ML20247B199

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Revised Pressure Temp Limits Rept for Byron Unit 2
ML20247B199
Person / Time
Site: Byron Constellation icon.png
Issue date: 03/24/1998
From:
COMMONWEALTH EDISON CO.
To:
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ML20247B188 List:
References
NUDOCS 9805070310
Download: ML20247B199 (24)
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i ATTACHMENT G BYRON UNIT 2 PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

(Revised March 24,1998) l l

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) 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

l 3.0 Reactor Vessel Material Surveillance Program 9

4.0 Supplemental Data Tables 11 5.0 References 18 Attachment NCAP-14824, Revision 2 " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwood," November, 1997.

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BYRON - UNIT 2 j

PRESSURE AND TEMPERATURE LIMITS REPORT l

l List of Figures I

Figure Page 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4

l 100 F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates 5

up to 100 F/hr) Applicable for the Fitst 12 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology)

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2.3 Byron Unit 2 Maximum Allowable Nominal PORV Setpoints for the Low 7

Temperature Overpressure Protection (LTOP) System Applicable for the First 12 EFPY 9

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BYRON - UNIT 2 i

i PRESSURE AND TEMPERATURE L-IMITS REPORT List of Tables Table Page 2.1 Byron Unit 2 Heatup and Cooldown Data Points at 12 EFPY 6

(Without Margins for Instrumentation Erars) 1 2.2 Data Points from Byron Unit 2 PORV Setpoints for the LTOP System 8

3.1 Byron Unit 2 Capsule Withdrawal Schedule 10 4.1 Byron Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data 12 4.2 Byron Unit 2 Reactor Vessel Material Properties 13 4.3 Summary of Byron Unit 2 Adjusted Reference Temperatures (ARTS) at the 1/4T 14 and 3/4T Locations for 12 EFPY 4.4 Byron Unit 2 Calculation of ^djusted Reference Temperatures (ARTS) at 15 12 EFPY at the Limiting React-- Vessel Material Weld Metal (Based on Surveillance Capsule Data) 4.5 RTp13 Values for Byron Unit 2 for 32 EFPY 16 4.6 RTns Values for Byron Unit 2 for 48 EFPY 17 i

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BYRON - LWIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

' 10 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

Reference to Technical Specifications (TS) numbers are given in both the Byron Station current Technical Specifications (CTS) and Improved Technical Specifications (ITS). The CTS number is presented first, followed by the ITS number in brackets [ ].

This PTLR for Unit 2 has been prepared in accordance with the requirements of l

TS 6.9.1.11/[lTS-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:

LCO 3.4.9.1 Pressure / Temperature Limits; and LCO 3.4.9.3 Overpressure Protection Systems.

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[ITS-LCO 3.4.3 RCS Pressure and Temperature (PU) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System].

2.0 Operating Limits The PTLR limits for Byron Unit 2 were developed using a methodology specified in the l

l Technical Specifications. The methodology listed in WCAP-14040-NP-A (Reference 1) was used with four exceptions:

Use of ENDF/B-IV neutron transport cross-section library and ENDF/B-V dosimetry l

a.

reaction cross-sections,

b. Optional use of ASME Code Section XI, Appendix G, Article G-2000,1996 Addenda, c.

Use of ASME Code Case N-514, and

d. Use of RELAP computer code for calculation of LTOP setpoints for Byron Unit I replacement steam generators.

WCAP-14824, Revision 2, is included as an attachment for reference. WCAP-14824, Rev. 2 l

contains the P/T curves for Byron Unit 1, along with the weld metal data integration for Byron and Braidwood Units 1 and 2 and the Byron /Braidwood fluence methodologyjustification for ENDF/B-VI cross sections. WCAP-14940 (Ref.14) contains the Byron Unit 2 Heatup and Cooldown Limit Curves applicable to 12 EFPY.

i 2.1 RCS Pressure and Temperature (PH) Limits (LCO 3.4.9.1/ [ITS-LCO 3.4.3])

2.1.1 The RCS temperature rate-of-change limits defmed in Reference 14 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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BYRON - UNIT 2 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 by Figure 2.1 and Table 2.1. The RCS PTT limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defined in Reference 14. Consistent with the methodology described in Reference 1, the RCS P/T 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 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 Overprenare Protection (LTOP) System Setpoints (LCO 3.4.9.3/

[lTS-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 5,13 and 15. The Residual Heat Removal (RH) Suction Relief Valves are also analyzed to individually provide low temperature overpressure protection. This analysis for the RH

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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 PTT 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 LTCP PORV maximum lift settings shown in Figure 2.3 and Table 2.3 account for appropriate instrument error.

