ML20199E880

Revised Byron Unit 2 Pressure Temperature Limits Rept (PTLR)

ML20199E880
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Site:	Byron
Issue date:	11/03/1997
From:	COMMONWEALTH EDISON CO.
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References
NUDOCS 9711240019
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Text

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BYRON UNIT 2 PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

(Revised November 3,1997)

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BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT-Table of Contents -

Section Page i

1.0

Introductions

1 s

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

l 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 Progiam 9

4.0 Supplemental Data Tables 11 i

5.0 References 18 Attacluner*

WCAP 14824, Revision 2 " Byron Unit 1 Heatup and Cooldown Limit Curves for Norr-Operation and Surveillance Weld Metal Integration for Byron and Braidwood," Novemv.r, 1997.

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i BYRONo UNIT 2

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PRESSURE AND TEMI ERATURE LIMITS REPORT

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List of Figures

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' 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitadons (Heatup Rates up to 4

100' F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix 0 Methodology) 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates 5

up to 100 'F/le) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors: Using 1996 Appendix G Methodology)

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

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

iii

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BYRON. UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 Byron Unit 2 Heatup and Cooldown Data Points at 12 EFPY 6

i (Without Margins for Instrumentation Errors) 2.2 Data Points from Byron Unit 2 PORY 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 L4.2 Byron Unit 2 Reactor Vessel Material Properties 13 4.3 Summary of Byron Unit 2 Adjusted Reference Temperatures (ARTS) at the 1/4T 14 i

and 3/4T Locations for 12 EFPY 4.4 Byror. Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 15 12 EFPY at the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data)

't 4.5 RTns Values for Byron Unit 2 for 32 EFPY 16 4.6 RTns Values for Byron Unit 2 for 48 EFPY 17 1

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BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 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 [ ).

l This PTLR for Unit I has been prepared in accordance with the requirements of TS 6.9.1.1liilTS 5.6.6). Revisions to the PTLR shall be provided to the NRC after issuance.

l l

The Technical Specifications addressed in this report are listed below:

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

[ITS 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 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP 14040 NP A (Reference 1) was used with four exceptions:

Un of ENDF/B IV neutron transport cross section library and ENDF/B V dosimeter a.

reaction cross s~ections,

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 Unit I replacement steam generators.

WCAP-14824, Revision 2, is included as an attachment for reference. WCAP 14824, Rev. 2 contains the P/T cunts for Byron Unit 1, along with the weld metal data integration for Byron and Braidwood Units 1 and 2 and the Byron /Biaidwood 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.

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.9.1/ [ITS LCO 3.4.3))

2.1.1 The RCS temperature rate of-change limits defined 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 changt ofless than or equal to 10'F in any 1-bour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit cunes.

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IlYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (Continued) 2.1.2 The RCS P/T limits for '.catup, insenice hydrostatic and leak testing, and criticality are specified by Figure 2.. and Table 2.1. The RCS P/T limits for cooldown are shown in j

Figuie 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 0, Article G 2000,1996 Addenda. In determining compliance with Figures 2.1 and 2.2 and Table 2.1, mstrument uncertainties need not be con >ldered since appropriate station operating procedures ensure that the limits contained in the figures rnd table are not exceeded. 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 0.

The P/T limits for core operation (excep 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 (l.CO 3.4.9.3/

[lTS LCO 3.4.12]).

The power operated relief valves (PORVs) shall each have lift settings in accordance with Figure 2.3 and Table 2.2. These limits are based on References 5,13 and 15.

The LTOP setpoints are based on P/T limits which were established in accordance.vith 10 CFR 50, Appendix G without ailowance for instmmentation error and in accordance with the methodology described in Reference 1. The LTOP 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 goveming 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. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

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-4 BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (Continued) 2,4 Reactor Vessel Boltup Temperature (Non Technical Specification)

The minimum boltup teraperature for the Reactor Vessel Flange shall be h 60'F. Boltop 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 Pressuriution Temperature (Non-Technical SpecMication)-

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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PRESSURE AND TEMPERATURE LIMITS REPORT 4,

MATERIAL PROPERTY BASIS i

LIMITING MATERIAL: CIRCUMFERENTIAL WELD iusing eurv. capsule data)

LIMITING ART VALUES AT 12 EFPY:

./ IT, 87,6'F i

3s4T. 71.S'F 2500

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Figure 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 'F/br)

Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) l 4

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD (using sury. capsule cata)

LIMITING ART VALUES AT 12 EFPY:

1/4T. 87.6'F 3/47, 71.5'F 2500

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- Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 *F/hr)

Applicable for the First 12 EFPY (Without Margios for Instrumentation Errors)

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1 BYRON. UNIT 2 l

PRESSURE AND TEM 7ERATURE LIMITS REPORT 4

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. Table 2,1 -

Byron Unit 2 Heatup and Cooldown* Data Points at 12 EFPY" (Without margins for instrumentation errors)

HEATUP CURVES COOLDOWN CURVES

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180 F HeetvP Test Curve Cettecemy Unit Limit stee8f State 25 000 F SO 000 F 180 000 F T

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P T

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191 2000 60 0

80 0

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to 0

80 821 212 849 212 2485 80 821 80 See to

$38 to 437 SS 821 212 692 SS 821 SS 801 SS 551 SS 452 70 821 212 878 70 821 70 814 70 885 70 488 75 821 212 889 75 421 75 821

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$21 80 597 80 SOS

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100 821 212 883 100 821 100 821 100 821 100 599 105 821 212 870 108 821 105 821 10!

821 105 821 110 821 212 679 110 821 110 821 110 821 110

$21 115 821 212 890 115 421 115 421 115 421 its 821 120 821 212 703 120 821 120 821 120 821 120 821 125 821 212 719 125 821 128 421 125 421 125 821 130 821 212 738 130

$1. I 1 821 130 821 130 821 135 821 212 758 135 821 135 821 135 i't21 135

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140 821 212 781 140 821 140 821 140 821 140 821 145 821 212 807 145 821 145 821 145 821 145 821 150 821 212 435 150 821 150 821 150 -

$21 150 821 150 738 215 888 150 1045 150 1028 150 tote 150 1008 155 758 220 900 1

155 41087 155 1074 155 1088 155 1088 180 781 225 937 l

150 1132 180 1123 100 1120 180 1131 185 807,

230 977 i

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~1521 300 2020 235 2400 240 1807 30$

2143 245 1899 310 2275 250 1794 315 2417 255 190$

280 2020 285 2143 270 2275 275 2417

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Heatup and Conidown data includes vessel flange requirements of 150 F and 621 psig per 10CFR50, Appeaduc G.

• • For each cooldown rate, the steady state pre 55ure values shell govern the temperature where no allowable pressure values are provided

BYRON UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

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80 100 150 200 250 300 354 400 AUCTIONEERED LOW RCS TEMPERATURE (DEG, F)

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

t HYRON o UNIT 3

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PRESSURE AND TEMPERATURE LIMITS REPORT I

Table 2.2 Data Points for Byron Uniti Mksimum Allowable PORY Setpoints for the LTOP System Applicable for the First 12 EFPY' PCV-455A PCV-456

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

(PSIG)

RCS TEMP. (DEG. F)

(PSIG)

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50 497 50 514 70 497 70 514 100 497 100 514 120 446 120 462 150 446 150 462 200 446 200 462 250 587 250 004 300 587 300 6 04 350 587 350 604 450 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 S trveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determiae changes in material properties. He removal schedule is provided in Table 3.1. Also, ine 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,"Reactot Vessel Radiation Surveillance Program." he material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTu, which is determined in accordance with ASME Section III, ND 2331. The empirical relationship between RTm and the fracture teughness of the reactor vessel ste:1 is developed in accordance with Appendix 0," 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 o UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Byron Unit 2 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time ("

Estimated Capsule (Degrees)

Factor (EFPY)

Fluence (n/cm')

U 58.5' 3.96 1.15 (Removed) 3.996 x 10" W

121.5*

3.89 4.634 (Removed) 1.211 x 10" X

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

301.5*

3.89 Wall (*?)

3.288 x 10" V

61.0' 3.64 Standby l

Y 241.0' 3.64 Standby (a) Effective Full Power Years (EFPY) from plar I stanup.

