NRC 2008-0014, Pressure and Temperature Limits Report
| ML080640451 | |
| Person / Time | |
|---|---|
| Site: | Point Beach  |
| Issue date: | 02/27/2008 |
| From: | Mccarthy J Florida Power & Light Energy Point Beach |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| NRC 2008-0014 | |
| Download: ML080640451 (16) | |
Text
FPL Energy Point Beach, LLC, 6610 Nuclear Road, Two Rivers, WI 54241 FPL Energy, Point Beach Nuclear Plant February 27,2008 NRC 2008-0014 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 Section 5.6.5, FPL Energy Point Beach, LLC, enclosed is Revision 3 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
,hca B
ew
,$/ygre E
n LLC Site Vice president I Enclosure 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 FPL ENERGY POINT BEACH, LLC POINT BEACH NUCLEAR PLANT, UNITS I AND 2 PRESSURE TEMPERATURE LIMITS REPORT REVISION 3, ISSUED FEBRUARY 18,2008 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."
I.O 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. I 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 (Ref 5.15) this PTLR is effective until June 2008.
(Ref 5.19)
The Technical Specifications addressed in this report are listed below:
1.1 3.4.3 Pressure/Temperature (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. All 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 all applicable limits of the safety analysis are met. All items that appear in capitalized type are defined in Technical Specification 1.1, "Definitions."
All EFPY values listed in this procedure are estimates based on reactor power of 1518.5 MWt. Applicability of the operating limits are determined by accumulated fluence values listed in Tables 3 and 4. This report will be revised with new P-T limits prior to exceeding the associated fluence values.
2.1 RCS Pressure and Tem~eratureLimits (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 ~ 1 0 ° Fper 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 (includes instrument uncertainty).
POINT BEACH TRM REV. 3 February 18,2008
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 270°F (includes instrument uncertainty for RCS Tc wide range).
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 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 must 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)
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, RTNDTlwhich is determined in accordance with ASTM E208. The empirical relationship between RTNDTand 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 El85-82.
POINT BEACH TRM REV. 3 February 18,2008
POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 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 AND DATA TABLES 4.1 The RTPTsvalues for the PBNP limiting beltline materials is 278°F for Unit 1 and 291°F for Unit 2 at 32 EFPY.
4.2 Tables Table Number Table Description Table 1 Point Beach Nuclear Plant, Unit 1 Reactor Vessel Surveillance Capsule Removal Schedule Table 2 Point Beach Nuclear Plant, Unit 2 Reactor Vessel Surveillance Capsule Removal Schedule Table 3 Point Beach Unit 1 RPV Beltline 32.2 EFPY Fluence Values Table 4 Point Beach Unit 2 RPV Beltline 34.0 EFPY Fluence Values Table 5 Point Beach Unit 1 RPV 114t Beltline Material Adjusted Reference Temperatures at 32.2 EFPY Table 6 Point Beach Unit 2 RPV 114t Beltline Material Adjusted Reference Temperatures at 34.0 EFPY Table 7 Point Beach Unit 1 RPV 314t Beltline Material Adjusted Reference Temperatures at 32.2 EFPY Table 8 Point Beach Unit 2 RPV 314t Beltline Material Adjusted Reference Temperatures at 34.0 EFPY POINT BEACH TRM REV. 3 February 18,2008
POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT 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-12794, "Reactor Cavity Neutron Measurement Program for Point Beach Unit 1," Rev. 4, February 2000 5.3 WCAP-12795, "Reactor Cavity Neutron Measurement Program for Point Beach Unit 2," Rev. 3, August 1995 5.4 EPRl TR-107450, "P-T Calculator for Windows, Version 3.0," Revision 0, December 1998 5.5 Westinghouse Report, "Pressure Mitigating Systems Transient Analysis Results,"
July 1977 5.6 Westinghouse Report, "Supplement to the July 1977 Report, Pressure Mitigating Systems Transient Analysis Results," September 1977 5.7 Wisconsin Electric Calculation 2000-0001, Revision 0, RCS P-T Limits and LTOP Setpoints Applicable through 32.2 EFPY - Unit 1 and 34.0 EFPY - Unit 2 5.8 Deleted 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 License Amendment Request 251, dated December 14,2006 (NRC 2006-0090),
Technical Specification 5.6.5, Reactor Coolant System Pressure and I Temperature Limits (Application for use of FERRET Code as approved methodology for determining RCS pressure and temperature limits) 5.14 NRC SE dated 10118/07 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. 3 February 18,2008
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 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 Westing house report, "Low Pressure Overpressure Protection System (LTOPS)
Setpoint Analysis for Nuclear Management Company, Point Beach Units 1 and 2," January 2007. (This report was added as a basis for making the LTOP maximum setpoint change in Revision 2. The report references PT curves generated using Master Curve methodology, which has not been approved for use at PBNP. The lower LTOP setpoint is based upon limiting plant conditions which are not applicable to Master Curve methodology. See AR01116807.)
