ML20207H756

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Pressure Temp Limits Rept (Ptlr)
ML20207H756
Person / Time
Site: Byron Constellation icon.png
Issue date: 06/28/1999
From:
COMMONWEALTH EDISON CO.
To:
Shared Package
ML20207H753 List:
References
NUDOCS 9907210202
Download: ML20207H756 (24)
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Text

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BYRON UNIT 1 PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

(June 281999)

II' h,7210202990712AM. D O C K 0 5 0 0 0, ,4 5 4 '

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CYRON - U, NIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page l

1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) 1

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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 j 2.4 Reactor Vessel Boltup Temperature 3 2.5 Reactor Vessel Minimum Pressurization Temperature 3 l

J 3.0 Reactor Vessel Material Surveillance Program 10 4.0 Supplemental Data Tables 12 5.0 References 19

p DYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4 100 F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors and Using the 1996 Appendix G Methodology)

. 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates 5

, up to 100 F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors and Using the 1996 Appendix G Methodology) 2.3 Byron Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low 8 Temperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY l

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- LYRON - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT i.-

List of Tables Table Page 2.1 l Byron Unit 1 Heatup ahd Cooldown Data Points at 16 EFPY ' 6 (Without Margins for Instrumentation Errors and Using the 1996 Appendix G Methodology)

, 2.2 Data Points from Byron Unit 1 PORV Setpoints for the LTOP System 9 3.1 ~ Byron Unit 1 Capsule Withdrawal Schedule 11 4.1 Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data 13 4.2. Byron Unit 1 Reactor Vessel Material Propenies 14 4.3 Summary of Byron Unit 1 Adjusted Reference Temperatures (ARTS) at the 1/4T 15 and 3/4T Locations for 16 EFPY -

4A. Byron Unit 1 Calculation of Adjusted Reference Temperatures (ARTS) at 16 j 16 EFPY at the Limiting Reactor Vessel Material Intermediate Shell Forging SP-5933 (Based on Surveillance Capsule Data) i 4.5 . RTns for Byron Unit 1 Beltline Region Materials at Life Extension (32 EFPY) 17 4.6 RTns for Byron Unit 1 Beltline Region Materials at Life Extension (48 EFPY) 18 I

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 ' Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

This PTLR for Unit I has been prepared in accordance with the requirements of TS 5.6.6. Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

TS-LCO 3.4.3 RCS Pressure and Temperature (P/r) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.

-2.0 Operating Limits .

The PTLR limits for Byron 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 ASME Code Section XI, Appendix G, Article G-2000,1996 Addenda, and b) Use of RELAP computer code for calculation of LTOP setpoints for Byron 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 16.

i WCAP-15124, Reference 17, provides the basis for the Byron Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for Byron and Braidwood Units 1 and 2 is documented in Reference 2.

2.1 RCS Pressure and Temperature (P/r) Limits TS-LCO 3.4.3 2.1.1 The RCS temperature rate-of-change limits defined in Reference 17 are:

a. A maximum heatup of 100*F in any 1-hour period,
b. A maximum cooldown of 100*F in any 1-hour period, and
c. A maximum temperature change ofless than or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

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B'/RON o UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (continued) 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified i by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in Figure 2 2 and TaNe j 2.1. These limits are defined in WCAP-15124, Rev. 0 (Reference 17). Consistent with the 1' methodology described in Reference I, the RCS P/r 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 G-2000,1996 Addenda The  ;

criticality limit curve specifies pressure-temperature limits for core operation to provide i '

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 i?

vessel must be at a temperature equal to or higher than the minimum temperature required for the f inservice hydrostatic test, and at least 40 F higher than the minimum permissible temperature in the corresponding P/r curve for heatup and cooldown 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints TS-LCO 3.412 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 5,13, and 14. 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 vali l with the current Appendix G limits contained in this PTLR document and will be reevaluated in th future as the Appendix G limits are revised.