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

(c) Derived from Tabe C 2 of WCAP 14824, Rev. 2 (Reference 2, which is the Anachment to this repon).

10

j BYRON. UNIT 2

^

PRI'SSURE AND TEMPERA'l URE LIMITS REPORT f

i 4.0 Supplemental Data Tables t

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The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96 03. Some cd 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 i

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).

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

l Table 4.6 provides RTnsvalues for Byron Unit 2 for 48 EFPY.

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

FFW Measured FF*ARTm (FF)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 -1 (Axial)

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

66.037 3.332 Chemistry Factor (0 = 66.037 + 3.332 = 19.8'F Byron 1 Weld Metal WF 336*

U 3.72x10" 0.727 0

0 0.00 0.529 X

1.39x10" 1.091 35 105")

114.56 1.190 Byron 2 Weld Metal WF-447("

U 3.996x10" 0.746 0

0 0

0.557 W

l.211x10" 1.053 30 90"'

94.77 1.110 Sum:

209.33 3.386 Chemistry Factor (* = 209.33+3.386= 61.8'F (a) FF = FRence Factor a-f"*** 0 (b) B> ton Unit 1 ARTm values were obtained from the surveillance Capsule X analysis (WCAP 13880, Reference 10).

The ByTon Uni I capsule fluence values were recalculated using the ENDF/B V scattering cross sections in 1994 and t

are documented in WCAP 14044 (Reference 8).

(c) Dyron 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 = E (FF'ARTm)/I((FF)')

(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 (68.007.0 = 2.5); However, for conservatism, ratio of 3.0 was used.

12

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DYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2 Byron Unit 2 Reactor Vessel Material Properties Material Description Cu(%)N Ni (%)(*)

Chemistry Initial FactorM RT m ('F)N Closure Head Flange Not 0.74 OM Reponed Vessel Flange Not 0.73 30M heported Inter. Shell Forging 0.01 0.70 20.0

-20 49D329/49C2971-1 Lower Shell Forging 0.0a 0.72 32.2

-20 49D330/49C2981-1 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 I

12).

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

13

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HYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 I

Summary of Hyron Unit 2 Adjusted Reference Temp:istures (ARTS'y at 1/4T and 3/4T Locations for 12 EFPY F

12 EFPY Material Description 1/4T ART ('F) 3/4T ART ('F)

Intermediate Shell Forging 12.1 1.7 49D329/49C2971-1 (R.G.1.99 Position 1 ()

Using credible surveillance 31.7 14.9 capsule data (RG Position 2(*))

Lower Shell Fo ging 11.8 1.5 49D330/49C2981 1 (RG Position l))

Circumfriential Weld 131.9 98.9 bi-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 I and 2 (Reference 12).

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

14

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BYRON UNIT 2 l

t PRESSURE AND TEMPERATURE LIMITS REPORT l

Table 4.4 Byron Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 12 EFPY at the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data)

Parameter Values Operating Time 12 EFPY '

l LocationN 1/4T ART 3/4T ART i

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Chemistry Factor, CF (*F) 68.0 68,0 Fluence (f), n/cm' 4.94x10" 1.78x10"

[

(E>l.0 Mev)*

Fluence Factor, FF 0.803 0.542 ARTum= CFxFF(*F) 49.63 33.50 Initial lit ym,1('F) 10 10

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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 fs(E>l.0 Mev) = 8.22x10 at 12 EFPY, WCAP 14940 (Reference 14).

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

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1 HYRON UNIT 2 PRESSURE AND TEMPERATURE 1.lMITS REPORT Table 4.5 RTna Valuer for flyron Unit 2 for 32 EFPY M

l RTune ARTns

RTn, CF p.i pp, 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 4

Using Surveillance 19.8 2.192 1.213 34.0

-20 24.0 38.0 Capsule Data"'

intennediate Shell Forging 20.0 2.192 1.213 24.3 20 24.3 28.6 MK 24 2 Circurnference Weld 68.0 2.192 1.21 I 56.0 10 82.5 148.5 Metal WF447 Using Surveillance 61.3 2.192 1.213 28.0 10 75.0 113.0 Capsule Data (3) 2.192x10 n/cm (Epl.0 Mev) for 32 EFPY from flyron 2 P t s report. WCAP 14054 (Reference 9) 3 (b) FF (Fluence Factor) = l'"'*'** 0 (c) Cakulated using a CF based on suncillance capsule data per RG l.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.

16

._-____=.___-_.-_m_---____-._-___------_---_.2_-

4 HYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6 RTn, Yalues for Ilyron Unit 2 for 48 EFPY M

RTm,mn ARTns

RTn, CF p.,

pp63 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 d

Using Surveillar.cc 19.8 3.288 1.312 34.0

-20 26.0 40.0 Capsule Data"'

Intermediate 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.f, 3.288 1.312 28.0 10 81.1 119.1 Capsule Data (a) 2.192x10" n/cm (E>l.0 Mev) for 32 ETPY from Dyron 2 PTS report. WCAP 14054 (Reference 9). The following 8

calculation provides the 48 F/PY Oueni e value:

8 2.192x10" + ((2.19.!x10".3 tkx10"y32 4.634 EFPY)) * (48 32 EFPY) = 3.288x10" n/cm (b) FF (Fluence Factor) = f* 8"** 0 (c) Calculated using a CF based on w.c ihnce capsule data per RO 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.

17

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

November,1997.

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 Prvgram," Terek, E., et al.,

October 1989.

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-10398," Commonwealth Edison Company, Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program," Singer,IlR., December 1983.

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 Shoch for Byron Unit 2," Peter, P.A., August 1994.

10. 10 CFR Part 50, Appendix 0," Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, dated December 19,1995.

I1. 10 CFR 50.61," Fracture Toughness Requirements for Protection Against Pressurized Thennal 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.

I8

HYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

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. Wr AP-14940,"11yron Unit 2 ileatup and Cooldown Limit Curves for Normal Ope.ation," Laubham, T. J., et al., October,1997.

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.

19

.. o,

\\

ATTACHMENT WCAP -24824, REVISION 2, BYRON UNIT 1 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION AND SURVEILLANCE WELD METAL INTEGRATION FOR BYRON AND BRAIDWOOD w

r ma

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section _

Page 1.0 Introduction 1

2.0 Operating Limits I

k 2.1 RCS Pressure and Temperature (P/T) Limits 1

2.2 Low Temperature Overpresure Protection (LTOP) System Setpoints 2

2.3 LTOP Enable Temperature 2

2.4 Reactor Vessel Boltup Tempe:ature 3

2.5 Reactor Vessel Minimum Pressurization Temperature 3

3.0 Reactor Vessel Material Surveillance Program 9

4.0 Supplemental Data Tables 11 5.0 References 18 M

Attachment WCAP-14824, " Byron Unit i Heatup and Cooldown Limit Curves for Normal Operation and Sune!!!ance Weld Metal Integration for Byron and Braidwood," Revision 2, Novembo.1997.

i-

s BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITT, REPORT List of Figures Figure Page 2.1 Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4

{

100'F/hr) Applicable for the First 16 EFPY (Without Margins for Instrementation Errors) 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up 5 to 100 *F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors)

\\

}

2,3 Braidwood Unit 1 Maximum Allowable Nominal PORY Setpoints for the Low 7

F Temperature Overpressure Protection (LTOP) System App::pW for the First 16 EFPY

\\

L k

r

=-

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT.

