ML050610038
| ML050610038 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 02/22/2005 |
| From: | Polson K Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW050018, TAC MC3283, TAC MC3284, TAC MC3285, TAC MC3286 | |
| Download: ML050610038 (50) | |
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Exelon Generation Company, LLC www.exeloncorp.com Nuclear Braidwood Station 35100 South Rt 53, Suite 84 Braceville, IL 60407-9619 Tel. 815-417-2000 February 22, 2005 BW050018 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and 50-457
Subject:
Corrected Pressure and Temperature Limits Reports (PTLRs),
Revision 3, Braidwood Station, Units 1 and 2
References:
(1) Letter from Kenneth A. Ainger (Exelon Generation Company, LLC) to NRC, "Request for a License Amendment to Incorporate Approved Pressure and Temperature Limits Report (PTLR) Methodology into Technical Specifications,"
dated May 21, 2004 (2) Letter from U.S. NRC to Christopher M. Crane, "Issuance of Amendments:
Revised Pressure-Temperature Limits Methodology; Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2 (TAC Nos. MC3285, MC3286, MC3283, MC3284), dated October 4, 2004 (3) Letter from Keith J. Poison to NRC, "Pressure and Temperature Limits Reports (PTLRs), Revision 3, Braidwood Station, Units 1 and 2," dated January 24, 2005 Copies of recently implemented revisions to the Braidwood Station, Units 1 and 2 Pressure and Temperature Limits Reports (PTLRs) were sent to the NRC by letter dated January 24, 2005 (Reference 3). The revised PTLRs were transmitted to the NRC in accordance with Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)" and as requested in Reference 2. This revision of the PTLRs was recently implemented and extended the current pressure-temperature (P-T) limits curves by an additional 2 effective full power years (EFPY) as described in Reference 1. However, it was subsequently identified that a pagination error existed in the Braidwood Station, Unit 2 PTLR transmitted to the NRC in Reference 3.
- Ag I
U.S. Nuclear Regulatory Commission Page 2 February 22, 2005 Therefore, Exelon Generation Company, LLC (EGC) is resubmitting the Braidwood Station, Unit 1 PTLR in Attachment 1 and providing the corrected Braidwood Station, Unit 2 PTLR in .
EGC apologizes for any inconvenience this administrative oversight may have caused. Should you have any questions regarding this matter, please contact Mr. Dale Ambler, Regulatory Assurance Manager, at (815) 417-2800.
Sincerely, Keith JPolson Site Vice President Braidwood Station Attachments: 1. Braidwood Unit 1 Pressure Temperature Limits Report, Revision 3
- 2. Braidwood Unit 2 Pressure Temperature Limits Report, Revision 3 (corrected) cc: Regional Administrator - NRC Region III NRC Senior Resident Inspector- Braidwood Station
BRAIDWOOD UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
Revision 3
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 Operating Limits 1 2.1 RCS Pressure and Temperature (P/T) Limits 1 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints 2 2.3 LTOP Enable Temperature 2 2.4 Reactor Vessel Boltup Temperature 3 2.5 Reactor Vessel Minimum Pressurization Temperature 3 3.0 Reactor Vessel Material Surveillance Program 11 4.0 Supplemental Data Tables 13 5.0 References 20 i
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup 4 Rates up to 1000 F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations 5 (Cooldown Rates of 0, 25, 50, and 100 'F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors) 2.3 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 9 Overpressure Protection (LTOP) System Applicable for the First 16 EFPY ii
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 a Braidwood Unit 1 Heatup Data Points at 16 EFPY (Without 6 Margins for Instrumentation Errors) 2.1b Braidwood Unit 1 Cooldown Data Points at 16 EFPY (Without 8 Margins for Instrumentation Errors) 2.2 Data Points for Braidwood Unit I Nominal PORV Setpoints for 10 the LTOP System Applicable for the First 16 EFPY 3.1 Braidwood Unit I Capsule Withdrawal Schedule 12 4.1 Braidwood Unit 1 Calculation of Chemistry Factors Using 14 Surveillance Capsule Data 4.2 Braidwood Unit 1 Reactor Vessel Material Properties 15 4.3 Summary of Braidwood Unit 1 Adjusted Reference Temperatures 16 (ARTs) at the 1/4T and 3/4T Locations for 16 EFPY 4.4 Braidwood Unit I Calculation of Adjusted Reference 17 Temperatures (ARTs) at 16 EFPY at the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data) 4.5 RTpTs Calculation for Braidwood Unit I Beltline Region 18 Materials at EOL (32 EFPY) 4.6 RTpTs Calculation for Braidwood Unit I Beltline Region 19 Materials at Life Extension (48 EFPY) iii
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 1 has been prepared in accordance with the requirements of Braidwood Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)".
Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications (TS) addressed in this report are listed below:
LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 Operating Limits The PTLR limits for Braidwood Unit I were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A (Reference
- 1) was used with the following exceptions:
a) Use of ENDF/B-IV neutron transport cross-section library and ENDF/B-V dosimeter reaction cross-sections, b) Use of ASME Code Case N-514, and c) Use of RELAP computer code for calculation of LTOP setpoints for Braidwood Unit I replacement steam generators.
These exceptions to the methodology in WCAP 14040-NP-A have been reviewed and accepted by the NRC in Reference 2.
WCAP 14243, Reference 3, provides the basis for the Braidwood Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. Reference 4 evaluated the effect of higher fluence from 5% uprate on the existing P/T curves.
The applicability periods for all areas previously evaluated for 14.0 EFPY have been extended by two additional years to 16.0 EFPY. This applicability period extension was reviewed and approved by the NRC in Reference 12.
2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3).
2.1.1 The RCS temperature rate-of-change limits defined in Reference 3 are:
- a. A maximum heatup of 100lF in any l-hour period,
- b. A maximum cooldown of 100lF in any l-hourperiod, and l
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT
- c. A maximum temperature change of less than or equal to 10F in any l-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1 a. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1b. These limits are defined in Reference 3. Consistent with the methodology described in Reference I and exceptions noted in Section 2.0, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error.
The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.
The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.
2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12).
The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 2.3 and Table 2.2. These limits are based on References 5, 6, and 7. The Residual Heat Removal (RH) Suction Relief Valves are also analyzed to individually provide low temperature overpressure protection. This analysis for the RH Suction Relief Valves remains valid with the current Appendix G limits contained in this PTLR document and will be reevaluated in the future as the Appendix G limits are revised.
