initial RTNDT and 0, defined as the standard deviation of ARTNOT.For example, for nozzle belt forging, heat no. 122P237, ART = 50 + (77 x 0.53) + 34 = 125°F. Calculated ART values are rounded to the nearest "F in accordance with the rounding-off method of ASTM Practice E29.

(Fj Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

(G' Deleted.

" EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP-15976 (Ref 5.2) for discussion of EFPY values.

POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 6 POINT BEACH UNIT 2 RPV 114T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY .(I)

Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"

May 1998. (Ref. 5.6).

Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5",without clad" Component Description Footnotes:

7n, See Table 4

(

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNDTand predicted ARTNDTbased on Table CF is less than 20 (34°F)

(

Credible surveillance data; see BAW-2325 for evaluation.

("

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

fEJ Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT + ARTNDT+ Margin, where ARTNDT= Chemistry Factor x Fluence Factor, and Margin = 2(0? + 0 2 ) ~ with

~ . ol defined as the standard deviation of the Initial RTNDT,and CTA defined as the standard deviation of A R T N ~ .For example, for nozzle belt forging, heat no. 123V352, ART = 40 + (76 x 0.60) + 34 = 120°F. Calculated ART values are rounded to the nearest OF in accordance with the rounding-off method of ASTM Practice E29.

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

fG' Deleted.

(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.5.7).

(IJ EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP-15976 (Ref 5.2) for discussion of EFPY values. I POINT BEACH TRM REV. 4 04/28/2008

POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 7 POINT BEACH UNIT 1 RPV 314T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY 'H' Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"

May 1998. (Ref 5.6) 1 Vessel Manufacturer: I Babcock & Wilcox I 1 Plate and Weld Thickness (without cladding): 1 6.5", without clad'"

I Component Description Lower Shell Long Seam (100%) 61782 (SA-847) -5 0.23 0.52 157.4 Table 0.88 19.7 28 68.47 202 138.5 Surv.

163.3 14 48.34 187 D ~ ~ ~ ( D ' 143.7 L Footnotes:

See Table 3.

'('

Credible Surveillance Data; see BAW-2325 for evaluation.

'(' -Nan-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between ratio-adiusted measured ARTN~T . - are predicted ARTN~T based on Table CF is less than 20 (56°F).

fD' Credible Surveillance Data: see WE Calculation Addendum 98-0156-00-A. "Evaluation of New Surveillance Data on Chemistrv 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.

)('

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial RTNDT+ ARTNOT + Margin, where ARTNDT= Chemistry Factor x Fluence Factor, and Margin = 2(012 + o:)~',with 01defined as the standard deviation of the Initial RTNDT,and 0, defined as the standard deviation of ARTNDT.For example, for nozzle belt forging, heat no. 122P237, ART = 50 + (77 x 0.37) + 34 = 113°F. Calculated ART values are rounded to the nearest OF in accordance with the rounding-off method of ASTM Practice E29.

(F)

Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.

(G' Deleted.

(H)

EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP-15976 (Ref 5.2) for discussion of EFPY values.

POINT BEACH TRM 2.2 - 13 REV. 4 0412812008

fl POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 8 POINT BEACH UNIT 2 RPV 3/4T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 36.9 EFPY ' I Unless otherwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"

May 1998. (Ref 5.6)

Vessel Manufacturer: I Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 1 6.5", without clad")

Footnotes:

lA' See Table 4.

)('

Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNoTand predicted ARTNDTbased on Table CF is less than 20 (56°F).

(" Credible surveillance data; see BAW-2325 for evaluation.

fD)

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

()

Adjusted reference temperature (ART) calculated per Regulatory Guide 1.99, Rev. 2. ART = Initial R T N ~+TARTNOT+ Margin, where ARTNDT= Chemistry Factor x Fluence Factor, and Margin = 2(oI2+ 0:)O5, with 01defined as the standard deviation of the Initial RTNDT, and 0, defined as the standard deviation of ARTNOT.For example, for nozzle belt forging, heat no. 123V352, ART = 40 + (76 x 0.44) + 34 = 107°F. Calculated ART values are rounded to the nearest "F in accordance with the rounding-off method of ASTM Practice E29.

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

fG) Deleted.

" 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 EFPY value listed here is based on various reactor fuel management strategies and reactor power levels. See WCAP-15976 (Ref 5.2) for discussion of EFPY values. I POINT BEACH TRM REV. 4 04/28/2008

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