List of Tables Table Page 2.1 Braidwood Unit 1 Hntup and Cooldown Data Points at 16 EFPY 6

(Without Margins for Instrumentation Errors) 2.2 Data Points for Braidwood Unit 1 Maximum Allowable PORV Setpoints 8

for the LTOP System Applicable for the First 16 EFPY 3.1 Braidwood Unit 1 Capsule Withdrawal Schedule 10 4.)

Braidwood Unit 1 Calculation of Chemistry Factors Using Surveillance 12 Capsule Data 4.2 Braidwood Unit 1 Reactor Vessel Material Properties 13 4.3 Summary of Braidwood Unit 1 Adjusted Reference Temperatures (ARTri) at 14 the 1/4T and 3/4T Locations for 16 EFPY-4.4 Braidwood Unit 1 Calculation of Adjusted Reference Temperatures (ARTS) at 15 16 EFPY at the Limiting Reactor Vessel Material Weld Metal WF562 (Based on

'Survei: lance Capsule Data) 4.5 RTns Values for Braidwood Unit I for 32 EFPY 16 4.6 RTns Values for Braidwood Unit I for 48 EFPY 17 iiF

~. -

BRAIDWOOD - UNIT 1 PRESSURE AND TEMI'EP.ATURE LIMITS REPORT 1.0 Introduction Reference to Technical Specifications (TS) numbers are given in both the Braidwooc' Station current Technical Specifications (CTS) and Improved Tect nical Specifications (ITS). The CTS number ir, presented fir:1, followed by the ITS number in brackets [ ).

This PTLR for Unit I has been prepared in accordance with the requirements of TS 6.9.1.11/[ITS-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.

[ITS 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 were developed using a methodology specified in the Technical Specifications.

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

a. Use of ENDF/B-IV neutron transpcrt cross-section library and ENDF/B-V dosimeter reaction cross-sections,

b. Optional use of ASME Code Section XI, Appendix G, Article G-2000,1996 Addenda (not used for Braidwood Unit 1 P-T curves),

c. Use of ASME Code Case N-514, and

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

WCAP-14824, Revision 2, is included as an attachment for reference. WCAP-14824, Rev. 2 contains the P/T curves for Byron Unit 1, along with the weld metal data integratior ir Byron and Braidwood Units 1 and 2 and the Byron /Braidwood fluence methodologyjustifica

z. lor ENDF/B-VI cross sections.

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.9.1/ [iTS-LCO 3.4.3])

2.1.1 The RCS temperature rate-of-change limits defined in Reference 7 are:

a. A raaximum heatup of 100 F ir, any 1-hour period,

b. A maximum cooldown of 100*F in any 1-hour period, and

c. A maximum temperature change ofless t.'en or equal to 10 F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

p

BRAIDWOOD - UNIT I P.RESSURE AND TEMPERATURE LIMITS REPORT

- Operating Limits (Continued) 2.1.2 The RCS Pff limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defined in Reference 7. Consistent with the methodology described in Reference 1, the RCS P/T limits for heatup and cooldown showm in Figures 2.1 and 2.2 are provided without margins for instrumeat error. In determining compliance with Figures 2.1 and 2.2 and Table 2.1, instrument uncertainties need not be considered since appropriate station operating procedures ensure that the limits contained in the figures and table are not exceedea. 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 Pff limits for core operation (except for low power physics testing) are that the reactor vi.ssel 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 3.4.9.3/

[ITS-LCO 3.4.12]).

The power operated relief valves (PORVs) shall e.ch 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 LTOP setpoints are based on P/T limits which were established by 10 CFR 50, Appendix G without allowance for instrumentation erroi in accordance with the methodology described in Reference 1. The LTOP PORY 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 15).

The TS required enable temperature for the PORVs shall be 2 350 F RCS temperature (Braidwood Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS r mperature of 350*F and below and disarming of LTOP for RCS temperatu:e 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. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

2

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Operating Lhnits (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 the RCS vented to atmosphere (Reference 7).

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 1 l

PRESSURE AND TEMPERATURE LIMITS REPORT l

MATERIAL PROPERTY BA315 LIMTTING MATERIAL WELD METAL LIMITING ART VALUES AT 16 EFPY:

1/4 76.6*F 3/4 65.4*F 2500 4

l i

1 e

I Ic i'

ii, i e t

• I f

I

/

l' 2250 m

m LEAK TEST L1MIT I

J l

t

/

/

l. l

~

as 2000 ll'

/

/ " '

=

.i i.

I

(

g l t f

f f I

I w

I

/

1750 UNACCEPTABLE OPERATION I

I' CJ I

i i

f f

y a

1500

/

/

l m

~

m s r/

I HrATur nATr

/ ;'

/

AccEPTAstE o

1250 ll' UP TO 400 t/Hr.

/

/

OPERATION a

j i i f

/

i *

,w i i :,

i i e

t i

/!

/

r i i

1000 'l l l ll ll l

/

=

., e i,

i i !

> + t

/

V g

e a it ii s

/

/

750

, e' ' '

i, i

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n e i e r O

i t e s

. i,

500 l

lll l

e e i

,ii i

1_

e e i t i e N

i 6 e i,e

. i X

i 250 CRITICALITT LikIT RASED OM I ' '

IM5ERVICE HYDROET& TIC TEST i

TEMPERATURE late FI FOR THE i

! a i!

i e it i e e **

SERVICE PERIOD UP 70 1.

0 EFPY l

i 0

0 5'O Id0 150 200 2$0 380 3$0 480 4$0 500 Indicated Temperature (Deg.F)

Figure 2.1 Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 'F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors)

'4

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT -

MATERIAL PROPERTY BASIS LIMITING MATERIAL WELD METAL I,IMITING ART VALUES AT 16 EFPY:

. t/4.t. 76.6*F

3/4-t, 65.4*F

.2500 l

}

sessessessase 8

4!

l J

2250.

j' l ;'

1 m

- ing._

J 1

2000

/

l f

00 m.

l

)

i i

/

l750

_'_uN A c c r P T A s t r-i cu-OPrRATioN J

I

/

1500

/

l sn

. cc

/

o 1250.

AccrPTAstr b

i fl !

oPrRATION

1. ',h 1

i r

1000 i

g.

2r i

s,,,

W e

f a

~~

o AP a

6

- C00LDOWN jyj

===8

- RATES 7

l,'

l

=

750 -- r/ur.

g.

it i

6 o

cx,-

'e

-500

'S

',f l l,

ll

,,l

~~

=

q i

6 250 4,,

0 0

50 Id0 150 2 d O' 250 300 350 400 450 500 I n d i c a t e-d Temperature (Deg..F)

Figure 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 *F/hr) Applicable for the First 16 EFP.Y (Without Margins for Instrumentation Errors)

~5 i>

n m.un www-w.:..