The LTOP setpoints are based on P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1. The LTOP PORV nominal lift settings shown in Figure 2.3 and Table 2.2 account for appropriate instrument error.
2.3 LTOP Enable Temperature The minimum required LTOP enable temperature is 2000 F (Reference 2).
Braidwood Unit I procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 3507F and below and disarming of LTOP for RCS temperature above 350'F.
2
- a BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Note that the last LTOP PORV segment in Table 2.2 extends to 450'F where the pressure setpoint is 2350 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.
2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification)
The minimum boltup temperature for the Reactor Vessel Flange shall be 2 60'F.
Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.
2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification)
The minimum temperature at which the Reactor Vessel may be pressurized (i.e., in an unvented condition) shall be 2 60'F, plus an allowance for the uncertainty of the temperature instrument, determined using a technique consistent with ISA-S67.04-1994.
3
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: WELD METAL LIMITING ART VALUES AT 16 EFPY: 1/4T, 76.60 F 3/4T. 65.40 F m 1600 rn
-a
, 1400 U)
(n W
C- 1200 a)
- s 1000 co is C,
0 50 100 150 200 250 300 350 400 450 500 ModeratorTemperature (deg. F)
Figure 2.1 Braidwood Unit 1 Reactor Coolant System Heatup Limitations (heatup rate up to 1001F/hr)
Applicable for the First 16 EFPY I (Without Margins for Instrumentation Errors) 4
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 7
MATERIAL PROPERTY BASIS LIMITING MATERIAL: WELD METAL LIMITING ART VALUES AT 16 EFPY: I14T, 76.61F 3/4T, 65.40 F 2400 2200 2000 1800 c& 1600 0.
2 1400 a,
a, 2f 1200 CL X~
0 1 000 O 800 600 400 200 0
0 50 100 150 200 250 300 350 400 450 500 ModeratorTemperature (deg. F)
Figure 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 0, 25, 50 and 100 'F/hr) Applicable for the First 16 EFPY I (Without Margins for Instrumentation Errors) 5
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a (Page I of 2)
Braidwood Unit 1 Heatup* Data Points at 16 EFPY (Without Margins for Instrumentation Errors)
Heatup Curve 100 F Heatup Criticality Leak Test Limit Limit T P T P T P 60 0 210 0 188 2000 60 565.09 210 611.83 210 2485 65 565.09 210 597.56 70 565.09 210 585.60 75 565.09 210 576.77-80 565.09 210 570.35 =
85 565.09 210 566.61 90 565.09 210 565.09 95 565.09 210 565.87 100 565.87 210 568.69 -I 105 568.69 210 573.56 110 573.56 210 580.30 =
115 580.30 210 588.84 120 588.84 210 599.36 = -
125 599.36 210 611.78 T 130 611.78 210 626.07 =
135 626.07 210 642.16 140 642.16 210 660.36 145 660.36 210 680.59 150 680.59 210 702.80 155 702.80 210 727.33 160 727.33 210 754.07 165 754.07 210 783.17 170 783.17 215 814.98 175 814.98 220 849.37 180 849.37 225 886.54 185 886.54 230 926.73 190 926.73 235 970.11 195 970.11 240 1016.91 200 1016.91 245 1067.33 205 1067.33 250 1121.63 210 1121.63 255 1180.01 215 1180.01 260 1242.62 220 1242.62 265 1309.84 225 1309.84 270 1382.03 230 1382.03 275 1459.45 235 1459.45 280 1542.27 240 1542.27 285 1630.97 245 1630.97 290 1726.05 250 1726.05 295 1827.80 6
BRAID WOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a Page 2 of 2 Heatup Curve 100 F Heatup Criticality Leak Test Limit Limit T P T P T P 255 1827.80 300 1936.51 260 1936.51 305 2052.39 265 2052.39 310 2176.33 270 2176.33 315 2308.42 275 2308.42 320 2449.09 280 2449.09
- Heatup and Cooldown data includes vessel flange requirements of I 10'F and 621 psig per I OCFR50, Appendix G.
7
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1b Page 1 of I Braidwood Unit I Cooldown* Data Points at 16 EFPY**
(Without Margins for Instrumentation Errors)
Cooldown Curves Steady State 25 OF 50 OF 100 OF Cooldown Cooldown Cooldown T T P T P T P 60 0 60 0 60 0 60 0 60 620.27 60 577.45 60 534.28 60 446.98 65 621.00 65 590.68 65 548.52 65 463.79 70 621.00 70 605.03 70 563.98 70 481.93 75 621.00 75 620.51 75 580.67 75 501.49 80 621.00 80 621.00 80 598.51 80 522.68 85 621.00 85 621.00 85 617.90 85 545.50 90 621.00 90 621.00 90 621.00 90 570.23 95 621.00 95 621.00 95 621.00 95 596.83 100 621.00 100 621.00 100 621.00 100 621.00 105 621.00 105 621.00 105 621.00 105 621.00 110 621.00 110 621.00 110 621.00 110 621.00 110 795.92 110 766.92 110 739.27 110 690.04 115 821.55 115 794.59 115 769.53 115 726.24 120 849.00 120 824.45 120 801.97 120 765.12 125 878.42 125 856.54 125 836.87 125 807.07 130 910.25 130 890.97 130 874.41 130 852.23 135 944.34 135 928.00 135 915.03 135 900.91 140 980.89 140 967.79 140 958.57 140 953.33 145 1020.15 145 1010.84 145 1005.42 145 1009.81 150 1062.35 150 1056.88 150 1055.76 155 1107.92 155 1106.38 160 1156.42 165 1208.78 170 1265.05 175 1325.37 _
180 1390.04 185 1459.41 190 1533.55 195 1613.49 _
200 1699.01 205 1790.55 210 1888.61 215 1993.61 =
220 2105.69 225 2225.77 230 2353.75
- Heatup and Cooldown data includes vessel flange requirements of I10TF and 621 psig per IOCFR50, Appendix G.
- For each cooldown rate, the steady-state pressure values shall govern the tempcrature where no allowabic pressure values are provided.
8
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 2400 2200 2000
--I/i 0
r5 a.