5.19 Root Cause Evaluation 01092944, "Apparent Non-compliance with TS 5 . 6 . 5 . ~ ~ "
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 46, Reactor Coolant Pump Operation 5.24 OP I A , Cold Shutdown to Hot Standby POINT BEACH TRM REV. 3 February 18,2008
POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT FIGURE 2 RCS PRESSURE-TEMPERATURE LIMITS FOR COOLDOWN POINT BEACH TRM REV. 3 February 18, 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 actual removal date will be adjusted depending on the implementation of a power uprate and operating history of Unit 2. (NRC SER dated 1212005, NUREG 1839)
POINT BEACH TRM REV. 3 February 18, 2008
POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 3 POINT BEACH UNIT 1 RPV BELTLINE 32.2 EFPY $ B ~ ~VALUES'~) ~.E~~.
Based on WCAP-12794, "Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 1," Rev. 4, February 2000. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimated fluence at 32 EFPY, due to changes in core design at certain points in the operating history of the unit. As intermediate input to further calculations, these values are not rounded in accordance with ASTM E29 (Ref. 11).
Vessel Manufacturer: I Babcock & Wilcox Plate and Weld Thickness (without cladding): 1 6.5, without clad fU' 32 EFPY 4 ~ e s t . ~ s t . 32.2 EFPY 4Best.Est. 32.2 EFPY 32.2 EFPY 32.2 EFPY 32.2 EFPY Inside Surface Inside Surface 4Best.Est. 114T 4~est.Est.114T +BI?s~.ES~.314T 46est.Est. 314T Component Description Heat or HeatlLot Fluence Fluence (El9 nlcm2) Fluence Fluence Fluence Fluence (El9 nlcm2) (A) (El 9 nlcm2) IS) Factor IC) (El9 nlcm2))(' Factor (')
Nozzle Belt Forging 122P237 0.547 0.550 0.3724 0.7269 0.1707 0.5322 Intermediate Shell Plate A9811-1 2.64 2.65 1.794 1.160 0.8225 0.9452 Lower Shell Plate (21423-1 2.24 2.25 1.523 1.116 0.6983 0.8993 Nozzle Belt to Intermed. Shell 8T1762 Circ Weld (100%) (SA-1426) 0.547 0.550 0.3724 0.7269 0.1707 0.5322 lntermediate Shell Long 1PO815 (SA-812) 1.74 1.75 1.185 1.047 NIA NIA Seam (ID 27%)
lntermediate Shell Long 1PO661 (SA-775) 1.74 1.75 NIA NIA 0.5431 0.8293 Seam (OD 73%)
Intermed. to Lower Shell Circ.
71249 (SA-l 2.24 2.25 1523 1.I16 0.6983 0.8993 Weld (100%)
Lower Shell Long Seam (100%) 61782 (SA-847) 1.54 1.55 1.049 1.013 0.481 1 0.7960 Footnotes:
Interpolationof neutron exposure (in units of E l 9 nlcm2, E>1 MeV) to a particular value of effective full power years (EFPY) is performed based on WCAP-12794, Revision 4. For example, for the nozzle belt forging, heat no. 122P237, fluence = 0.547 + 1 0.796 - 0.547 1x (32.2 EFPY - 32.0 EFPY) = 0.550 E l 9 n/cm2 (8 EFPY 32 EFPY)
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 where fSudis 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. 122P237, at 32.2 EFPY, at a depth of 114 of the 6 . 5 vessel wall (1.625"), f = 0.550 x e0.24('.625)
= 0.3724 E l 9 n/cm2.
The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulator Guide 1.99, Revision 2: ff = t@.28-0.'0'ogo , where f is the fluence in units of E l 9 n/cm2.