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 BRAIDWOOD Unit 1 Heatup* and Cooldown* Data Points at 16 EFPY" (Without margins for instrumentation errors)

HEATUP CURVES COOLDOWN CURVES l 100 F Heatup Leak Test curve Criticahty Limit UmH Steady State 26 DEO F So DEO F 100 DEO F P

T P

T I

P T 4 P T I P

T I P

T 4 P T

80 0 00 210 l 0 00 188l2000 60 1 0 00 60 1 0 00 60 1 0 00 60 f 0 00 60 565 09 210 1 611 83 210l2495 80 ! 620 27 60 l 577 45 80 534 28 60 1 446 98 65 565 09 210 l 597.56 I

65 1 621.00 65 1 590 68 65 548 52 85 i 463 79 70 565 09 210 ! 585 60 i

70 f 621.00 70 1 605 03 70 1 563 98 70 ' 481.93 565 09 210 i 576 77 I

75 l 62100 75 l 620 51 75 1 560 67 75 1 501 49 75 i

80 585 09 210 1 570 35 I

80 ! 621 00 80 i 621 00 80 l 598 51 80 I 522 88 85 84509 210 6 586 61 1

85 l 621.00 85 l 62100 85 I 617 90 85 6 545 50 90 565 09 210 j 565 09 l

90 1 621 00 90 1 621 00 90 1 621.00 90 1 570.23 596 83 95 585 09 210 1 565.87 l

95 1 621.00 95 621.00 95 1 621 00 95 i

100 565 87 210 f 568 89 i

100 l 62100 100 621 00 100 l 621.00 1{l 621 00 105 568 89 210 i 573 56 l

105 621.00 105 1 621 00 105 1 621 00 e

621.00 110 573 58 210 i 580 30 i

110 1 621.00 110 1 621 00 110 1 621.00 110 t 62140 115 580 30 210 1 585 84 110 I 795 92 110 1 786 92 110 l 739.27 110 l 890 04 120 1 588 84 210 I 599.36 115 ! 821 55 115 7e4 59 115 1 769 53 115 l 726.2 125 599 36 210 1 811.78 120 I 849 00 120 824 45 120 } 801 97 120 1 785 12 130 611.78 210 i 626 07 l

125 1 818 42 125 856 54 125 1 836 87 125 I 807.07 135 626C7 210 i 64216 l

130 1 91025 130 890 97 130 1 874 41 130 1 852.23 140 642 16 210 I 860 36 1

135 l 944 34 135 l 928 00 135 i 915 0",

135 } 900 91 145 660 36 210 1 680 59 l

140 i 980 89 140 i 967.79 140 l 958 57

'40 1 953.33 150 680 59 210 1 702 80 1

145 1 1020.15 145 1 1010 84 145 11005 42 145 l 1009 81 155 i 702 80 210 ! 727.33 i

150 l 1062.35 150 i 1056 88 150 11055 76 l

727.33 210 1 754 07 i

155 t 1107.92 155 j 1106.38 l

l 160 i

165 754 07 210 t 783 17 160 11156 42

}

I 17:

78317 215 I 814 98 i

165 ! 1208 78 l

175 814 98 223 I 849 37 I

170 I 1265 05 I

I 175 i1325.37 i

i i

180 849 37 225 ! 886 54 185 886 f _

230 ! 926 73 1

180 1 1390 04 l

l l

190 926 73 235 i 970 11 185 l 1459 41 I

I i

195 970 11 240 1 1016 91 190 1 1533 55 i

1 I

200 1016 91 245 t 1067.33 195 i 1613 49 l

205 1067.33 250 i 1121 63 200 11699 01 l

l 210 1121 63 255 1 1180 01 205 ' 1790.55 l

t i

215 1180 01 260 l 1242 62 210 i 1888 81 I

I i

220 1242 62 285 i 1309 84 215 619C3 61 I

I 225 1309 84 270 ' 1382 03 220 : 2105 6e i

230 1382 03 275 1 1459 45 225 2225.77 i

235 1459 45 280 t 1542.27 230 ! 2353 75 6

i i

240 1542.27 285 : 1630 97 4

245 1630 97 290 i 1728 05 i

I i

250 1726 05 295 1 1827 80

+

l i

255

'827 80 300 ' 1936 51 I

i 260 193651 305 ; 2052 39 565 2052 39 310 ! 2176 33 i

i 270 2176 33 315 2308 42 i

i 275 2308 42 320 ' 2449 09 1

280 2449 09 i

• Heatup and C,ooldown data include vessel flange requirernents of 110 F and 621 psig pr 10CFRSO, Appendix G

" For e8ch cooldown rate, the steady-state pressure values shall govern the ternperature where no allowab 8 pressure values are provided 6-

GRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT i i i I i l I Il lII i

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't i i i ! i i i i i l i i j i i i j -

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! i 1

t l

! ! ! i iil i i i! ! l l 1 I i i l 0

so 100 160 200 260 300 sao AUCTIONEERED LOW RCS TEMPERATURE (DEG, F)

Figure 2.3 Braidwood Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY 7

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Braidwood Unit ! Maximum Allowable PORV Setpoints for the LTOP Sy5 tem Applicable for the First 16 EFPY PCV-455A PCV-456 (1TY 0413M)

(1TY-0413P) b AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. F)

(PSIG) 8tCS TEMP. (DEG. F)

(PSIG) 50 497 50 513 70 497 70 513 iOO 497 100 513 110 497 110 513

(

160 497 160 513 200 616 203 634 250 603 250 619 300 588 300 604 350 588 350 604 450 2350 450 2350 Note: To determine maximum allowable fift setpoints for RCS Pressure and RCS Temperatures greater than 350 F, linearly interpolate betwetn the 350 F and 450 F data points shown above.

r n

E

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0

- Reactor Vessel Material Sun'e.illance 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, The results of these anslyses 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 standards utilize the reference nil-ductility temperature, RTmr,

which is determined in accordance with ASME Section III Paragraph NB-2331. The empirical relationship between RTer nd the fracture toughness of the reactor vessel steel is developed in a

accordance with Appendix G," Protection Against Non-Ductile Failure," to Section XI of the AShE Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E18F 82.

i

BRAIDWOOD-UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Braidwood Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time'"

Estimated Capsule 2

(Decrees)

Factor (EFPY)

Fluence (n/cm )

U 58.5*

4.03 1.10 (Removed) 3.814 x 10'8 X

238.5' 4.03 4.234 (Removed) 1.144 x 10

W 121.5' 4.03 7.95 (EOL Wall) 2.239 x 10'*

11.93 (1.5 EOL Z

301.5*

4.03 Wall'4) 3.359 x 10" __

V 61.0' 3.73 Standby Y

241.0*

3.73 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-3 of WCAP-14824, Rev. 2 (Reference 2, which is the Attachment to this report).

10

BRAIDWOOD - UNIT 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 4.3 provides a summary of the Braidwood Unit 1 adjusted referer.ce temperatures (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 Braidwood Unit I reactor vessel material -Weld Metal WF562 (Based on Surveillance Capsule Data).