1800 __ I I i I /
w
- 1600 U,
1 co w
Ix
- a. 1400 Unaccep able Operalltn 0
C 1200 Acceptablt
________I - O-veratlon 0 1000 z
800 _-58 1--
600 400 .I .. ,. .. . . I I I PCvI i 0 50 100 150 200 250 300 350 400 450 500 AUCTIONEERED LOW RCS TEMPERATURE (Deg F)
Figure 2.3 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the first 16 EFPY I 9
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Braidwood Unit I Nominal PORV Setpoints for the LTOP System Applicable for the First 16 EFPY PCV-455A PCV-456 (ITY-0413M) (ITY-0413P)
AUCTIONEERED RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE LO\\'
RCS TEMP. (DEG. F) (PSIG) RCS TEMP. (DEG. F) (PSIG) 50 497 50 513 70 497 70 513 100 497 100 513 110 497 110 513 160 497 160 513 200 618 200 634 250 603 250 619 300 588 300 604 350 588 350 604 450 2350 450 2350 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 350'F, linearly interpolate between the 350'F and 450'F data points shown above.
10
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 8) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME Section III, NB-233 1. 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 E185-82.
The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in material properties. The surveillance capsule testing has been completed for the original operating period.
11
BRAIDW'OOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Braidwood Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time(b) Estimated Capsule l (Degrees) Factor(a) (EFPY) Fluence (n/cm 2 ) (a)
U 58.50 4.37 1.10 3.87x 1018(c)
X 238.50 4.23 4.234 1.24 x 1019(c)
W 121.50 4.20 7.61 2.09 x 1019(c)
Z 301.50 4.20 12.01 (d) l V 610 3.92 Standby l Y 2410 3.92 12.01 (d) l (a) Updated in Capsule W dosimetry analysis, (Reference 9).
(b) Effective Full Power Years (EFPY) from plant startup.
(c) Plant specific evaluation.
(d) Capsule removed and is stored in the spent fuel pool. Capsule has not been analyzed and therefore capsule fluence has not been estimated.
12
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.
Table 4.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data. The values of the CF listed in Table 4.1 are those obtained from the most recent Unit 1 Capsule data, Capsule W, (Reference 9). However, these values were not used in calculating the Adjusted Reference Temperature (ART) values that were used to generate the Braidwood Unit I Heatup and Cooldown Curves. The ART values listed in Table 4.3, based on Capsules U and X data, continue to be the basis for the Braidwood Unit 1 curves (Reference 10)
Table 4.2 provides the reactor vessel material properties table.
Table 4.3 provides a summary of the Braidwood Unit 1 adjusted reference temperature (ARTs) at the 1/4T and 3/4T locations for 16 EFPY. The ART values listed in Table 4.3 are based on Capsules U and X data and continue to be the basis for the Braidwood Unit I curves (Reference 10).
Table 4.4 shows the calculation of ARTs at 16 EFPY for the limiting Braidwood Unit 1 reactor vessel material, i.e. weld WF-562 ( HT # 44201 1, Based on Surveillance Capsules U and X Data).
Table 4.5 provides RTPTS calculation for Braidwood Unit I Beltline Region Materials at EOL (32 EFPY), (Reference 11).
Table 4.6 provides RTpTs calculation for Braidwood Unit I Beltline Region Materials at Life Extension (48 EFPY), (Reference 11).
13
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 4.1 Braidwood Unit I Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule f'a) FF(b) ARTNDT(0) FF*ARTNDT FF2 Lower Shell Forging U 0.387 0.737 5.78 4.26 0.543 49D867/49C813-1 X 1.24 1.060 38.23 40.52 1.124 (Tangential) W 2.09 1.201 24.14 28.99 1.442 Lower Shell U 0.387 0.737 0.0 0.0 0.543 Forging 49D867-1 X 1.24 1.060 28.75 30.48 1.124 49C813-1 V 2.09 1.201 37.11 44.57 1.442 (Axial)
SUM: 148.82 6.218 CFForging = X(FF *A RTNDT) + ( FF2) = (148.82) . (6.218) = 23.9°F Braidwood Unit I U 0.387 0.737 17.06 12.57 0.543 Surv. Weld X 1.24 1.060 30.15 31.96 1.124 Material (Heat2#442011) V 2.09 1.201 49.68 59.67 1.442 Braidwood Unit 2 U 0.40 0.746 0.0 0.0 0.557 Surv. Weld X 1.23 1.058 26.3 27.83 1.119 Material (Heat # 442011)
W 2.25 1.220 23.9 29.16 1.488 SUM: 161.19 6.273 CF = X:(FF
- ARTNDT)
- Z( FF2 ) = (161.19) . (6.273) = 25.7°F Notes:
(a) f= Calculated fluence, (x 1019 n/cm2 , E> 1.0 MeV)
(b) FF = fluence factor =-o2- 0.log f)
(c) ARTNDT values are the measured 30 ft-lb shift values.
14
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2 Braidwood Unit I Reactor Vessel Material Properties Material Description Cu (%) Ni (%) Chemistry Initial l _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ __ _ _ _ _ Factor(a) RT NDT (OF)(a)
Closure Head Flange Heat # 5P7381/3P6406 0.11 0.67 -20 Vessel Flange _
Heat # 122N357V 0.77 -10 Nozzle Shell Forging
- 0.04 0.73 26.OOF(b) 10 Heat # 5P-7016 Intermediate Shell Forging
- Heat # 49D383-1/49C344-1 0.05 0.73 31.OOF(b) -30 (also referred to as the Upper Shell forging)
Lower Shell Forging
- 31.0oF~b)
Heat # 49D867/49C813-1 0.05 0.74 23.90F(c) -20 Circumferential Weld
- 4 41.0F(b)
(Intermediate Shell to Lower Shell) 0.03 0.67 25.71F(c) 40 WF-562 (HT# 442011)
Upper Circumferential Weld *
(Nozzle Shell to Intermediate Shell) 0.04 0.46 54.0oF(b) -25 WF-645 (HT# H4498)
- Beltline Region Materials a) The Initial RTNDT values for the plates and welds are based on measured data.
b) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position 1.1.
c) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position 2.1.
15
BRAIDW'OOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Braidwood Unit 1 Adjusted Reference Temperatures (ARTs) at 1/4T and 3/4T Locations for 16 EFPY(C) 16 EFPY(C) I Material Description 1/4T ART(0 F) 3/4T ART(-F)
Intermediate Shell Forging 25.1 8.2 Heat # 49D383-1/49C344-1 (RG Position 1)
Lower Shell Forging 26.2 12.1 Heat # 49D867/49C813-1 (RG Position 1) 13.4 3.2 Using Surveillance Data(a)
(RG Position 2(a))
Circumferential Weld (Intermediate Shell to Lower Shell)
WF-562 (HT# 442011) 112.9 90.5 (RG Position 1)
Using credible surveillance 7 6 .6 (b) 65.4(b)
Data (RG Position 2(a))
(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Position 2.