2.3 LTOP Enable Temperature The as-analyzed LTOP enable temperature is 200 F (Reference 16).

The TS required enable temperature for the PORVs shall be 2 350 F RCS temperature.

(Byron Unit 2 procedures governing the heatup and cooldown of the RCS require the l

arming of the LTOP System for RCS temper tre of 350 F and below and disarming of LTOP for RCS temperature above 350 F).

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (continued)

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)

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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 2).

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.

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l BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT l

l MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using surv. capsule data)

LIMITING ART VALUES AT 12 EFPY:

1/4T, 87.6*F 3/4T,71.5 F l

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0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F)

Figure 2.1 Byron Unit 2 Reactor Coolant System IIcatup Lirnitations (IIcatup Rates up to 100 F/hr)

Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) 4

f HYRON - UNIT 2 PRESSURE AND TEMPERATUiu LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using surv. capsule data)

LIMITING ART VALUES AT 12 EFPY:

1/4T, 87.6*F 3/4T,71.5 F 2500 l

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0 50 100 150 200 250 300 350 400 450 500 Moderat.or Temperature (Deg.F)

Figure 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cea'down Rates un to 100 *F/hr)

Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) 5 4

BYRON o UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit 2 Heatup and Cooldown* Data Points at 12 EFPY**

(Without margins for instrumentation errors; Using 1996 Appendix G Methodology)

HEATUP CURVES COOLDOWN CURVES 100 F Heatup Leak Test Curve Criticauty Limit Limit Steady State 25 DEG F 50 DEG F 100 DEG F T I P T I P T I P T f P T I P T l t T I P 60 0

212 0

191 2000 60 0

60 0

60 0

60 0

60 621 212 649 212 2485 60 621 60 589 60 538 60 437 65 621 212 692 65 j 621 65 601 65 551 65 452 70 621 212 678 0

621 70 614 70 565 70 468 75 621 212 669 75 621 75 621 75 581 75 486 80 621 212 662 80 621 80 621 80 597 80 55 85 621 212 658 85 621 85 621 85 614 85 526 90 621 212 657 90 621 90 621 90 621 90 548 95 621 212 659 95 621 95 621 95 621 95 572 100 621 212 663 100 621 100 621 100 621 100 599 105 621 212 670 105 621 105 621 105 621 105 621 110, 621 212 65

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110 621 110 621 110 621 110 621 115 l 621 212 690 115 621 115 621 115 621 115 621 120 l 621

, 212 703 120 621 120 621 120 621 120 621

[621 125 212 719 125 621 125 621 125 621 125 621 130 i 621 212 738 130 621 130 621 130 621 130 621 135]621 35 621 135 621 135 621 135 ' 621 212 758 140 621 212 781 140 621 140 621 140 621 140 621 145 621 J2 807 145 621 145 621 145 621 145 621 150 621 212 835 150 621 150 621 150 621 150 621 150 738 215 866 150 Q 045 150 1028 150 1016 150 1006 155 758 220 900 155 1087 155 1074 155 1066 155 1066 160 781 225 937 160 1132 160 1123 160 1120 160 1131 165 807

,230 (977, 165 1180 165 1177 185 1177 170 835 235 l1021 170 1232 175 666 240 l1068 175 1288 180 900 245 }1119 180 1345 185 937 250 i 1174 185 1413 190 977 255 1233 190 1483 95] 1021 260 1297 195 1558 200 1068 265 1366 200 1638 205 1119 270 1441 205 1725_

l 210 1174 275 1521 210 1818 l

215 1233 280 1607 215 1918 l

220 1297 285 1699 220 2026 225

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290 1798 225 2141 i

230 1441 295 1905 230 2266 i

235 1521 300 2020 235 2400 l

240 1607 305 2143 i

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" For each cooldown rate, the steady-state pressure values shall govern the temperature where no allowable pressure values are provided