The LTOP setpoints are based on P/r 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 1. The LTOP PORV maximum lift settings shown i igure 2.3 and Table 2.3 account for appropriate instrument error. j 2.3 LTOP Enable Temperature j The as analyzed LTOP enable temperature is 200 F (Reference 15 andl7)

The required enable temperature for the PORVs shall be s 350 F RCS temperature (Byron Unit [

l 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)

Note that the last LTOP PORV segment in Table 2 2 extends to 450 F where the pressure setpoint is 2350 psig Th3 is intended to prohibit POP V lift for an inadvertent LTOP system arming at power 2

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (continued) -

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 (Reference 17).

2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification) l Based on the steady-state limits specified in Table 2.1, the minimum temperature at which the Reactor Vessel may be pressurized (i.e., in an unvented condition) shall be 2 65 F, plus an allowance for the uncertainty of the temperature instrument, determined using a technique consistent with ISA-S67.04-1994.

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Figure 2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100

'F/hr) Applicable for the First 16 EFPY (Without Magins for Instrumentation Errors and Using the 1996 Appendix G Methodology) 4

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2500 ,!

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Figure 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 j

  • F/hr) Applicable for the Fint 16 EFPY (Without Margins for Instrumentation Errors and Using the I 1996 Appendix G Methodology)

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit 1 Heatup and Cooldown Data Points at 16 EFPY (Without Margins f for Instrumentation Errors and Using the 1996 Appendix G Methodology)

Heatup Curve Cooldown Curves 100 F Criticality Leak Test 25 F 50 F 100 F Heatup Limit Limit Steadu State Cooldown Cooldown Cooldown T P T P T P T P T P T P T P l 60 0 225 0 204 2000 60 0 60 0 60 0 60 0 60 587 225 587 225 2485 60 613 60 561 60 509 60 402 ;

65 587 225 587 65 621 65 572 65 520 65 414 l 70 621 70 582 70 5?1 70 427 j 70 587 225 587 '

75 587 225 587 75 621 75 594 75 544 75 442 80 587 225 587 80 621 80 607 80 557 80 458 85 587 225 587 85 621 85 620 85 572 85 475 90 587 225 587 90 621 90 621 90 588 90 494 95 587 225 587 95 621 95 621 95 605 95 514 1 00 587 225 587 100 621 100 621 100 621 100 535 105 587 225 587 105 621 105 621 105 621 105 559 110 587 225 587 110 621 110 621 110 621 110 584 115 587 225 588 115 621 115 621 115 621 115 611 120 588 225 591 120 621 120 621 120 621 120 621 125 591 225 596 125 621 125 621 125 621 125 621 130 596 225 602 130 621 130 621 130 621 130 621 135 602 225 611 135 621 135 621 135 621 135 621 140 611 225 622 140 621 140 621 140 621 140 621 145 621 225 634 145 621 145 621 145 621 145 621 ;

150 621 225 648 150 621 150 621 150 621 150 621 j 155 621 225 665 155 621 155 621 155 621 155 621 i 160 621 225 683 160 621 160 621 160 621 160 621 165 621 225 703 165 621 165 621 165 621 165 621 170 621 225 725 170 621 170 621 170 621 170 621 l 175 621 225 750 175 621 175 621 175 621 )

180 621 225 777 180 621 180 621 180 750 230 806 180 1207 180 1205 185 777 235 838 185 1261 190 806 240 872 190 1319 195 838 245 910 195 1382 200 872 250 950 200 1449 205 910 255 994 205 1521 210 950 260 1041 210 1599 215 994 265 1092 215 1683 220 1041 270 1147 220 1773 225 1092 275 1206 225 1869 230 1147 280 1269 230 1973 235 1206 285 1338 235 2085 6

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Heatup Curve Cooldown Curves 100 F Criticality Leak Test Steady 25 F 50 F 100 F Heatup Limit Limit State Cooldown Cooldown Cooldown T P T P T P T P T P T P T P 240 1269 290 1411 240 2205

i. 245 '1338 295 1490 245 2334 250 1411 300 1575 250 2473 255 1490 305 1666 260 1575 310 1764 265 1666 315 1869 270 1764 320 1982 275 1869 325 2104 280 1982 330 2234 285 2104 335 2374 290 2234 295 2374 Note 1: Heatup and Cooldown data includes the vessel flange requirements of 180 'F and 621 psig per 10CFR50, Appendix 0.

Note 2: For each cooldown rate, the steady-stats pressure values shall govern the ; .,,..ne where no allowable pressure values are provided.