(b) These ART values were used to generate the Braidwood Unit I Heatup and Cooldown curves, (Reference 3).
(c) The applicability date has been increased from 14 EFPY to 16 EFPY based on an evaluation approved by the NRC in Reference 12.
16
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 Braidwood Unit 1 Calculation of Adjusted Reference Temperatures (ARTs) at 16 EFPY(b) at the Limiting Reactor Vessel Material I WVeld Metal (Based on Surveillance Capsule Data)
Parameter Values Operating Time 16 EFPY(b) I Location(c) 1/4T ART(-F) 3/4T ART(-F)
Chemistry Factor, CF (IF) 20.6 20.6 Fluence(f), n/cm2 6.73 x 1018 2.43 x10l8 (E>1 .0 Mev)(a)
Fluence Factor, FF 0.889 0.616 ARTNDT= CFxFF(0 F) 18.31 12.70 Initial RTNDT., I(0 F) 40 40 Margin, M (0 F) 18.31 12.70 ART= I+(CF*FF)+M,OF 76.6 65.4 per RG 1.99, Revision 2 ._.
(a) Fluence f, is based upon fsud (E > 1.0 Mev) = 1.120 x 10 19 at 14 EFPY for uprated conditions.
(b) The applicability date has been increased from 14 EFPY to 16 EFPY based on an evaluation approved by the NRC in Reference 12.
(c) The Braidwood Unit I reactor vessel wall thickness is 8.5 inches at the beltline region.
17
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5 RTpTs Calculation for Braidwood Unit I Beltline Region Materials at EOL (32 EFPY)
Material Fluence FF CF ARTpTs(c) Margin (IF) RTNDT(U)(2) RTPTS(b) 2 (1019 n/cm , (°F) (OF) (OF) (OF)
E>1.0 AfeN)
Intermediate Shell Forging 2.05 1.20 J 31.0 37.2 34 -30 41 Heat # 49D383-1/49C344-1 Lower Shell Forging 2.05 1.20 31.0 37.2 34 -20 51 Heat # 49D867/49C813-1 2.05 1.20 23.9 28.7 17 -20 26 Lower Shell Forging (Using S/C Data)
Nozzle Shell Forging 0.608 0.86 26.0 22.4 22.4 10 55 Heat # 5P-7016 II Inter. to Lower Shell Circ. Weld 1.99 1.19 41.0 48.8 48.8 40 138 WF-562 (HT# 442011) 1 2 3 2 4 1.99 1.19 25.7 30.6 28 40 99 Inter. to Lower Shell Circ. Weld 1.99 Using S/C Data _ _ I I I I Nozzle Shell to Inter. Shell Circ. 0.608 0.86 54.0 46.5 46.5 -25 68 W eld _ _ _ _ ___ __
WF-645 (HT# H4498) IIIII (a) Initial RTNDT values are measured values.
(b) RTPTS = RTNDT(u) + ARTPTS + Margin (OF)
(c) ARTpTs = CF
- FF 18
BRAID WOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6 RTPTS Calculation for Braidwood Unit I Beltline Region Materials at Life Extension (48 EFPY)
Material Fluence) I FF CF ARTprTstc) Margin (0f) RTNDT(U)(a) RTpTs(b)
(10"sn/Cm2' (OF;) (OF~) (Oi;) (OF)r E>1.O MeV)
Internediate Shell Forging 3.06 1.30 31.0 40.3 34 -30 44 Heat # 49D383-1/49C344-1 _
Lower Shell Forging 3.06 1.30 31.0 40.3 34 -20 54 Heat # 49D867/49C813-1 Lower Shell Forging 3.06 1.30 23.9 31.1 31.1 -20 42 Using S/C Data Nozzle Shell Forging 0.909 0.97 26.0 25.2 25.2 10 60 Heat # 5P-7016 Inter. to Lower Shell Circ. Weld 2.98 1.29 41.0 52.9 52.9 40 146 Metal WF-562 (HT# 442011)
Inter. to Lower Shell Circ. Weld 2.98 1.29 25.7 33.2 28 40 101 Using S/C Data Nozzle Shell to Inter. Shell Circ. 0.909 0.97 54.0 52.4 52.4 -25 80 Weld Metal WF-645 (HT# H4498)
(a) Initial RTNDT values are measured values.
(b) RTprs = RTNJTzu) + ARTpms + Margin (fF)
(c) ARTprs = CF
- FF 19
BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References
- 1. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Andrachek, J.D., et. al., January 1996.
- 2. NRC Letter from R. A. Capra to O.D. Kingsley, Commonwealth Edison Company, "Byron Station Units 1 and 2 and Braidwood Station Units I and 2, Acceptance for referring of pressure temperature limits report, (M98799, M98800, M98801, and M98802)," January 21 1998.
- 3. WCAP-14243, "Commonwealth Edison Company, Braidwood Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," March 1995.
- 4. Westinghouse Calculation CN-EMT-01-8, "Braidwood Units 1 and 2, Development of New Pressure Temperature Limit Curves and Evaluation of Byron Units I and 2 PT Curves EFPY."
- 5. Westinghouse Letter to Commonwealth Edison Company, CCE-95-186, "Braidwood Unit 1 LTOPS Setpoints Based on 16 EFPY P/T Limits," June 5, 1995.
- 6. ComEd Calculation BRW-96-9061/BYR 96-293, "Channel Accuracy for Power Operated Reief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 1 Original Steam Generators and Replacement Steam Generators)," Revision 0.
- 7. ComEd Nuclear Fuel Services Department, NDIT No. 960194, "Maximum Allowable LTOPS PORV Setpoints for Braidwood Unit I with RSGs," Revision 2.
- 8. WCAP-9807, "Commonwealth Edison Company, Braidwood Station Unit I Reactor Vessel Radiation Surveillance Program," February 1981.
- 9. WCAP-15316, "Analysis of Capsule W from the Commonwealth Edison Company Braidwood Unit I Reactor Vessel Radiation Surveillance Program," December 1999.