For example, the 32.2 EFPY 114T fluence factor for nozzle belt forging, heat no. 122P237, ff = 0.3724' . 8'0.'0 0.3724) = 0.7269.
Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969 (Ref. 12).
EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates.
POINT BEACH TRM REV. 3 February 18, 2008
POINT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 4 POINT BEACH UNIT 2 RPV BELTLINE 34.0 EFPY $Best.Est. VALUES'~)
Based on WCAP-12795, "Reactor Cavity Neutron Measurement Program for Wisconsin Electric Power Company Point Beach Unit 2," Rev.3, August 1995. Note that the estimated fluence at a specific point in time is not linearly interpolated between zero and the estimatedfluence at 32 EFPY, due to changes in core design at certain points in the operating history of the unit. As intermediate input to further calculations,these values are not rounded in accordance with ASTM E29 (Ref. 11).
Vessel Manufacturer: Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 6.5", without clad fU' 32 EFPY 34.0 EFPY 34.0 EFPY 34.0 EFPY 34.0 EFPY 34.0 EFPY best.Est. best.Est. $~est.Est. best.Est. 4~est.Est. &est.Est Heat or Component Description Inside Surface Inside Surface 114T 114T 314T 314T HeatlLot Fluence Fluence Fluence Fluence Fluence Fluence (El9 nlcm2) (El9 nlcm2) (El9 nlcm2)()' Factor IC) (El9 nlcm2)IS) Factor IC)
Nozzle Belt Forging 123V352 0.548 0.5775 0.3910 0.7399 0.1792 0.5435 Intermediate Shell Forging 123V500 3.01 3.174 2.149 1.208 0.9851 0.9958 Lower Shell Forging - 122W195 - 2.52 - 2.654 - 1.797 1.161 0.8237 0.9456 Nozzle Belt to Intermed. Shell 2 1935 0.548 0.5775 0.3910 0.7399 0.1792 0.5435 Circ Weld (100%)
Intermed. to Lower Shell Circ 72442 2.49 2.606 1.764 1.156 0.8088 0.9405 Weld (100%) (SA-1484)
Interpolationof neutron exposure (in unik of E l 9 n/cm2, MeV) to a particular value of effective full power years (EFPY) is performed based Revision 3. For example, for the nozzle belt forging, heat no. 123~352, fluence = 0.548 + J 0.784 - 0.548 1x (34 EFPY - 32 EFPY) = 0.5775 E l 9 n/cm2 (48 EFPY - 32 EFPY) fBJ 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.0.24x,where fSudis 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 34.0 EFPY, at a depth of 114 of the 6.5" vessel wall (1.625), f = 0.5775 x e*0.24(1.62"
= 0.3910 E l 9 nlcm2.
"' The dimensionless fluence factor is calculated using the fluence factor formula from equation (2) of Regulatory Guide 1 99 Revision 2: ff = ~0.28-0.10'og0 ,where f is the fluence in units of E l 9 n/cm2. For example, the 34.0 EFPY 114T fluence factor for noule belt forging, heat no. 123V352, ff = 0.3910~0.28-0'10'dg0.3910)
= 0.7399.
(nJ Instruction Manual, Reactor Vessel, Point Beach Nuclear Plant No. 2, Combustion Engineering, CE Book #4869, October 1970 (Ref. 13).
(EJ EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates.
POINT BEACH TRM REV. 3 February 18, 2008
POlNT BEACH NUCLEAR PLANT TRM 2.2 TECHNICAL REQUIREMENTS MANUAL PRESSURE TEMPERATURE LIMITS REPORT TABLE 5 POINT BEACH UNIT 1 RPV 114T BELTLINE MATERIAL ADJUSTED REFERENCE TEMPERATURES AT 32.2 EFPY h3est.Est.(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. 14),
including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All beltline materials are included for comparison.
I Babcock & Wilcox I
Vessel Manufacturer:
Plate and Weld Thickness (without cladding): 1 6 . 9 , without cladft)
Component Description I I I I I I I I I I Footnotes:
See Table 3 I 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 Calculation Addendum 98-0156-00-A, "Evaluation of New Surveillance Data on Chemistty Factor for Weld Wire Heat 61782, Point Beach Unit 1," (Ref.