Table 4.5 provides the RTns alues for Braidwood Unit 1 for 32 EFPY.

v Table 4.6 provides the RTns alues for Braidwood Unit i for 48 EFPY.

v It

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 1 Calculation of Chemistry Factors Usine Surveillance Capsule Data Fluence Material Capsule (ri/cm,

FF")

Measured FF*ARTmr (FF)2 2

E> l. 0 ARTsm Mev), f Lower Shell Forging U

3.814 x10" 0.733 5

3.666 0.538 24-3 (Tangential)

X 1.144 x10" 1.038 30 31.127 1.077 Lower Shell Forging U

3.814 x10" 0.733 0

0 0.538 24-3 (Axial)

X 1.144x10" 1.038 25 25.939 1.077 Sum:

60.733 3.228 Chemistry Factor d' = 60.733 + 3.228 = 18.8'F Braidwood 1 Weld U

3.814x10" 0.733 10")

7.333 0.538 Metal WF562*'

X 1.144x10" 1.038 25")

25.95 1.077 Braidwood 2 Weld U

3.933x10" 0.741 0

0 0.550 Metal WF562")

X 1.126 x I0" 1.033 20"'

20.66 1.067 Sum:

53.943 3.232 Chemistrv Factor "' = 53.943+ 3.232 = I6.7'F (a) FF = Fluence Factor = f*2" "'*"'

(b) Braidwood Unit 1 ARTxor values were obtained from WCAP-14243 (Reference 7).

l (c) Braidwood Unit 2 capsule fluence, FF, and ARTxo1 values obtained from WCAP-14970 (Reference 12)-

(d) Chemistry Factor = I (FF*ARTxoT) / I ((FF) 2)

(e) ARTsot per Ratio Procedure of 10 CFR 50.61 (Reference 10)is not affected, since Ratio = 1.0. See Table B-4 of WCAP 14824, Rev. 2 (Reference 2).

12 l

f

BRAIDWOOD - UNIT 1 PRESSURE AND TEh!PERATURE LIMITS REPORT Table 4.2 Braidwood Unit 1 Reactor Vessel Material Pro perties Chemistry Initial Material Cu (%)")

Ni (%)")

Factor'"

RT mr ('F)'"

Description Closure Head

-20'"

Flange

- 10"'

Vessel Flange Lower Shell 0.04 0.74 26.0

-20 Forging 24-3 Intermediat-Shell 0.05 0.73 31.0

-30 Forging 24 2 Weld Metal (i.03 0.67 41.0 40 WF562 (a) Chemist y Factors are calculated from Cu and Ni values per Regulatory Guide 199, Position 1 (Reference 11).

(b) initial RTer alues are measured, WCAP-14243 (Reference 7) v (c) Closure head and vessel flange Initial RTer values are used for considering flange requirements for the heatup/cooldown curves, WCAP-14243.

13 1

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Braidwood Unit 1 Adjusted Reference Temperatures (ARTS) at the 1/4T and 3/471.ocations for 16 EFPY")

16 EFPY Material Description 1/4T ART (*F) 3/4T ART (*F)

Lower Shell Forging

-26.2 12.1 24-3 Using credible surveillance capsule data 13.4*)

3.2*'

7 (RG Position 2))

Intermediate Shell 25.1 8.2 Forgi..g 24-2 (RG Position 1)

Circumferential Weld 112.9 90.5 (RG Position 1)

~

Using credible 76.6*'

65.4*'

surveillance capsule data (RG Position 2))

(a) Calculated using a chemistry factor based on Pegulatory Guide (RG) 1.99, Position 2 (Reference 11).

(b) These ART values were used to generate the Braidwood Unit I heatup and cooldown curves, WCAP-14243 (Reference 7).

(c) Based on evaluation of WCAP-14243; however, applicability date has been increased to 27.9 EFPY per evaluation in WCAP-14824, Rev. 2-Appendix D: Weld Mctal Integration for Braidwood Units 1 and 2 (Reference 2).

14 l

l

l BRAIDWOOD - UNIT 1 1

PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 Braidwood Unit 1 Calculation of Adjusted Reference Temperatures (ARTa) at 16 EFPi*' at the Limiting Reactor % ssel Mcterial Weld Metal WF562 (Based on Surveillance Capsule Data)

Parameter Values r

Operating Time 16 EFPi*'

Location"'

1/4T ART 3/4T ART fhemistry Factor, CF ('F) 20.6 20.6 Fluence (f), n/cm' 0.73x10" 2.43x10" (E>1.0 Mev))"'

Fluence Factor, FF 0.889 0.616 ARTmr= CFxFF ('F) 18.31 12.70 Initial RT mr, I (*F) 40 40 Margin, M ( F) 18.31 12.70 ART = 1+(CF*FF)+M, *F 76.6 65.4 per RG 1.99. Rwision 2 (a) Fluence, f, is based upon f.s(E>l.0 Mev) = 1.120x10" at 16 EFPY, WCAP-14241 (Reference 3).

(b) Based on evaluation of WCAP 14243; however, applicability date has been increased to 27.9 EFPY per evaluation in WCAP-14824, Rev. 2 Appendix B: Weld Metal Integration for Braidwood Units 1 and 2 (Reference 2).

(c) The Braidwood Unit I reactor vescel wall thickness is 3.5 inches at the beltline region.

t 15

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT T

Table 4.5 RTers Values for Braidwood Unit 1 for 32 EFPY CF p.

ppm M

RTane ARTns RTns Material

(*F)

(*F)

(*F)

(*F)

(*F)

Lower Shell Forging 26.0 2.239 1.218 31.68

-20 31,68 43.4 MK 24 3 Using Sun'eillance 18.8 2.239 1.218 17.0

-20 22.90 19.8 Capsule Data"'

~

Intermediate Shell Forging 31.0 2.239 1.218 34.0

-30 37.77 41.8

~

24 2 Circumferential Weld 41.0 2.239 1.218 49.95 40 49.95 ! 139,9 Metal WF562 Using Sun eillance 16.7 2.239 1.218 20.34 40 20.34 80.7 Capsule Data")

(a) 2.239x10 n/cm'(E>l 0 Mev) for 32 EFPY from Braidwood 1 PTS report, WCAP-14242 (Reference 8).

(b) FF (Fluence Factor) = fo.2w io w,o r

(c) Calculoted with CF based c. surveillance capsule data per RG 1.99, Position 2 (Reference 11).

16

BRAIDWOOD - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6 RTns Values for B aidwood Unit 1 for 48 EFPY CF M

RTmrrm)

ARTns RTns f.>

ppm Material

('F)

('F)

('F)

('F)

('F)

Lower Shell Forging 26.0 3.358 1.317 34.0

-20 34.25 48.3 MK 24 3 Using Surveillance 18.8 3.358 1.317 17.0

-20 24.76 21.8 Capsule Data"'

Intermediate Shell Forging 31.0 3.358 1.317 34.0

-30 40.83 44.8 24-2 C'rcumferential Weld 41.0 3.358 1.317 54.00 40 54.00 148.0 Metal WF562 Using Surveillance 16.7 3.358 1.317 21.99 40 21.99 84.0 Capsule Data")

(a) 3.358x10 n/cm (E>l.0 Mev) for 48 EFPY from Braidwood 1 PTS repon, WCAP-14242 2

(Reference 5).

(b) FF (Fluence Factor) = f"#'*80 (c) Calculated with CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 11).

17

BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT.

5.0 References 1.

Andrachek, J.D., et al., WCAP-14040-NP-A, " Methodology used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1996.

2.

Grendys, P.A., WCAP-14824, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood,"

Revision 2, November 1997 3.

Peter, P.A., et al., WCAP-14241, " Analysis of Capsule X from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program,"

March 1995.

4.

Terek, E., et al., WCAP-12685, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program," August 1990.

5.

Westinghouse Letter to Conunonwealth Edison Company, CCE-95 186, "Braidwood Unit 1 LTOPS Setpoints Based on 16 EFPY Pfr Limits," June 5,1995.