Note 3: Temperatures and pnesares are given in' F and pais, respectively.

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0 80 100 180 200 250 300 300 400 AUCTIONEERED LOW RCS TEMPERATURE (DEG. F) i l

l Figure 2.3 Byron Unit 1 Maximum Allowable Nominal PORY Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY i

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT i

Table 2.2 Data Points for Byron Unit 1 Maximum Allowable Setpoints l

for the LTOP System Applicable for the First 16 EFPY l

PCV-455A PCV-456 (1TY-0413M) (1TY-0413P)

AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F) (PSIG) RCS TEMP. (DEG. F) (PSIG) 65 497 65 514 70 497 70 514 100 497 100 514 120 446 120 462 j 150 446 150 462 j 200 446 200 462 l 250 587 250 604 300 587 300 604 350 587 350 604 450 2350 450 2350 Note: To determine maximum allowable lift setpoints for RCS Pressure and RCS Temperstares greater than 350*F,lineady interpolate between the 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.)

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DYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 ' Remeter Vessel Material Surveillance Program The pressure vessel material surveillance program (Ref. 6) 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, RTer, which is determined in accordance with ASME,Section III, NB-2331. The empirical-relationship between RTer 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 El85-82.

The third and final reactor vessel material inadia: ion surveillance specimens have been removed and analyzed to determine changes in the reactor vessel material propenies. The surveillance capsule testing has been completed for the original operating period. Other capsules will be removed to avoid excessive fluence accumulation should they be needed to support life extension. The removal schedule is provided in Table 3.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.

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1 BYRON - UNIT 1 l

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PRESSURE AND TEMPERATURE LIMITS REPORT I

l Table 3.1 Byron Unit 1 Capsule Withdrawal Schedule l

Capsule Vessel Location Capsule Lead Removal Time (a) Estimated Capsule l 2

(Degrees) Factor (EFPY) Fluence (n/cm )

i U 58.5 4.22 1.15 (Removed) 4.04 x 10

X 238.5 4.27 5.64 (Removed) 1.57 x 10

W 121.5 4.20 9.24 (Removed) 2.43 x 10(b)

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i Z 301.5* 4.20 12.55(c) 3.27 x 10(d) t V 61.0' 3.97 13.6(c) 3.27 x 10(d)

Y 241.0 3.97 13.6(c) 3.27 x 10(d)

(a) Effective Full Power Years (EFPY) from plant stanup.

2 (b) Maximum end oflicena (32 EFPY) inner vessel wall fluence is estimated to be 1.95 x 10" n/cm .

(c) Standby capsule to be used for future life extension. Derived from Table 7-1 of WCAP-IS123, Rev.1 (Reference 18).

2 (d) 54 EFPY projected peak wall fluence is 3.27 x 10" n/cm ,

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DYRON - UNIT 1 1

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.

Table 4.2 provides the reactor vessel material properties table.

Table 4.3 provides a summary of the Byron Unit I adjusted reference temperature (ARTS) at the 1/4T and 3/4T locations for 16 EFPY.

Table 4.4 shows the calculation of ARTS at 16 EFPY for the limiting Byron Unit I reactor vessel material (Intermediate Shell Forging 5P-5933).

Table 4.5 provides RTersvalues for Byron Unit I for 32 EFPY obtained from Reference 9.

Table 4.6 provides RTersvalues for Byron Unit I for 48 EFPY obtained from Reference 9.

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4 BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Calculation of Chemistry Factors Using Surveillance Capsule Data Fluence Material Capsule (n/cm ,

2 ppm Measured FF*ARTer (FF)2 E>1.0 Mev), ARTer f

Inter. Shell U 4.04x10" 0.748 28.55 21.36 0.560 Forging SP-5933*) X 1.57x10" 1.124 9.82 I1.04 1.263 j (Tangential)' W 2.43x10 ' 1.239 49.20 60.96 1.535 Inter. Shell U 4.04x10 O.748 18.52 13.85 0.560 Forging SP-5933*) X 1.57x10 l.124 53.03 59.61 1.263 (Axial) W 2.43x10 1.239 29.34 36.35 1.535 Sum: 203.17 6.716 Chemistry Factor * = 203.17 + 6.716= 30.3*F Byron 1 Weld U 4.04x10 O.748 11.22 (5.61)