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT l

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50 100 150 200 250 300 350 400 AUCTIONEERED LOW RCS TEMPERATURE (DEG. F) i

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Figure 2.3 Byron Unit 2 Maximum Allowable Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 12 EFPY 7

l BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Byron Unit 2 Maximum Allowable PORV Setpoints for the LTOP System Applicable for the First 12 EFPY*

PCV-455A PCV-456 (2TY-0413M)

(2TY-0413P)

AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F)

(PSIG)

RCS TEMP. (DEG. F)

(PSIG) 50 497 50 514 70 497 70 514 100 497 100 514 1

120 446 120 462 150 446 150 462 200 446 200 462 I

250 587 250 604 300 587 300 604 350 587 350 604 I

450 2350 450 2350 Note: To determine maximum allowable 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.

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

  • 3.0 Reactor Vessel Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. The removal schedule is provided in Table 3.1. Also, the results of these analyses shall be used to update Figures 2.1 and 2.2, and Table 2.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.

The pressure vessel material surveillance program (Reference 6)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, RTer, which is determined in accordance with ASME Section III, NB-2331. The empirical relationship between RTmr 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.

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT l

l Table 3.1 l

Byron Unit 2 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal TimeN Estimated Capsule (Degrees)

Factor (EFPY)

Fluence (n/cm )

2 U

58.5 3.96 1.15 (Removed) 3.996 x 10'8 W

121.5 3.89 4.634 (Removed) 1.211 x 10

X 238.5 3.89 8.23 (EOL Wall) 2.192 x 10N 12.34(1.5 EOL Z

301.5 3.89 WallW) 3.288 x 10

V 61.0 3.64 Standby Y

241.0 3.64 Standby (a) Effective Full Power Years (EFPY) from plant startup.

(b) Maximum end oflicense (32 EFPY) inner vessel wall fluence.

(c) Derived from Table C-2 of WCAP-14824, Rev. 2 (Reference 2, which is the Attachment to this report).

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I BYRON - UNIT 2 PRESSUIE AND TEMPERATURE LIMITS REPORT

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' 4.0 Supplemental Data Tables

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The following tables provide suppkmental information on reactor vessel material properties and are provided to be consistrat with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/F 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 2 adjusted reference temperature (ARTS) at the 1/4T and 3/4T locations for 12 EFPY.

Table 4.4 shows the calculation of ARTS at 12 EFPY for the limiting Byron Unit 2 reactor vessel material weld metal (Based on Surveillance Capsule Data).

l Table 4.5 provides RTns values for Byron Unit 2 for 32 EFPY.

Table 4.6 provides RTn3 values for Byron Unit 2 for 48 EFPY.

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table'4.1 Calculation of Chemistry Factors Using Surveillance Capsule Data Fluence Material Capsule (n/cm,

ppto Measured FF* ART,

(FF)2 2

E> 1.0 ARTm Mev), f Lower Shell Forging 49D330/

U 3.996x10" 0.746 0

0 0.556 49C298 1 1

(Tangential)

W l.211x10" 1.053 5

5.267 1.110 Lower Shell Forging 49D330/

U 3.996x10" 0.727 25 18.65 0.556 49C298 1 (Axial)

W l.211x10" 1.053 40 32.73 1.110 Sum:

66.037 3.332 Chemistry Factor (* = 66.037 + 3.332 = 19.8*F Byron 1 Weld Metal WF-336N U

3.72x10" 0.727 0

0 0.00 0.529 X

1.39x10" 1.091 35 105(0 114.56 1.190 Byron 2 Weld Metal WF-447(4 U

3.996x10" 0.746 0

0 0

0.557 W

1.211x10" 1.053 30 90(d 94.77 1.110 Sum:

209.33 3.386 Chemistry Factor (0 = 209.33+3.386= 61.8*F (a) FF = Fluence Factor = fo2:4i ioso (b) Byron Unit 1 ARTm values were obtained from the surveillance Capsule X analysis (WCAP-13880, Reference 10).