6.

Yanichko, S.E., et al., WCAP-9807, " Commonwealth Edison Company, Braidwood Station Unit i Reactor Vesse! Surveillance Program," February 1981.

7.

Peter, P.A, WCAP-14243, "B,raidwood Unit i Heatup and Cooldown Limit Curves for Normal Operation," March 1995.

8.

Peter, P. A., WCAP-14242, " Evaluation of Pressurized Thermal Shock for Braidwood Unit 1," March 1995.

9.

Lippencott, E.P., WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation," April 1994, 10, 10 CFR 50.61, " Fracture Toughaess Requirements for Protection Against Pressurized Thermal Shock Events," (PTS Rule) January 18,1996.

I1. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revisicn 2, May 1988.

12. Laubham, T. J., WCAP-14970,"Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," October 1997 and Erratta Sheet (Westinghouse Letter CAE 210, CCE 97-289).

18

1; i

I' BRAIDWOOD - UNIT l' PRESSURE AND TEMPERATURE LIMITS REPORT References (Continued)

13. Comed Ca'culation BRW-96-9061/BYR 96-293, " Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 1 Original Steam Generators and Replacement Steam Generators)," Revision 0.

14. Comed Nuclear Fuel Services Department, NDIT No. 960194, " Maximum Allowable LTOPS PORV Setpoints for Braidwood Unit I with RSGs," Revision 2.

15. Westinghouse Letter to Commonwealth Edison Company, CAE 97-211, CCE 290,

" Comed Byron and Braidwood Units 1 and 2 ATmetal Evaluation for Commonwealth Edison," November 7,1997.

T 4

19 i

ATTACHMENT 1 WCAP-14824, REVISION 2, BYRON UNIT 1 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION AND SURVEILLANCE WELD METAL INTEGRATION FOR BYRON AND BRAIDWOOD 20 -

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- BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT i

Table of Contents

- Section Page 1.0 - Introduction 1

.- 2.0 Operating Limits 1-2.1 RCS Pressure and Temper ture (PR) 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 9

4.0 Supplemental Data Tables 11 5,0 References 18 Attachment WCAP-14824, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron and Braidwood," Revision 2, November 1997.

i-

• i BRAIDWOOD - UNIT 2.l PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Firure Page 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4

i 100.*F/hr) Applicable for the First 12 EFPY (Without Margins for Instmmentation Errors) 2.2 Braidwoc,d Unit 2 Reactor Coolant System Cooldowa Limitations (Cooldown Rates up 5 to 100 'F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) 2.3

_ Braidwood Unit 2 Maximum Allowabic Nominal PORY Setpoints for the Low 7

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

'Es _

CRAIDWOOD. UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tabler Table -

Page 2.1 Braidwood Unit 2 Heatup and Cooldown Data Points at 12 EFPY

> 6

-(Without Margins for Instrumentation Errors) 2.2 Data Points for Braidwood Unit 2 Maximum Allowable PORY Setpoints 8

for the LTOP System Applicable for the First 12 EFPY 3.1 Braidwood Unit 2 Cap.me Withdrawal Schedule 10 4

-4.1-Braidwood Unit 2 Calculation of Chemistry Factors Using Surveillance 12 Ct.psule Data 4.2 -

Braidwood Unit 2 Reactor Vessel Material Properties 13 4.3 Summary ofBraidwood Unit 2 Adjusted Reference Temperatures (ARTS) at 14 the 1/4T and 3/4T Locations for 12 EFPY 4.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 15 12 EFPY at the Limiting Reactor Vessel Material Weld Metal WF562 (Based on Sun eillance Capsule Data) i 4.5 RTus Values for Braidwood Unit 2 for 32 EFPY 16 4.6 RTm Values t'or Braidwood Unit 2 for 48 EFPY 17 tii

• i HRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction Reference to Technical Specifications (TS) numbers are given in both the Braidwood 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 I has been prepared in accordance with the requirements of.

TS 6.9.1.11/[ITS-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 /femperature Limits; and LCO 3.4.9.3 Overpressure Protection Systems.

[ITS-LCO 3,4.3 RCS Pressure and Temperature (Pff) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System].

2.0 Operating Limits The FTLR limits were developed using a methodology specified in the Technical Specifications.

The methodolcgy listed in WCAP-14040-NP-A (Reference 1) was used with four exceptions:

a Use of ENDF/B-IV neutron transport cross-section library and ENDF/B-V dosimeter reaction cross-sections,

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

c. Use of AShE Code Case N-514, and

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

WCAP-14824, Revision 2, is included s an attachment for reference. WCAP-14824, Rev,2 contains the D/T curves for Byron Unit 4, along with th: weld metal data integration for Byron and Braidwood Units 1 and 2 and the Byron /Braidwood fluence methodologyjustifica: ion for ENDF/B-VI cross sections.

21 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.9.1/ [ITS-LCO 3.4.3])

2.1.1 The RCS temperature rate-of-change limits defined in Reference 7 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 j.

inseivice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

I-

BRAIDWOOD - UNIT 2 1

PRESSURE AND TEMPERATURE LIMITS REPORT Operating Limits (Continued) 2.1.2 The RCS P/T limits for heatup, insenice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defmed in Reference 7. 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. In determining compliance with Figures 2.1 and 2.2 and Table 2.1, instrument uncertainties need not be considered since appropriate station operating procedures ensure that the limits contained in the figures and table are not exceeded. 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 insenice 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 Setpoir.ts (LCO 3.4.9.3/

[ITS-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 14.

The LTOP setpoints are based on P/T limits which were established by 10 CFR 50, Appendix G without allowance for instrumentation error in accordance with the methodology described in Reference 1. The LTOP PORY maximum lift settings shown in Figure 2.3 and Table 2.3 account for appropriate instrument error.

2.3 LTOP Enable Temperature The as taalyzed LTOP enable temperature is 200 F (Reference 15)

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

(Braidwood Unit 2 procedures governing tne 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 PORY segmer : 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.

t 2

's BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Op: rating 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 7).

2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification)

The minimum temperature at which the Resctor Vessel may be pressurized (i.e., in an unvented concDion) 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 TEMPERATURE LIMITS REPORT Material Property Basis LIMITING MATERIAL: CIRCUMFREtGIAL WELD (using sun'. capsule data)

LIMITING ART VALUES AT 12 EFPY:

1/4T,66.9'F 3/4T, 58.I'F 2500 i.,......

t I

i g

• Ii i

l' 2250 M

/,

'l

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gn LEAK TEST LIMIT 7I

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r f

f

'I s

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2000

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uMACCEPTABLE OPERATION

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6

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UP To 300 r/Er.

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OPERATION

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INSERVICE RYDROST4"lC TEST TEMPERaTURS dIst F i FOR TWE f

I SERVICE PER!dD UP PO 88.0 EFPY 0.

0 50 100 150 200 250 300 350 400 450 500 Moderator Temperature (Deg.F) l Figure 2.I Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 'F/hr)

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

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Material Property Basis LIMIT!NG MATERIAL: CIRCUMFRENTIAL WELD (using sury capsule data)

LIMITING ART VALUES AT 12 EFPY:

1/4T. 66.9'F 3/4T, 58.I'F 2500

.,. 4 u...-

i i

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fi 1250 ACCEPTABLE 6

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i 88 250~,

N Boltup 4

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g P.

6 i...