  • 8.39 0.529 Metal WF-336*)- X 1.57x10 l.124 80.22 (40.11)
  • 90.17 1.263 W 2.43x10" 1.239 102.68 (51.34)
  • 127.22 1.535 Byron 2 Weld U 4.05x10" 0.749 16.88 (8.44)
  • 12.64 0.561 Metal WF-447M W l.27 x10" 1.067 57.76 (28.88) W 61.63 1.138 X 2.30 x 10" 1.225 108.02 (54.01)
  • 132.32 1.500 Sum: 432.46 6.561 1

Chemistry Factor * = 432.46+6.561= 65.9'F* ,

(a) The Byron Unit land 2 capsule fluence values were recalculated using the ENDF/B-VI scattering cross sections and are documented in WCAP-15123 (Reference 18) and WCAP-15176 (Reference 20). FF

= Fluence Factor = ("#''8 0 (b) Byron Unit 1 ARTmvahes were obtained from the surveillana Capsule W analysis, WCAP-15123 (Reference 18).

Credibility assessment for Byron Unit I surveillance data alone is provided in WCAP-15183 (Reference 21).

v were obtained from the surveillance Capsule X analysis, WCAP-15176 (Reference 20).

(c) Byron Unit 2 ARTwr alues Credibility assessment for Byron Unit 2 surveillance data alone and the combined Unit I ar.d Unit 2 data is provided in WCAP-15180 (Reference 21).

(d) Chemistry Factor = I (FF*ARTer) / I ((FF)')

(c) A4 justed ARTer per Ratio Procedure of Regulatory Guide 1.99, Rev. 2 (Ref.12). Ratio = 2.0 See Table 4.8 of WCAP 15178, Rev. O. (Ref. 22).

(f) Surveillana data credibility menessment determined data from both Units 1 and 2 produced the limiting CF, WCAP- <

15180, Reference 21.

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BYRON - UNIT 1 j

)

PRESSURE AND TEMPERATURE LIMITS REPORT

. Tchte 4.2: Reactor Vessel Beltline Material Unirradiated Toughness Properties

. Material Description Cu(%)* Ni (%)* Initial RTsn/"

Closure Head Flange 124K358 val -- O.74 60 Vessel Flange 123J219 val --- 0.73 10 Nozzle Shell Forging 123J218 0.05 0.72 30 Intennediate Shell Forging SP 5933 0.04 0.74 40 Imer Shell Forging SP-5951 0.04 0.64 10 Intermediate to Lower Shell Forging Cire. Weld 0.04 0.63 -30 Seam WF-336 (Heat # 442002)

Nozzle Shell to Intennedsate Shell Forging Cire. 0.03 0.67 10 Weld Scam WF-501 (Heat # 442011)"

Byron Unit 1 Surveillance Program 0.02 0.69 ---

Weld Metal (Heat # 442002)

Byron Unit 2 Surveillance Program 0.02 0.71 ---

Weld Metal (Heat # 442002)

Braidwood Units 1 & 2 Surveillance Program 0.03 0.67, 0.71 ---

Weld Metals (Heat # 442011)  !

N91CL (a) TheinitialRTervalues for the plates and welds are based on measured data per Reference 2.

(b) Best Estimate Cu% and Ni% Per Reference 2 and 22.

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 4-3 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 16 EFPY Material 16 EFPY 1/4T ART 3/4T ART Intermediate Shell Forging 5P-5933 84 70

- Using Surveillance Data") 100*) 92*) I l

Lower Shell Forging SP-5951 54 40 Circumferential Weld WF-336 62 33

- Using Credible Surveillance Data") 54 37 Circumferential Weld WF-501 54 37

- Using Credible Surveillance Data 28 21 form Braidwood I and 2 Nozzle Shell Forging 123J218 64 51 NOTES:

(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Revision 2, Position 2 along with a full margin since it was determined that this data was not credible and the Table chemistry factor was non conservative.

(b) These ART values were used to generate the Byron Unit 1 16 EFPY heatup and cooldown cun es.

I (c) Calculated using the chemistry factor from the Byron Unit I and 2 integrated sun eillance data as repried in WCAP-15178 (Reference 22).

