ML20199E875

Revised Byron Unit 1 Pressure Temperature Limits Rept (Ptlr)

ML20199E875
Person / Time
Site:	Byron
Issue date:	11/03/1997
From:	COMMONWEALTH EDISON CO.
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ML20199E810	List: ML20199E806 ML20199E828 ML20199E853 ML20199E875 ML20199E880
References
NUDOCS 9711240018
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Text

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5 BYRON UNIT 1 PRESSURE TEMPERATURE LIMITS REPORT (PTLR) -

- (Revised November 3,1997) t 4

9711240018 971118

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PDR ADOCK 05000454' P

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BYRON - UNIT I 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

3.0 Reactor Vessel Material Surveillance Program 9

4.0 Supplemental Data Tables 11 5.0 References 18 Attachment WCAP-14824, Revision 2, " Byron Unit I Heatup and Cooldown Limit Curves for Normal

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

November 1997.

ii

- BYRON e UNIT 1

- PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures.

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

4 100 'F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors) 2.2 '

Byron Unit 1 Reactor Coolant System Cooldown Limitations 'Cooldown Rates

5 up to 100 *F/hr) Applicable for the First 12 EFPY (Without Margins for Instrumentation Errors)

[2.3 Byron Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low 7

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

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BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT

- c.

List of Tables -

Table Page 2.1.

Byron Unit _2 Heatup and Cooldown Data Points at 12 EFPY

'6 (Without Margins for Instrumentation Errors) 2.2 Da'a Points from Byron Unit 1 PORV Setpoints for the LTO'P System 8-3.1 Byron Unit 1 Capsule Withdrawal Schedule 10 4.1 Byren Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data 12-4.2 Byron Unit 1 Reactor Vessel Meterials Properties 13 4.3 Summary of Byron Unit 1 Adjusted Reference Temperatures (ARTS) at the 1/4T 14 anio 3/4T Locations for 12 EFPY 4.4 Byron Unit i Calculation of Adjusted Reference Temperatures (ARTS) at

-15 12 EFPY at tne Limiting B-actor Vessel MaterialIntermediate Shell Forging SP 5933 (Based on Surveillance Capsule Data) 4.5 RTns Values for Byron Unit I for 32 EFPY 16 4.6 R fns Values for Byron Unit I for 48 EFPY 17 bl c

iY -

1

BYRON - UhTr 1 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 [ ].

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 L ied below:

LCO 3.4.9.1 Pressureffemperatuie Limits; and LCO 3.4.9.3 Overpressure Protection Systems.

[LCO 3.4.3 RCS Pressure and Temperatur (Pff) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System),

2.0 Operating Limits The PTLR limits for Byron Unit I were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A (Reference 1) was used with four exceptions:

a) Use of ENDF/B IV neutron transport cross-section library and ENDF/B V dosimetry reaction Cross sections, b) Optional use of ASME Code Section XI, Appendix G, Article G 2000,1996 Addenda (not used for Byron Unit i P-T curves contained in this attachment),

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 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.

2.1 RCS Pressure and Temperature (Pff) Limits (LCO 3.4.9.1/ [3.4.3])

2.1.1 The RCS temperature rate-of-change limits defined in Reference 2 cre:

a. A maximum heatup of 100'F in any 1-hour period.

b A maximum cooldown of 100*F in any 1-hour period I

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4

- 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.--

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 P/T limits for cooldown are shown in i

. Figure 2.2 and Table 2.1. These limits are defined in WCAP-14824, Rev. 2 (Reference 2). 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. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10CFR50, 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 c(pal 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/

[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 li. nits which were established in accordance with 10CFR50, Appendix G without allowance for instrumentation 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 (Non Technical Specification)

The as analyzed LTOP enable temperature is 200*F (Reference 15).

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

- (Byron Unit 1 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperatu e 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.

2

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. BYRON - UNIT I i

PRESSURE AND TEMPERATURE LIMITS REPORT 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 t'o the =

~,

studs, 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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BYRON - UNIT 1 d

PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS

. LIMITING MATERIAL:- INTERMEDIATE SHELL FORGING SP-5933 (using suu. capsule ca:a).-

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BYRON o UNIT 1 4

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit i Heatup and Cooldown* Data Points at 12 EFPY" (Without margins for instrumenation errors) -

- HEATUP CURVES COOLDOWN CURVES

,m 100 F Heatup Test Curve Criticality Limit Limit 61eedy 5:ste 25 DEG r 50 TJG F 100 DEG F T l P

T l P T l P T l P T l P T i P T l P 60 l

0 203 0

162 2000 60 0

60 0

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65 621 65 610 65 570 65 489 70 621 203 0

70 621 70 621 70 567 70 609 75 621 203 671 75 621 75 621 75 605 75 531 60 621 203

-657 90 621 80 621 80 621 80 554 85 621 203 646 85 621 85 621 85 621 65 579 9?