0 0

5'O Id0 150 20 2$0 300 3$0 4 0 480 500 Moderator Temperature (Deg.F)

Figure 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100*F/hr)

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

~

~

~

orwvvvi.ivuivivu e

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~

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 BRAIDWOOD Unit 2 Hestup* and Cooldown* Data Points at 12 EFPY"

(Using 1996 Appendix G Methodolgy Without margins for instrumentation errors)

HEATUP CURVES

' COOLDOWN CURVES 100 F Heat 4 Leek Test curve CatacelHy Lindt Undt Stee8y 8tste 28 D80 F 88 DSO F 100 080F T

P T

1 P T I P T i P T i P T i P T + P 80 C

192 1 0 17012000 80 1 0 to 1 0 80 i- 0 80 1 0 to 821 1 92 708 192i2485 80 l 821 to i 821 to I 801 80 ' 511 85 821 192 741 f

85. 821 85 t 821 85 ! 819 85 i 532 70 821-192 729 1

70 I 821 70 l 821 70 1 821 70 1 555

-75 821 192 I 720 I

75 821 75 i 821

- 75 1 821 75 1 579 821 192 I 715 l

80 821 80 l 821 80 821 80 { 805 80 85 821 192 4 713 1

85 821 85 i 821 85 821 85 l 821 90 i 821 192 1 114 20 821 to 821 to 821 90 1 821 95 f'71 192 1 719 95 l 821 95 821 95 I 821 95 i 821 100 821 192 i 728 6

100 l 821 100 421 100 I 821 100 i 821 105 821 192 1 735 1

105 1 821

-105 I 821 105!

821

-105 1 821

'110 821 192 748 i

'10 1 821 110 l 821 110 821 110 4 821 115 : 821 192 783 f

115 i 821 115 l 821 1151 821 115 821 120 ; $21 192 700 1

120 821 120 l 821 120-821 120 821 125 821 192 I 801 l

125 821 125 821 125 821 125 821 130 i - 821 192 1 824 1

130 821 130 821 130 821 130 1 821 135 821 195 ! 850 1

135 821 135 821 135 821 135 I 821 140 821 200 1 879 140 i 821 140 1 821 140 1 821 f

140 700

_.205 911 l

140 1 1138 140 ! 1131 140 l 1128 1

145 1 801 210 [ 948 1

145 l 1187 145 ! 1184 145 1188 i

824 215 l 984 i

150 ! 1240

(-

i 150 155 850 220 ' 1028 6

155 i 1297 l

180 879 225 i 1071 1

180 6 1357 L

I i

185 911 230 1 1120 l

185 1 1423 I

i l

170 948 235 t 1174 170 l 1493 I

I i

175 984 240 t 1231 1F5 l 1589 I

i l

180 1028 245 1 1294 i

180 f 1850 f

I i

185 1071 250 1381 i

185 i 1738 i

l i

190 1120 255 l 1433 190 1 1832 l

l t

195 1174 280 ; 1511 195 1 1933 I

I t

200 1231 285 ! 1595 200 ? 2041 i

I 1

205 1294 270 1886 205 i 2158 i

i 1

210 1381 27f 1783 210 1 2284 i

I 215 1433 280 1888 215 2419 i

220 1511 285 ' 2000 i

t i

i 225 1595 290 2121 t

I 230 1886 295 2251 i

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235 1783 300 ' 2391 i

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250 2121' 4

255 2251 1

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280 2391 I

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• HeStup and C00100wn data inclu0e vessel flange requirements of 140 F Snd 621 psig per 10CFR50. Appendix G l

" For each c00ldOwn rate, the steady-state pressure values shall govern the temperature where no ellowable pressure values are provided.

l:

6 l

GRAIDWOOD - UNIT 2 4

L PRESSURE AND TEMPERATURE LIMITS REPORT

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I ! i I liiiliI i l l l l l l l I i

.t I i i l I II t ' i: it i I II I i l l 1II

!I I II I O

M 100 1H 200 2H 300 3M AUCTIONEERED LOW RCS TEMPERATURE (DEG, F)

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

'(

7

BRAIDWOOD - UNIT 2 s

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Braidwood Unit 2 Madmum 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)

(PalG)

RCS TEMP. (DEG. F)

(PSIG) 50 497 50 513 70 497 70 513 100-497 100 513 120 452 120 469 170 451 170 468 200 614 200 630 250 599 250 615 300 584 300 600 350 584 350 600 450 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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8 1

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Reactor Vesse! Material Surveillance Program The reactor vessel material irradiation surveillance specimens shall be removed and analyzed to determine changes in material properties. Tne removal schedule is provided in Table 3.1. 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 rnost 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 standards ntilize the reference nil-ductility temperature, RTr,w,

which is determined in accordance with ASME Section III Paragraph NB 2331. The empirical relationship between RTum 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.

I 9

r

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORf Table 3.1 Braidwood Unit 2 Capsule Withdrawal Schedule Capsule Vessel Location Capsule La.ad Removal Time"'

Estimated Capsule 2

(Degrees)

Factor (EFPY)

Fluence (n/cm )

U 58.5' 4.00 1.15 (Removed) 3.933 x 10" X

238.5*

4.02 4.215 (Removed) 1.126 x 10" W

121.5*

4.02 7.97 (EOL Wall) 2.199 x 10"N 11,95 (1.5 EOL Z

301.5' 4.02 Wall'")

3.299 x 10" V

61.0*

3.70 Standby Y

t 241.0*

3.70 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-4 of WCAP-14824, Rev. 2 (Reference 2, which is the Attachment to this report).

10

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0

. Suppicmental Data Tables The following tables provide supplemental information on reactor vessel material propenies and are provided to be consistent with Geaeric 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 4.3 provides a summary of the Brsidwood Unit 2 adjusted reference temperatures (ARTS) at I

the 1/4T and 3/4T locations for 12 EFPY.

Table 4.4 shows the calculation of ARTS at 12 EFPY for the limiting Braidwood Unit 2 reactor vessel material -Weld Metal WF562 (Based on Surveillance Capsule Data).

Table 4.5 provides the RTns values for Braidwood Unit 2 for 32 EFPY.

i Table 4.6 provides the RTns alues for Braidwood Unit 2 for 48 EFPY, v

l Il i

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

FF"' -

- Measured FF*ARTm1

- (FF)2 E>l.0 ARTet Mev). f Lower Shell Forging U

3.933 x10

O.741 0

0.000 0.550 50D102/50C97 (Tangential)

X 1.126 x10

l.033 2

3.099 1.067 Lower Shell Forging U

3.933 x10

O.741 5

3.707 0.550 50D102/50C97 (Axial)

X 1.126x10

l.033 35 36.160 1.067 Sum:

42.996 3.234 d

Chemistry Factor ' = 42.996 + 3.234 = 13.3*F Braidwood 1 Weld U

3.814x10

O.733 10"'

7.333 0,538 Metal WF562*'

_X 1.144x10

l.038 25")

25.95 1.077 Braidwood 2 Weld U

3.933 x 10

O.741 0

0 0.550 Metal WF562"'

X l.126 x10

l.033 20'"

2066 1.067 Sum:

53.943 3.232 Chemistry Factor d' = 53.943+ 3.232= 16.7'F (a) FF = Fluence Factor = fo.2a innn (b) Braidwood Unit 1 ARTer alues were obtained from WCAP-14243. (Reference 12).

v (c) Braidwood Unit 2 capsule fluence, FF, and ARTm1 values were obtained from WCAP-14970 (Reference 7).

(d) Chemistry Factor = I (FF* ARTmT) / I ((FF) 2),

(e) ARTer per Ratio Procedure of 10 CFR 50.61 (Reference 10) is not affected, since Ratio = 1.0.