- 10. Letter from J. D. von Suskil (Exelon Generation Company, LLC) to U.S. NRC, "Braidwood Station Response to U. S. NRC Request for Additional Information Regarding the Braidwood Station Pressure-Temperature Limits Report", dated August 30, 2002.
- 11. WCAP-15365, Revision 1, "Evaluation of Pressurized Thermal Shock for Braidwood Unit 1," September 2000.
- 12. NRC Letter from G. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC, "Issuance of Amendments: Revised Pressure-Temperature Limits Methodology; Byron Station, Units I and 2, and Braidwood Station, Units 1 and 2," dated October 4, 2004.
20
BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
Revision 3
BRAIDWOOD - UNIT 2 PRESSURE AND TErVIPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 1 2.0 Operating Limits 1 2.1 RCS Pressure and Temperature (P/T) Limits I 2.2 Low Temperature Overpressure Protection (LTOP) 2 System Setpoints 2.3 LTOP Enable Temperature 2 2.4 Reactor Vessel Boltup Temperature 2 2.5 Reactor Vessel Minimum Pressurization Temperature 3 3.0 Reactor Vessel Material Surveillance Program II 4.0 Supplemental Data Tables 13 5.0 References 20 i
BRAID WOOD - UNIT 2 B}RAIDNVOOD -UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations 4 (Heatup Rates up to 1000 F/hr) Applicable for the First 16 EFPY Using the 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations 5 (Cooldown Rates of 0, 25, 50 and 1000 F/hr) Applicable to the First 16 FPPY using 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) 2.3 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature 9 Overpressure Protection (LTOP) System Applicable for the First 16 EFPY ii
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1a Braidwood Unit 2 Heatup Data at 16 EFPY using 1996 6 Appendix G Methodology (Without Margins for Instrumentation Errors) 2.1b Braidwood Unit 2 Cooldown Data Points 16 EFPY using 8 1996 Appendix G Methodology (Without Margins for Instrumentation Errors) 2.2 Data Points for Braidwood Unit 2 Nominal PORV 10 Setpoints for the LTOP System Applicable for the First 16 EFPY 3.1 Braidwood, Unit 2 Capsule Withdrawal Schedule 12 4.1 Braidwood Unit 2 Calculation of Chemistry Factors Using 14 Surveillance Capsule Data 4.2 Braidwood Unit 2 Reactor Vessel Material Properties 15 4.3 Summary of Braidwood Unit 2 Adjusted Reference 16 Temperatures (ARTs) at the 1/4T and 3/4T Locations for 16 EFPY 4.4 Braidwood Unit 2 Calculation of Adjusted Reference 17 Temperatures (ARTs) at 16 EFPY at the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data) 4.5 RTprs Calculation for Braidwood Unit 2 Beltline Region 18 Materials at EOL (32 EFPY) 4.6 RTprs Calculation for Braidwood Unit 2 Beltline Region 19 Materials at Life Extension (48 EFPY)
Hii
BRAII)WOOD - UNIT 2 PRESSURE AND TEIMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 2 has been prepared in accordance with the requirements of Braidwood Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)". Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications addressed in this report are listed below:
LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 Operating Limits The PTLR limits for Braidwood Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A (Reference 1) was used with the following exception:
a) Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, This exception to the methodology in WCAP 14040-NP-A has been reviewed and accepted by the NRC in Reference 2.
WCAP 15626, Reference 3, provides the basis for the Braidwood Unit 2 PIT curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. Reference 4 evaluated the effect of higher fluence from 5% uprate oni the existing PIT curves.
The applicability periods for all areas previously evaluated for 14.0 EFPY have been extended by two additional years to 16.0 EFPY. This applicability period extension was reviewed and approved by the NRC in Reference 10.
2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3).
2.1.1 The RCS temperature rate-of-change limits defined in Reference 3 are:
- a. A maximum heatup of 100°F in any 1-hour period.
- b. A maximum cooldown of 1000 F in any 1-hour period, and I
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REIPORT
- c. A maximum temperature change of less than or equal to 10F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.2 The RCS P/T limits for hcatup, inservice hydrostatic and leak testing, and criticality arc specified by Figure 2.1 and Tablc 2.1a. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1b. These limits are defined in Reference 3. Consistent with the methodology described in Reference l, with the exception noted in Section 2.0. the RCS PIT limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error.
These limits were developed using ASME Code Section XI, Appendix G, Article G2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.
The PIT limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding PW' curve for heatup and cooldown.
2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12).
The power operated relief valves (PORVs) shall each have nominal lift settings in accordance with Figure 2.3 and Table 2.2. These limits are based on Reference 5. The Residual Heat Removal (RH) Suction Relief Valves are also analyzed to individually provide low temperature overpressure protection. This analysis for the RH Suction Relief Valves remains valid with the current Appendix G limits contained in this PTLR documernt and will be reevaluated in the future as the Appendix G limits are revised.
The LTOP setpoints are based on PIT limits that were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error. The LTOP setpoints were developed using the methodology described in Reference l.
The LTOP PORV nominal lift settings shown in Figure 2.3 and Table 2.2 account for appropriate instrument error.
2.3 LTOP Enable Temperature The minimum required LTOP enable temperature is 200'F (Reference 6).
Braidwood Unit 2 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350'F and below and disarming of LTOP for RCS temperature above 350'F.
2'
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Note that the last LTOP PORV segment in Table 2.2 extends to 450'F where the pressure setpoint is 2335 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.
2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification)
Tile minimum boltup temperature for the Reactor Vessel Flange shall be 2 60'F.
Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere.
2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification)
The minimum temperature at which the Reactor Vessel may be pressurized (i.e.,
in an unvented condition) shall be 2 60'F, plus an allowance for the uncertainty of the temperature instrument, determined using a technique consistent with ISA-S67.04-1994.
3::.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMIPERATURE LIMITS REPORT Material Propertv Basis Limiting Material: Weld Meetal Limiting ART Valucs at 16 EFl'Y 1/4T 821F 3/4T 680 F I
2500 2250 2000 1750 a
U,1500 E.
1250 1
c t000 1
0 750 500 250 0
0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2.1 1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100°F/hr) Applicable for the First 16 EFPY Using the 1996 Appendix G Methodolkgy (Without Margins for Instrumentation Errors) 4.:.
BRAID WOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Material Property Basis Limiting Material: Vclcl Metal Limiting ART Values at 16 EFPY l/4T 821F 3141' 68F0 2500 lopedim Version:5.0 Run:23482 l 2250 ..... I. .. .. . . ...