The Byron Unit I capsule fluence values were recalculated using the ENDF/B-V scattering cross sections in 1994 and are documented in WCAP-14044 (Reference 8).

(c) Byron Unit 2 capsule fluence, FF, and ARTm values were obtained from the surveillance Capsule W analysis of WCAP-14064 (Reference 3) using the ENDF/B-V scattering cross sections.

(d) Chemistry Factor = Z (FF*ARTm) / Z ((FF)2)

(c) Adjusted ARTm per Ratio Procedure of Regulatory Guide 1.99, Rev. 2 (Ref.12). Ratio = 3.0. See Table 2 of WCAP 14824, Rev. 2, (Ref. 2). Actual ratio is 2.5 (68.0/27.0 = 2.5); However, for conservatism, ratio of 3.0 was used.

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

- Table 4.2 l

Byron Unit 2 Reactor Vessel Material Propertics Material Description Cu (%)(')

Ni(%)N Chemistry Initial Factor (')

RT sm ( F)N Closure Head Flange Not 0.74 OM Reported Vessel Flange Not 0.73 30M Reported Inter. Shell Forging 0.01 0.70 20.0

-20 49D329/49C297-1-1 Lower Shell Forging 0.05 0.71 32.2

-20 49D330/49C298-1-1 i

Circumferential Weld 0.05 0.62 68.0 10 WF-447 a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Rev. 2 (Reference 12).

b) initial RT,7 values are measured, WCAP-14063 (Reference 7) c) Closure head and vessel flange Initial RToy values are used for considering flange requirements for the heatup/cooldown curves, WCAP-14940 (Reference 14).

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Byron Unit 2 Adjusted Reference Temperatures (ARTS) t at 1/4T and 3/4T Locations for 12 EFPY 12 EFPY Material Description 1/4T ART ( F) 3/4T ART ( F)

Intermediate Shell Forging 12.1 1.7 49D329/49C297-1-1 (R.G.1.99 Position 1 (*))

Lower Shell Forging 31.7 14.9 49D330/49C298-1-1 (RG Position 1('))

Using capsule data 11.8 1.5 (RG Position 2('$

Circumferential Weld 119.2 83.7 i

WF-447 (RG Position 1('))

Using credible surveillance 87.6*)

71.5*)

capsule data (RG Position 2(*))

(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Positions 1 and 2 (Reference 12). Note that surveillance data for lower shell forging was determined to be NOT credible per WCAP-14824, Rev. 2, however, the results are reported here for completeness.

(b) These ART values were used to generate the Byron Unit 2 Heatup and Cooldown Curves, WCAP-14940 (Reference 14).

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 Byron Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 12 EFPYat the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data) l Parameter Values l

Operating Time 12 EFPY l

LocationM 1/4T ART 3/4T ART Chemistry Factor, CF ( F) 68.0 68.0 2

Fluence (f), n/cm 4.94x10" 1.78x10" (E>l.0 Mev)W l

Fluence Factor, FF 0.803 0.542 l

ARTm7= CFxFF( F) 49.63 33.50 i

Initial RTer,I( F) 10 10 Margin, M ( F) 28.00 28.00 ART = I+(CF*FF)+M, F 87.6 71.5 per RG 1.99, Revision 2 (a) Fluence, f, is based upon f,,,(E>l.0 Mev) = 8.22x10 at 12 EFPY, WCAP-14940 (Reference 14).

(b) The Byron Unit 2 :: actor vessel wall thickness is 8.5 inches at the beltline region.

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i BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT I

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Table 4.5 i

RT Values for Byron Unit 2 for 32 EFPY rrs CF M

RT ARTns RT pm ppm mu) ns Material (F)

(F)

(F)

(F)

( F)

Lower Shell Forging 32.2 2.192 1.213 34.0

-20 39.1 53.1 MK 24-3 i

Using Surveillance 19.8 2.192 1.213 34.0

-20 24.0 38.0 d

Capsule DataM Intermediate Shell Forging 20.0 2.192 1.213 24.3

-20 24.3 28.6 MK 24-2 Circumference Weld 68.0 2.192 1.213 56.0 10 82.5 148.5 Metal WF447 I

Using Surveillance 61.8 2.192 1.213 28.0 10 75.0 113.0 Capsule Data (a) 2.192x10" n/cm (E>l.0 Mev) for 32 EFPY from Byron 2 PTS report, WCAP-14054 (Reference 9)

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2 (b) FF (Fluence Factor) = ft 2"' *80 (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 12).