See Table E.-4 ofWCAP 14824, Rev. 2 (Reference 2)

,12

BRAIDWOOD - UNIT 2 PRESStJRE AND TEMPERATURE LIMITS REPORT Table 4.2 Braidwood Unit 2 Reactor Vessel Material Pro perties Chemistry Initial Material Cu (%)"'

Ni (%)"'

Factor"'

RT er ('F)*'

Description Closure Head Not 0.75 20"'

Flange Reported Vessel Flange 0.07 0.70 20"'

Lower Shell 0.06 0.77 37.0

-30 Forging Upper Shell 0.03 0.71 20.0

-30 Forging Weld Metal 0.03 0.67 41.0 40 WF562 (a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Position 1 (Reference 11).

(b) Initial RTervalues are measured, WCAP-14970 (Ref:rence 7).

(c) Closure head and vessel Dange Initial RTervalues are used for considering Dange requirements for the heatup/cooldown curves, WCAP-14970.

13

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Braidwood Unit 2 Adjusted Reference Temperatures (ARTS) at the 1/4T and 3/4T Locations for 12 EFPY 12 EFPY Material Description 1/4T ART (*F) 3/4T ART ('F)

Lower Shell Forging 29.5 10.1 Using credible surveillance capsule datt.

(RG Position 2'")

-8.6

-15.6 Upper Shell Forging 2.2

-8.3 (RG Position l'")

Weld Metal 105 o 84.4 (RG Position l'")

Using credible 66.9*

58.l*

surveillance capsule data (RG Position 2'")

(a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Position 1 and Position 2 (Reference 11).

(b) These ART values were used to generate the Braidwood Unit 2 heatup and cooldown curves, WCAP-14970 (Reference 7).

14

BRAIDWOOD - UNIT 2 t

PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 12 EFPY at the Limiting Reactor Vessel Material Weld Metal WF562 (Based on Surveillance Capsule Data)

Parameter Values Operating Time 12 EFPY Location")

1/4T ART 3/4T ART f.

Chemistry Factor, CF ('F) 16.7 16.7 2

Fluence (f), n/cm 4.95x10" 1.78x10" (E>1.0 Mev)i')

Fluence Factor, FF 0.804

_ 0.542 ARTmn= CFxFF('F) 13.43 9.05 Initial RTm. I('F) 40 40 Margia, M(*F) 13.43 9.05 ART = 1+(CF'FF)+M, 'F 66.9 58.1 per RG 1.99. Revision 2 (a) Fluence, f, is based upon f r(E>l.0 Mev) = 1.100x10'* at 16 EFPY, WCAP 14228 (Reference 3).

(b) The Braidwood Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.

15

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5 RTru Values for 13raidwood Unit 2 for 32 EFPY CF p.>.

pgi

.M RTsum ARTeis-RTers Material-('F)

(*F)

('F)

('F)

('F)-

Upper Shell Forging 20.0 2.199 1.214 24.28

-30 24.28 18.6 MK 24 3 Lower Shell Forging 37.0 2 199 1.214 34,0

-30 44.92 48.9 Using Surveillance Capsule Data 13.3 2.199 1214 34.0")

30 16.15 20.2 Circumferential Weld 41.0 2.199 1.214 49.77 40 49.77 139.5 Metal WF562 Using Surveillance 16.7 2.199 1.214 20.27 40 20.27 80.5 Capsule Data")

(a) 2.199x10" n/cm' (E>l.0 Mev) for 32 EFPY from Braidwood 2 PTS report WCAP-14229 (Reference 8)

(b) FF (Fluence Factor) = fa2:en n (c) Calculated using CF based on surveillance capsule data per RG 1.99, Position 2 (Refaence 1l)

(d) Double margin is used here due to the base metal surveillance capsule data exceeding the one s:gma criteria from the credibility evaluation. See WCAP-14824, Rev. 2 (Reference 2).

16

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4 6 RTns Values for Braidwood Unit 2 for 48 EFPY CF p.>

pp M

RTsortv)

ARTns RTns Matenal

("F)

(*F)

(*F)

('F)

('F)

Upper Shell Forging 20.0 3.298 1.313 26.26

-30 26.26 22.5 Lower Shell Forging 37.0 3.298 1.313 34.0

-30 48.58 52.6 N

Using Surveillance 13.3 3.298 1.313 34.0 )

-30 17.46 21.5 Capsule Data (*)

Circumferential Weld 41.0 3.298 1.313 53.83 40 53.83-147.7 Metal WF562 Using Surveillance 16.7 3.298 1.313 21.93 40 21.93 83.9 Capsule Data'*)

(a) 3.298x10" rt cm (E>l.0 Mev) for 48 EFPY from Braidwood 2 PTS report WCAP-14229

/

2 (Reference 8).

(b) FF (Fluence Factor) = f*2'#'*"*'O (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 11).

(d) Double margin is used here due to the base metal surveillance capsule data exceeding the one sigma criteria from the credibility evaluation. See WCAP-14824 Rev. 2 (Reference 2).

l l

l 17

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References 1.

Andracheck, J.D., et al., WCAP-14040-A. " Methodology used to Develop Cold Overpressure / Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2 January 1996.

2.

' Grendys, P.A., WCAP-14824, " Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood,"

Revision 2, November 1997.

i 3.

Peter, P.A., et al., WCAP-14228, " Analysis of Capsule X from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program,"

March 1995.

i 4.

Terek, E., et al., WCAP-12845, " Analysis of Capsule U from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program,"

March 1991.

5.

Westinghouse Letter to Commonwealth Edison Company, CCE-96-104, "Braidwood Unit 2 LTOPS Setpoints Based on 16 EFPY P/T Limits," January 24,1996.

6.

Singer, L.R., WCAP-11188, " Commonwealth Edison Company, Braidwood Station Unit 2 Reactor Vessel Surveillance Program," December 1986.

7.

Laubham, T. J., WCAP-14970, "Braidwood Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," October 1997 and Erratta Sheet (Westinghouse Letter CAE-97 210, CCE-97 289).

8.

Peter, P. A., WCAP 14229, " Evaluation of Pressurized Thermal Shock for Braidwood Unit 2," March 1995.

9.

Lippencott, E.P. WCAP-14044, " Westinghouse Surveillance Capsule Neutron Fluence Reevaluation," April 1994.

10.

10 CFR 50.61, " Fracture Tou'ghness Requirements for Protection Against Pressurized Thermal Shock Events,"(PTS Rule) January 18,1996.

11.

U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, "Radation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.

12.

Peter, P.A., WCAP-14243, "Braidwood Unit i Heatup and Cooldown Limit Curves for Normal Operation," March 1995.

ERAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT References (Continued) 13.

Comed Calculation BRW 96-9061/BYR 96-293," Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 2 Origimd Steam Generators)," Revision O.

14.

Westinghouse Letter to Commonwealth Edison Company, CCE-97-278, "CemEd Braidwood Unit 2 COMMS Setpoint for 12 EFPY," October 23,1997.

15.

Westinghouse Letter to Commonwealth Edison Company, CAE-97-211, CCE-97 290,

" Comed Byron and Braidwood Units 1 and 2 ATmetal Evaluation for Commonwealth Edison," November 7,1997, 1

19

e, ATTACHMENT I WCAP-14824, REVISION 2, BYRON UNIT 1 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPERATION AND SURVEILLANCE WELD METAL INTEGRATION FOR BYRON AND BRAIDWOOD 20-

,