I I .I{
2000
.- , - I, i^ "-'-.
-'. -'"'"1 --
1750 Unacceptable En Operatlon /i C,i
'- 1500 ......... 1-... .. . ... . . .-
0 Aceptable e Operation f-;
(a
, .... ... ........... . ...... . ~
0 Coohlown Rates i ,
750 / i r. ........ .... .... ........ .. . . .. . . . ................. i... ... ..... ... ..... .
Steady..-- ---
500 2..-. ........... ......... _.. .. ...........
250 so~-
_ .I.
0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50 and 1000FAihr) Applicable to the First 16 EFPY using 1996 Appendix G Methodology (Without Margins of Instrumentation Errors) 5 :i.:
BRAIDWOOD - UNIT 2 PRESSURE AND TEJMPERATURE LIMITS REPORT Table 2.1a (Page 1 of 2)
Braidwood Unit 2 Heatup* Data Points at 16 EFPY Using the 1996 Appendix GI Methodology (Without Margins for Instrumentation Errors)
Ifcatup Curve 100 F Heatup Criticality Leak Test Limit Limit T P T P T P 60 0 207 0 186 2000 60 617 207 621 207 2485 65 617 207 621 70 617 207 621 75 617 207 621 80 617 207 621 85 617 207 621 90 617 207 621 95 617 207 621 1(0 617 207 621 105 619 207 621 110 621 207 621 _ _ _
115 621 207 621 120 621 207 621 125 621 207 621 130 621 207 621 135 621 207 621 140 621 207 696 140 621 207 715 140 696 207 736 145 715 207 760 150 736 207 786 _
i55 760 207 815 160 786 210 846 165 815 215 880 170 846 220 917 _
175 880 225 957 180 917 230 10()0 185 957 235 1047 190 1000 240 1097 195 1047 245 1152 200 1097 250 1210 _
205 1152 255 1273 6!,
BRAIDWOOD) - UNIT 2 PRESSURE AND TEMIPERATURE LIMITS REPORT Table 2.1a Page 2 of 2 Heatup Curve 100 F 1aentup Criticality Leak Test Limit Limit T P T P T P 210 1210 260 1341 215 1273 265 1415 220 1341 270 1493 225 1415 275 1578 230 1493 280 1669 235 1578 285 1766 240 1669 290 1871 245 1766 295 1984 250 1871 300 2105 255 1984 305 2235 _
260 2105 310 2374 l _
265 2235 _ _ _ _ _ _-
270 2374 ; .
- Ileatup and Cooldown data includes thc vessel flangc requiremencts of 140 0Fand 621 psig per I OCFRSO, Aprpnlndix G..
7I .
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1b (Page 1 of 1)
Braidwood Unit 2 Cooldown* Data at 16 EFr4PYi-* Using the 1996 Appendix G I
Methodology (Without Margins for Instrumentation Errors)
Cooldown Curves Steady State 25 TF 50 T 100 'F Coollown Cooldown Coolcloxw T P T 1P T P T P 60 0 60 0 60 0 60 0 60 621 60 602 60 554 60 455 65 621 65 616 65 568 65 471 70 621 70 621 70 583 70 489 75 621 75 621 75 599 75 508 80 621 80 621 80 617 80 529 85 621 85 621 85 621 85 552 90 621 90 621 90 621 90 576 95 621 95 621 95 621 95 603 100 621 100 621 100 621 I00 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 140 621 140 621 140 621 140 1010 140 991 140 975 140 957 145 1050 145 1034 145 1022 145 1013 150 1092 150 1080 150 1072 150 1074 155 1137 155 1129 155 1126 155 1137 160 1186 160 1183 160 1185 160 1186 165 1239 165 1239 165 1239 165 1239 170 1295 170 1295 170 1295 170 1295 175 1356 175 1356 175 1356 175 1356 180 1422 180 1422 180 1422 180 1422 185 1492 185 1492 185 1492 185 1492 190 1567 190 1567 190 1567 190 1567 195 1649 195 1649 195 1649 195 1649 200 1736 200 1736 200 1736 200 1736 205 1830 205 1830 205 1830 205 1830 210 1931 210 1931 210 1931 210 1931 215 2039 215 2039 215 2039 215 2039 220 2156 220 2156 220 2156 220 2156 225 2281 225 2281 225 2281 225 2281 230 2416 230 2416 230 2416 230 2416
- I Icatup and Cooldown data iicludes the vessel nlangc rcquiremrents or 140 °F and 621 psig per I OCFR50, Appendix 0..
- For cach couldown ratc, thc steady-state pressure values shall goverm the temperature where no allowable pressLlre values arc provided.
8'
BRAIDNVOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2400e .
2200 :--- --- 1- -w 2000 _ _ ___ - __.- ___ l /
in aU) 1800 .
w 0.
C 140 __ ___ = = __ _ _ __ -
A. Unacteplable OJ tIl l
-J i
0 1000 __ __. /1 8 .___ _ _CV45 Op 400 0 50 100 150 200 250 300 350 400 450 500 AUCTIONEERED LOW RCS TEMPERATURE (Dcg. F)
Figure 2.3 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY I 9:
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points lor Braidwood Unit 2 Nominal PORV Setpoints for the LT17OP System Applicable for t(le First 16 EFPY I PCV-455A PCV-456 RCS TEMP. RCS Pressure RCS TEMP. RCS Pressure (DEG. F) (PSIG) (DEG. F) (PSIG) 50 495.8 5( 539.5 1(5 495.8 105 539.5 110 451.0 110 496.0 155 451.0 155 496.0 205 496.4 205 540.1 250 551.7 250 639.0 350 551.7 350 639.0 450 2335.0 450 2335.0 Note: To determine nominal lifl setpoints for RCS Prcssure and RCS Temperatures grcater than 350°F, linearly interpolate between thc 350°F and 450°F data points shown above. (Setpoints extend to 450°F to prevent PORV liftoff from an inadvertent LTOP system arming while at power).
I 0!.::
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 7) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME, Section 111, NB-233 1. 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 Xl of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets (he requirements of ASTM El 85-82.
The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in material properties. The surveillance capsule testing has been completed for the original operating period.