(d) Double margin is used here due to the base metal surveillance capsule data exceeding the one sigma criteria from the credibility evaluation.

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1 BYRON - UNIT 2 l

PRESSURE AND TEMPERATURE LIMITS REPORT 1

Table 4.6 RT Values for Byron Unit 2 for 48 EFPY ers CF 7 93 ppm M

RT ARTp1s RTers ag Material

( F)

(F)

(F)

(F)

( F)

Lower Shell Forging 32.2 3.288 1.312 34.0

-20 42.2 56.2 MK 24-3 Using Surveilh. rec 19.8 3.288 1.312 34.0

-20 26.0 40.0 d

Capsule Data")

Intennediate Shell Forging 20.0 3.288 1.312 26.2

-20 26.2 32.4 MK 24-2 Circumference Weld 68.0 3.288 1.312 56.0 10 89.2 155.2 Metal WF447 Using Surveillance 61.8 3.288 1.312 28.0 10 81.1 119.1 Capsule Data (a) 2.192x10" n/cm (E>l.0 Mev) for 32 EFPY from Byron 2 PTS report, WCAP-14054 (Reference 9). The following 2

calculation provides the 48 EFPY fluence value:

2.192x10" + ((2.I92x!0"-3.174x 10"y32-4.634 EFPY)) * (48-32 EFPY) = 3.288x10" n/cm 2 (b) FF (Fluence Factor) = f':""" 5 0 (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 12).

(d) Double margin is used here due to the base metal surveillance capsule data exceeding the one sigma criteria from the credibility evaluation.

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BYRON - UNIT 2 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.

WCAP-14824, Revision 2 " Byron Unit 1 Heatup and Cooldown Limit Curves for l

Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood,"

November,1997 plus Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE-97-231 and CAE-97-233, dated January 6,1998.

3.

WCAP-14064," Analysis of Capsule W from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," Malone, M.J., et al.,

July 1994.

4.

WCAP-12431," Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," Terek, E., et al.,

l October 1989.

5.

Westinghouse Letter to Commonwealth Edison Company, CAE-96-106," Byron Unit 1 and 2 LTOPS Setpoints Based on 10 and 12 EFPY #f Limits," January 17,1996.

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6.

WCAP-10398," Commonwealth Edison Company, Byron Station Unit 2 Reactor Vessel

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Radiation Surveillance Program," Singer, L.R., December 1983.

j 7.

WCAP-14063," Commonwealth Edison Company, Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," Peter, P.A., November 1994.

8.

WCAP-14044," Westinghouse Surveillance Capsule Neutron Fluence Reevaluation,"

Lippencott, E.P., April 1994, 9.

WCAP-14054," Evaluation of Pressurized Thermal Shock for Byron Unit 2," Peter, P.A., August 1994.

10. 10 CFR Part 50, Appendix G," Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, dated December 19,1995.

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11. 10 CFR 50.61, " Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events,"(PTS Rule) May 15,1991.
12. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99," Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.

18 i

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

. References (continued)

13. Comed Calculation BRW-96-9071/BYR 96-294," Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 2 Original Steam Generators)" Revision 0.
14. WCAP-14940," Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," Laubham, T. J., et al., October,1997 plus Westinghouse errata letters CAE-97-201 dated November 4,1997 and CAE-97-231 and CAE-97-232 dated January 6,1998.
15. Westinghouse Letter to Comed, CAE-97-202," Byron Unit 2 COMS Setpoints for 12 EFPY," October 23,1997.
16. Westinghouse Letter to Comed, CAE-97-211/CCE-97-290," Byron and Braidwood Units 1 and 2 ATmetal Evaluation," November 7,1997.

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1 ATTACIIMENT WCAP-14824, REVISION 2, BYRON UNIT 1 IIEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION AND SURVEILLANCE WELD METAL INTEGRATION FOR BYRON AND BRAIDWOOD s -. j

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