621 203 639 90 621 90 621 90 621 90 607 95 621 203 6 34 95 621 95 621 95 621 95 621 100 l 621 203 632 100 621 100 621 100 621 100 l 621 105 l 621 203 i 633 105 621 105 l 621 105 621 105 1 621 110 l 621 203 l 637 110 621 110 l 621 110 621 110l621 115-621 203 l 642 115 621 115 621 115 621 110 l 621 120 621 203 l 651 120 621 120 621 120 621 120 1 621 125 621-203 l 661 125 621 125 621 125 621 125 l 621 130 621 203 674 130 621 130 621 130 621 1a0 1 621 135 621 203 689 135 1 621 135 621 135 621 135 } 621 140 l 621 203 1 707 140 l 621 140 1 621 140 621 140 l 621 i

145 l 621 203 ! 727 145 l 621 145 1 621 145' 621 l

150 621 203 ! 749 tb0 l 621 150 l 621 1

155 1 621 203 1 774 l

155 l 621 l

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175 l 621 l

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180 l 621 I

l 180 900 230 l 980 l

180 1483 l

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190 l 983 240 l 1075 190 1540 l

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• Heatup and Cooldown data includes vessel flange requirements ef 180 F and 621 psig per 10CFR50, Appendix G.

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

BYRON - UNIT 1 i

. PRESSURE AND TEMPERATURE LIMITS REPORT l

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l

BYRON - UNIT 1 PRESSURE AND TEMPERA 1URE LIMITS REPORT Table 2.2 Data Points for Byron Unit 1 Maximu n allowable Setpoira for the LTOP System Applicable for the First 12 EFPY PCV-455A PCV-456 pTY 0413M)

(1TY-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 70 514 100 497 100 514 120 446 120 462 150 446.

150 462 200 446 200 462 250 587 250 604 300 587 300 604 350 587 350 604 450 2350 450 2350 Note: To determine maximum allowable lift seipoints for RCS Pressure. and RCS Temperatures greater than 350'F, linearly interpolate between the 350*F and 450*F data points shown above.

8

BYRON - UNIT I 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 Figure 2.1, Figure 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 approavaing the withdrawal schedule.

The pressure vessel material ser /eillance program (Ref. 6) is in compliance with Appendix H to 10 CFR 50, " Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTu, which is determined in accordance with ASME,Section III, NB 2331. The empirical relationship between RT, 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.

9-l c

BYRON - UNIT 1 l

PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Byron Unit 1 Capsule Withdrawal Schedule Capsule -

Vessel Location.

Capsule Lead Removal Time

• Estimated Capsule (Degrees)

- Factor (EFPY)

Fluence (n/cm )

2 U

58.5' 3.85 1.15 (Removed' ) -

3.72 x 10" X

238.5*

3.79 5.64 (Removed) 1.39 x 10" -

W 121.5*

3.79 8.44 (EOL Wall) 2,159 x 10"N 12.66(1.5 EOL Z

301.5*

3.79 WallM) 3.238 x 10"

.V 61.0*

3.59 Standby Y

241.0*

3.59 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-1 of WCAP-14824, Rev. (Reference 2, which is Attachment I to this report).

10

BYRON-UNIT 1 i -

PRESSURE AND TEMPERATURE LIMITS REPORT

- 4.0 Supplemental Data Tables

- The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.

Table 4.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 4.2 provides the reactor vessei material properties table.

Table 4.3 piovides a summary of the Byron Unit 1 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 I reactor vessel material (Intermediate Shell Forging SP-5933).

Table 4.5 provides RTm values for Byron Unit I for 32 EFPY.-

Table 4.6 provides RTmvalues for Byron Unit I for 48 EFPY.