II1.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE IAMITS REPORT Table 3.1 Braidwood Unit 2 Capsule Withdrawal Schedule Capsule Location Capsule Lead Removal Time"b) Estimated Capsule (Degrees) Factorea' (EFPY) Fluence (n/cm 2 ) (a)
U 58.5° 4.41 1.15 4.00x 1018 (c)
X 238.50 3.85 4.215 1.23 x 1019 (c)
W 121.50 4.17 8.53 2.25 x 1019 (C)
Z 301.50 4.17 12.78 (d) l V 61.00 3.92 Standby I Y 241.00 3.92 12.78 (d)
Notes:
(a) Updated in Capsule W dosimetry analysis (Reference 8).
(b) Effective Full Power Years (EFPY) from plant startup.
(c) Plant specific evaluation.
(di) Capsule has been removed and stored in the spent fuel pool. Capsule has not been analyzed and therefore capsule fluence has not been estimated.
12.
BRAID WOOD - UNIT 2 PRESSURE AND TEMPENRATURE LIMITS REPORT 4.0 Supplemental Data Table The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of tile material property values shown were used as inputs to the PIT limits.
Table 4.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data (Reference 8).
Table 4.2 provides the reactor vessel material properties table.
Table 4.3 provides a summary of the Braidwood Unit 2 adjusted reference temperatures (ARTs) at the I14T and 3/4T locations for 16 EPPY.
Table 4.4 shows the calculation of ARTs at 16 EFPY for the limiting Braidwood Unit 2 reactor vessel material.
Table 4.5 provides RTIvrs Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPY), (Reference 9).
Table 4.6 provides RTIrTs Calculation for Braidwood Unit 2 Beltline Region Materials at Life Extension (48 EFPY), (Reference 9).
13:
BRAID WOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule fn FF(b) ARTNDr (C FF:YARTND (FF)2 Lower Shell Forging U 0.400 0.746 0.0 0.0 0.557 (50D102-1/50C97-1) X 1.23 1.058 0.0 0.0 1.119 (Tangential) _ 2.25 1.220 4.53 5.53 1.488 Lower Shell Forging U 0.400 0.746 0.0 0.0 0.557 (50D102-1/50C97-1) X 1.23 1.058 33.94 35.91 1.119 (Axial) W 2.25 1.220 33.2 40.50 1.488 Sum: 81.94 6.328 Chenmistry Factor = I(FF*ARTNDT, + Y(FF 2) = (81.94 ) + (6.328) = 12.9°F Braidwood I Surv.Weld Material U 0.387 0.737 17.06 12.57 0.543 X 1.24 1.060 30.15'd' 31.96 1.124 W 2.09 1.201 49 .68 md3 59.67 1.442 Braidwood 2 Surv. Weld Material U 0.40 0.746 0.0 0.0 0.557 X 1.23 1.058 26.3"'d 27.83 1.119 W 2.25 1.220 23.9("d 29.16 1.488 Sum: 161.19 6.273 Chemistry Factor = 1(FF*ARTNDTI + 1(FF1) = (161.19 ) (t6.273) = 25.7°F NOTES:
(a) f = Calculated nuence. ( x 10'9 n/cm 2, E > 1.0 MeV)
(b) FF= fluence factor = l4O.2 -o8l log n (c) ARTNDr values are the measured 30 ft-lb shift values (d) The surveillance weld metal ARTNDr values have not been adjusted.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LINITS REPORT Table 4.2 Braidwood Unit 2 Reactor Vessel Material Properties Material Description Cu (%) Ni (%) Chemistry Initial RTNDlT Factor(u) (OF) ( )
Closure Head Flange --- 0.75 20 Heat # 3P6566/5P7547/4P6986 Serial # 2031-V-1I Vessel Flange 0.07 0.70 20 Heat # 124P455 Nozzle Shell Forging
- 0.04 0.90 26.0oF(h) 30 Heat # 5P7056 Intermediate Shell Forging
- 0.03 0.71 20.0°F(b) -30 Heat # 49D963/49C904- I-1)
(also rceerred to as thc Upper Shcll Forging)
Lower Shell Forging
- 0.06 0.76 37.0°F(b) -30 Heat # 50D102/5OC97-1-l 12.9°F(c)
Circumferential Weld 0.03 0.67 41.0 F(b) 40 (Intcrmcdiatc Shell to Loc-r Shell) 25.7F(c)
Weld Scam WF-562 Heat # 442011 Circumferential Weld *
(Nozzlc Shcll to Intcrmediate Shell) 0.04 0.46 54.00 F(b) -25 Weld Seam WF-645 Heat # H4498
- Bcltline Region Materials (a) The initial RT Nm)- values for the plates and welds are based on measured data.
(b) Chemistry Factor calculated for Cu und Ni values per Regulatory Guide 1.99, Rev.2, Position 1.1 (c) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2. Position 2.1 1512
BRAIDNVOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Braidwood Unit 2 Adjusted Reference Tcmperaturc (ART's) at 1/4T and 3/4T Location for 16 EFP'Y(a)(b)
I Material 16 EFPY'b) 1/4T ART (OF) 3/4T ART (0 F)
Intermediate Shell Forging 3 -8 Heat # 49D963/49C904- I- 1)
Lower Shell Forging 30()
Heat # 50D I02/50C97- I-
-Using Surveillance Data 15 II Circumferential Weld 106 85 (Intcrmediate Shell to Lowver Shell)
Weld Scam WF-562 Heat # 44201 1
-Using Surveillanee Data 82 ( 68 '
Circumferential Weld (Nozzle Shell to Intcrmediate Shell) 29 8 Weld Seam WF-645 Heat # H4498 Nozzle Shell Forging 56 46 Heat # 5P7056 . ._._ ._.
(a) These ART values were used to calculate the Heatup and Cooldown curves in Figures 2.1 and 2.2 using the 1996 Appendix G Methodology.
(b) The applicability date has been increased from 14 EFPY to 16 EFPY based on an evaluation approved by the NRC in Reference 10.