- 15) 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 + ARTNDT+ Margin, where ARTNDT= Chemistry Factor x Fluence Factor, and Margin = 2(012+ 0?)0.5,with GI defined as the standard deviation of the lnitial RTNDT and oAdefined as the standard deviation of ARTNOT.For example, for nozzle belt forging, heat nO.122P237, ART = 50 + (77 x 0.7269) + 34 = 140°F. Calculated ARTvalues are rounded to the nearest O F in accordancs with the rounding-off method of ASTM Practice E29.
Instruction Manual, 132-Inch I.D. Reactor Pressure Vessel, Babcock & Wilcox, September 1969.
By inspection, these are the limiting material properties.
EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates POINT BEACH TRM REV. 3 February 18, 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 34.0 EFPY 4Best.Est."'
Unless othelwise noted, all ART input data obtained from BAW-2325, "Response to Request for Additional Information (RAI) Regarding Reactor Pressure Vessel Integrity,"
May 1998 (Ref. 14), including the most recent best-estimate chemistry values for welds, applying current B&EWOG mean-of-the-sou rces approach. All beltline materials are included for comparison.
Vessel Manufacturer: I Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): 1 6.5, without cladf')
Component Description Heat or HeatlLot 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 I
than 20 (34°F)
Credible surveillancedata; see BAW-2325 for evaluation.
Non-crediblesurveillance 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 ARTNDT = Chemistry Factor x Fluence Factor, and Margin = 2(tsI2+ o?)~.', with (TI defined as the standard deviation of the Initial RTNDT, and (TA defined as the standard deviation of ARTNDT.For example, for nozzle belt forging, heat no. 123V352,ART = 40 + (76 x 0.7399) + 34 = 130°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.
By inspection, these are the limiting material properties.
Table CF value based on best-estimate chemistry data from CEOG Report "Best Estimate Copper and NickelValues in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 (Ref. 6).
EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates.
POINT BEACH TRM REV. 3 February 18, 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 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, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All beltline materials are included for comparison.
Vessel Manufacturer: 1 Babcock & Wilcox Plate and Weld Thickness (without cladding): 1 6.5, without cladft' Footnotes:
I See Table 3. I
('
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 ARTNDT are predicted ARTND~ based on Table CF is less than 20 (56°F).
Credible Surveillance Data; see WE Calculation Addendum 98-0156-OOA, "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 Regulatoty Guide 1.99, Rev. 2. ART = Initial RTN, + ARTNDT + Margin, where ARTNDT = Chemistry Factor x Fluence Factor, and Marain = 21s: + 0?)0.5.with 01defined as the standard deviation of the Initial RTNDT, and (s, defined as the standard deviation of ARTNDT.For example, for nozzle belt forging, hearno. 1 2 2 ~ 2 3 7 =: 50 ~ ~+ ~(77 x 0.5322) + 34 = 125OF. 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.
(GJ By inspection, these are the limiting material properties.
(H) EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates POINT BEACH TRM REV. 3 February 18, 2008
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 34.0 EFPY $Best.Est.(')
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, including the most recent best-estimate chemistry values for welds, applying current B&WOG mean-of-the-sources approach. All beltline materials are included for comparison. '
Vessel Manufacturer: I Babcock & Wilcox and Combustion Engineering Plate and Weld Thickness (without cladding): ( 6.5",without clad")
Component Description Footnotes:
See Table 4. I Non-credible surveillance data; see BAW-2325 for evaluation. Table CF conservative because difference between measured ARTNOTand predicted ARTNOT based on Table CF is less than 20 (56OF).
("
Credible surveillance data; see BAW-2325 for evaluation.
fD)
Non-credible surveillance data; Table CF value based on best-estimatechemistry is higher than best fit calculated using surveillance data, and therefore, conservative.
fE)
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(0: + 0,3°5,with ol 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. 123V352, ART = 40 + (76 x 0.5435) + 34 = 115°F. Calculated ART values are rounded to the nearest OF 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) By inspection, these are the limiting material properties.
(") Table CF value based on best-estimate chemistry data from CEDG Report "Best Estimate Copper and Nickel Values in CE Fabricated Reactor Vessel Welds," CE NPSD-1039, Revision 2, Final Report, June 1997 fl EFPY value listed here is based on a reactor power of 1518.5 MWt. See Section 2.0, "Operating Limits," for discussion of applicability dates.
POINT BEACH TRM REV. 3 February 18, 2008