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 1

4 Byron Unit 1_ Calculation of Adjusted Reference Temperatures (ARTS) at 16 EFPY at the Limiting Reactor Vessel Material Intermediate Shell -

Forging 5P-5933 (Conservatively, Based on Surveillance Capsule Data)

Parameter Values Operating Time 16 EFPY l

~ Location *) 1/4T ART 3/4T ART Chemistry Factor, CF ("F) 30.3 30.3 Fluence (f), n/cm' 5.91x10 2.13x10

(E>1.0 Mev))(')

l Fluence Factor, FF 0.853 0.585 ARTun= CFxFF(*Fj 25.8 17.7 Initial RTmn., I( F) 40 40 j

Margin, M(*F) 34 34 ART = I+(CF*FF)+M, *F 100 92 l per RG 1.99, Revision 2 i l

a) Fluence, f, is based upon 6 (E>l.0 Mev) = 9.85x10

b) The Byron Unit I reactor vessel wall thickness is 8.5 inches at the beltline region.

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5: RTers for Byron Unit 1 Beltline Region Materials at EOL (32 EFPY) 1 Material Fluence FF CF ARTns'" Margin RT.wm > RTns*'

(n/cm', (*F) ("F) (*F) (*F) ( F)

E>1.0 MeV)

Intermediate Shell Forging SP-5933 1.95 x 10 l.18 26.0 30.7 30.7 40 101 Intermediate Shell Forging SP-5933 1.95 x 10 l.18 30.3 35.8 34 40 110

  • using S!C Data (d)

Lower shell Forging SP-5951 1.95 x 10 l.18 26.0 30.7 30.7 10 71 Inter, to Lower Shell Cire. Weld 1.95 x 10 l.18 54.0 63.7 56 -30 90 Metal WF-336 (442002)

Inter. to Lower Shell Cire. Weld Metal (442002) 1.95 x 10 l.18 65.9 77.8 28 -30 76

-+ using S/C Data (*)

Nozzle Shell Forging 123J18 5.83 x 10 O.849 31.0 26.3 26.3 30 83 Nozzle Shell to Inter. Shell Cire.

5.83 x 10 O.849 41.0 34.8 34.8 10 80 Weld Metal WF-501 (442011)

Nozzle Shell to Inter. Shell Cire. l Weld Metal (44201I) 5.83 x 10 O.849 16.7 14.2 14.2 10 38

-+ using S/C Data Notes.

(a) InitialRTervalues are measured values (See Table 4.2)

(b) RTns = RTam + ARTns + Margin (*F)

(c) ARTns = CF

  • FF (d) Surveillance data is considered not credible, however, since the chemistry factor (CF) from the Reg. Guide Tables (Pos.1.1) is lower (i.e. CF via Pos. 2.1 > CF via Pos.1.1), then the Pos. 2.1 CF is used to determine FTS with a full c amargin term, i.e.17 'F.

(c) Based on Byron Unit I and 2 integrated surveillance data chemistry factor from WCAP-15178 (Reference 22).

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6: RTers for Byron Unit 1 Beltline Region Materials at Life Extension (48 EFPY)

Material Fluence FF CF ARTns") Margin RTmyra3"' RTns" (n/cm'. (*F) ("F) ('F) ('F) (*F)

E>1.0 MeV)

Intermediate Shell Forging SP-5933 2.91 x 10 l.28 26.0 33.3 33.3 40 107 Intermediate Shell Forging SP-5933 2.91 x 10 1.28 30.3 38.8 34 40 113

-+ using S/C Data"'

Lower shell Forging 5P-5951 2.91 x 10 l.28 26.0 33.3 33.3 10 77 Inter. to Lower Shell Circ. Weld Metal WF-336 (442002)

Inter. to Lower Shell Cire. Weld Metal (442002) 2.91 x 10 1.28 65.9 84.4 28 -30 82

-+ using S/C Data")

Nizzle Shell Forging 123J218 8.70 x 10 O.%1 31.0 29.8 29.8 30 90 Nozzle Shell to Inter. Shell Cire.

8.70 x 10 O.%1 41.0 39.4 39.4 10 89 Weld Metal WF-501 (442011)

NIzzle Shell to Inter. Shell Cire.