16a.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTs) at 16 EFPY(d) at the Limiting Reactor Vessel Material Weld I Metal WF562 (Based on Surveillance Capsule Data)
Parameter Values Operating Time 16 EFPY(d)
Locationeb) 1/4T ART (°F) 3/4T ART(°F)
Chcmistty Factor, CF (°F) 25.7 25.7 Fluence(f), n/cm2 5.0 3 xlO b 1.81xlO01 (E>1.0 Mev))___
Fluence Factor, FF
__ _ __ _ _ 0.808 0.546 ARTNDr- CFxFF(°F) 20.77(c) 14.04 Initial RT NDT., 1(F) 40 40 Margin, M(°F) 20.77 14.04 ART= 1+(CF viFF)+M, oF 82 68 per RG_1.99,_Revision_2 __________ _________
a) Fluence, f, is the calculated peak clad/base nietal interface fluence (E> 1.0 Mev) =8.37x I O's n/cm 2 at 14 EFPY (Reference 3).
b) The Braidwood Unit 2 reactor vessel wvall thickness is 8.5 inches at the beltline region.
c) Using Regulatory Guide 1.99, Revision 2.
d) The applicability date has been increased from 14 ElPY to 16 EEPY based on an evaluation approved by the NRC in Reference 10.
17 :
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT TablIc 4.5 RTpvrs Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPY)
Material Fluence FF CF (0 F) ARTrms( Margin (°F) Rr'TonT(u) RTrrsz(l CI"tcm,(F) (°F) (°IF)
E>l.0 MIeV)
Intermediate Shell Forging 9 0 2. 36-01 49D963149C904- -II Heat #Het#4D634C0*I- 1.96 I .18 20 23.6 23.6 -30 17 Lower Shell Forging Heat #50Dl02/5oC97-r - 1.96 1.18 37 43.7 34 -30 48 Lower Shell Frging (Using S/C Dta) 'j' 1.96 1.18 12.9 15.2 34 -30 19 Nozzle Shell Forging 0.567 0.841 26 21.9 21.9 30 74 Heat # 5P-7056 ____ ____
Circumferential WelcI (Intenndiatc SmhcFl 6to Locr Slicil) 1.89 1.17 41.0 48.0 48.0 40 136 Heat # 442011 Circumferential Weld (Intcmiediate Shcl lto Lowcr Shell) 1.89 1.17 25.7 30.1 28 40 98 (Using SIC Data)
Circumferential Weld (Nozzle Shell to Intermediate Shcll) 0.567 0.841 54 45.4 45.4 -25 66 Weld Seam WF-645 Heat # H4498 (a) Initial RTNDT values are measured values.
(b) RTrrs = RT ,-,T(t,) + ARTtrs + Margin ( 0F).
(c) ARTI'rs = CF
- FF (d) Surveillance data is considered not credible. In addition, the Table chemistry factor is conservative and would normally be used for calculating RT Fr.. However, because the chemistry factor predicted by the Regulatory Guide I .99 Position 2.1 for the forging surveillance data was greater that the Position 1. I chemistry factor. then the Position 2.1 chemistry factor will be used to determine the RTlrrs with a full r, margin lerm.
18;_
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6 RTrrs Calculation for Braidwood Unit 2 Beltline Region Materials at Life Extension (48 EFPY)
Material Fluence FF CF (°F) ARTITscI I\largin RTNI)T(Ui'(a RT~rsp(b)
(I0"in/ckm2, (°F) (OF) (OF) (OF)
E>1.(O McV)
Intermediate Shell Forging 2.94 1.29 20 25.8 25.8 -30 22 Heat # 49D963t49C904- I- 2 Lower Shell Forging 2.94 1.29 37 47.7 34 -30 52 Heat # 50D 102/50C97 2 Lower Shell Forging 2.94 1.29 12.9 16.6 34 -30 21 (Using S/C Data) (d) 2.94_1_29 12_9_16_6 34 _ 30_2__
Nozzle Shell Forging 0.849 0.954 26 24.8 24.8 30 80 H eat 4# 5P.7056__ _ _ _ _ __ _ _ _ _ _ _ _ __ _ _ _ __ _ __ _ _ _ _ _
Circumferential Weld (lntrndelatc ShFll-5oLower Shll) 2.83 1.28 41.0 52.9 52.9 40 145 Weld Seam WF-562 Heat # 44201 1 Circumferential Weld (IntcrmncditccShelltotLowcrS ScIll 2.83 1.28 25.7 32.9 28 40 101 (Using SIC Data) _
Circumferential Weld (NWele Shell to Interccdiatc Shell) 0.849 0.954 54 51.5 51.5 -25 78 Weld Seam WF-645 Heat # H4498______
(a) Initial RTNDT valucs are measured values .
(b) RTpT!; = RTrNI) + ARTrrs + Margin (OF)
(c) ARTprs = CF
- FF (d) Surveillance data is considered not credible. In addition the Table chemistry factor is conservative and would normally be used for calculatin' RT ,I-s. 1 However, because the chemistry factor predicted by the Reg. Guide 1.99 Position 2.1 for the forging surveillance data was greater than the Position 1.1 chemistry factor then the Postion 2.1 chemistry factor will be used to determine the RTrrs with a full cr margin term.
19'.
BRAIDWOOD - UNIT 2 PRESSURE, AND TEMPERATURE LIMITS REPORT 5.0 References
- 1. WCAP-14040-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D.
Andrachek, et. al., January 1996.
- 2. Letter from G. F. Dick, NRC, to 0. D. Kingsley, Commonwealth Edison Company, "Exemption from Requirements of 10 CFR 50.60 - Byron, Units 1 and 2, and Braidwood, Units 1 and 2," dated January 16, 1998.
- 3. WCAP-15626, "Braidwood Unit 2 12 and 14 EFPY Heatup and Cooldown Limit Curves for Normal Operation using Uprated Fluences," January 2001.
- 4. Westinghouse Calculation CN-EMT-01-8, "Braidwood Units 1 and 2, Development of New Pressure Temperature Limit Curves and Evaluation of Byron Units I and 2 PT Curves EFPY."
- 5. Braidwood Station Design Change Package 9900519 (Setpoint Scaling Change Request 00- 106), "Revise Unit 2 Low Temperature Overpressure Protection System setpoints/Scaling for Pressurizer Power Operated relief Valves."
- 6. NRC Letter from R. A. Capra to O.D. Kingsley, Commonwealth Edison Company, "Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Acceptance for referring of pressure temperature limits report, (M98799, M98800, M98801, and M98802)," January 21 1998.
- 7. WCAP- 1188, "Commonwealth Edison Company, Braidwood Station Unit 2 Reactor Vessel Surveillance Program," December 1986.
- 8. WCAP-15369, "Analysis of Capsule W from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program,"
March 2000.
- 9. WCAP-15381, "Evaluation of Pressurized Thermal Shock for Braidwood Unit 2", T.J. Laubham, September 2000.