Weld Metal (442011) 8.70 x 10 O.%1 16.7 16.0 16.0 10 42

-+ using S/C Data Notes (a) InitialRTumvalues are measured values (See Table 4.2)

(b) RTns = RTmnan + ARTns + Margin (*F)

(c) ARTns = CF

  • FF (d) Surveillance data is considered not credible, however, since the chemistry factor (CF) from the Reg. Guide Tables (Pos.1.1) is lower (i.e. CF via Pos. 2.1 > CF via Pos.1.1), then the Pos. 2.1 CF is used to determine PTS with a full c amargin term, i.e.17 'F.  ;

(e) Based on Byron Unit I and 2 integrated surveillance data chemistry factor from WCAP-15178 (Reference 22). I l

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BYRON - UNIT 1 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", J.D. Andrachek, et. al., January 1996.

2. LWCAP-14824, Revision 2," Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood", November 1997 with Westinghouse errata letters CAE-97-220, dated November 26,- 1997 and CAE-97-231/CCE-97-314 and CAE-97-233/CCE-97-316, dated January 6,1998.
3. WCAP-13880," Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", P.A. Peter, et. al., January 1994.
4. WCAP-12685, " Analysis of Capsule U from the Conunonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", E. Terek, et. al., August 1990.
5. Westinghouse Letter to Commonwealth Edison Company, CAE-96-106, " Byron Unit I and 2 LTOPS Setpoints Based on 10 and 12 EFPY P/T Limits", January 17,1996.'
6. WCAP-9517, " Commonwealth Edison Company, Byron Station Unit 1 Reactor Vessel Surveillance Program", J.A. Davidson, July 1979.
7. Westinghouse Letter Report to Commonwealth Edison Company, FDRT/SPRO-009(94), " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation", P.A. Peter, January 1994.
8. . WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation", E.P.

Lippencott, April 1994.

9. WCAP-15125," Evaluation ofPressurized Thermal Shock for Byron Unit 1", Revision 0, T.-J.

Laubham et al., November 1998.

10.10 CFR Part 50, Appendix G, " Fracture Toughness Requirements", Federal Register, Volume 60, No. 243, dated December 19,1995, 11.10 CFR 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal l Shock Events", May 15,1991. (PTS Rule)

12. Regulatory Guide 1.99, Revision 2, " Radiation Embrittlement of Reactor Vessel Materials", U.S.

Nuclear Regulatory Commission, May 1988.

- 13. Comed Calculation BRW-96-906I/BYR 96-293, Rev. O " Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 1 Original Steam Generators and Replacement Steam Generators)".

19

GYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 - References (continued)

14. Comed Nuclear Fuel Services Department NDIT No. 960186, Revision 1 " Maximum Allowable LTOPS PORV Setpoints for Byron Unit I with RSGs".
15. Westinghouse Letter to Comed, CAE-97-211/CCE-97-290," Byron and Braidwood Units 1 and 2 AT metal Evaluation," November 7,1997,
16. NRC Letter from R A.. Capra,-NRR, to O. D. Kingsley, Commonwealth Edison Co., " Byron Station, Units 1 ar.J 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limit 4 Report (TAC Numbers M98799, M98800, M98801, and M98802),"

January 21,1998.

- 17. WCAP- 15124, Revision 0, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, et al., November 1998.

18. WCAP-15123, Revision 1, " Analysis of Capsule W from Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et al., January 1999.
19. WCAP-15183, Revision 0, " Commonwealth Edison Company Byron Unit 1 Surveillance Program Credibility Evaluation," T. J. Laubham, et al., June 1999.

\

20. WCAP-15176, Revision 0, " Analysis of Capsule X from Commonwealth Edison Company Byron i Unit 2 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et al., March 1999. 1 i
21. WCAP-15180, Revision 0, " Commonwealth Edison Company Byron Unit 2 Surveillance Program Credibility Evaluation," T. J. Laubham, et al., June 1999.
22. WCAP- 15178, Revision 0, " Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, et al., June 1999.

20

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Attachment B Byron Station Unit 2 PTLR dated June 28,1999 l

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PDebyttrsWOO95. doc

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