4 11

l BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 h

Calculation of Chemistry Factors Using Surveillance Capsule Data t

Fluence Material Capsule (n/cm,

FF(4 Measured FF* ART, (FF)'

2 E>l.0 ARTm Mev), f Inter. Shell Forging SP-5933 U

3.72x10" 0.727 0

0 0.529 (Tangential)

X l.39x10" 1.091 30 32.73 1.190 Inter. Shell Forging SP-5933 U

3.72x10" 0.727 0

0 0.529 (Axial)

X 1.39x10" 1.091 30 32.73 1.190 u

Sum:

65.46 3.44 Chemistry Factor * = 65.46 + 3.44 = 19.0'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(4 114.56 1.190 Byron 2 Weld Metal WF-447M U

3.996x10" 0.746 0

0 0

0.557 W

l.211x10" 1.053 30 90M 94.77 1.110 Sum:

209.33 3.386 Chemistry Factor * = 209.33+3.386= 61.8*F (a) FF = Fluence Factor = f"'N o (b) Byron Unit 1 ARTm values s,ere obtained from the surveillance Capsule X analysis, WCAP 13880 (Reference 3).

The Byron Unit I capsule fluence values were recalculated using the ENDF/B V scattering cross sections in 1994 and are documented in WCAP 14N4 (Reference 8).

(c) Byron Unit 2 capsule fluence, FF, and ARTmvalues were obtained from the surveilla-ce Capsule W analysis (WCAP 14064) using the ENDF/B V scattering cross sections.

- (d) Chemistry Factor = I(FF'ARTm) /I((FF)')

(e) 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 ration is 2.5 (68.0.27.0 = 2.5); however, for conservatism, ratio of 3.0 was used.

12

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2 Byron Unit 1 Reactor Vessel Material Properties Chemistry Initial Material Cu(%)W Ni(%)N FactorN RT a ('F)N Description Closure Head 0.74 60N Flange Vessel Flange 0.73 10M Intermediate Shell 0.0364 0.747 23.8 40 Forging SP-5933 Lower Shell 0.04 0.64 26.0 10 Forging 5P 5951 Circumferential 0.05 0.62 68.0

-30 Weld (WF-336) a) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position I (Ref.12) b) Initial RTswvalues are measured, WCAP 13880 (Ref. 3) c) Closure head and vessel flange initial RTswvalues are used for considering Dange requirements for the heatup/cooldown curves in WCAP-14814, Rev. 2 (Ref. 2).

l 13

i BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Byron Unit 1 Adjusted Reference Temperatures (ARTS) at 1/4T and 3/4T Locations for 12 EFPY 12 EFPY Materir! Description 1/4T ART (*F) 3'/4T ART (*F)

Intermediate Shell 78 66 Forging SP 5933 (RG Position 1(')

Using credible 70*)

60*)

surveillance capsule data (RG Position 2(*))

Lower Shell Forging 52 38 SP-5951 (RG Position 1(*D Circumferential Weld 79 43 WF-336 (RG Position 1(*))

Using credible 47 31 surveillance capsule data (RG Position 2(*D a) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Revision 2, Position I and 2 (Ref.12).

b) These ART values were used to e,enerate the Byron Unit I heatup and cooldown curves in WCAP 14824, Rev. 2 (Ref. 2).

14 s

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS RE70RT Table 4.4 Byron Unit 1 Calculation of Adjusted Reference Temperatures (ARTS) at 12 EFPY at the Limiting Reactor Vessel Material Intermediate Shell Forging SP-5933 (Based on Surveillance Capsule Data) b Parameter Values Operating Time 12 EFPY LocationN 1/4T ART 3/4T ART Chemistry Factor, CF (*F) 19.0 19.0 s

2 Fluence (f), n/cm 4.86x10" 1.75x10" (E>1.0 Mev))(')

Fluence Factor, FF 0.799 0.538 ARTer= CFxFF('F) 15.2 10.2 Initial RTor,1(*F) 40 40 Margin, M(*F) 15.2 10.2 ART = 1+(CF*FF)+M, 'F 70 60 per RG 1.99, Revision 2 a) Fluence, f, is based upon fu (Erl.0 Mev) = 8.10x10 at 12 EFPY (WCAP-14824, Rev. 2).

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

f 15 Y

t r

BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT l

Table 4.5 t

RTns Values for Byron Unit 1 for 32 EFPY CF p

ppm M

RT,erm ARTns RTns Material

('F)

('F)

(*F)

('F)

('F)

Intermediate Shell Forging 23.8 2.159 1.209 28.3 40 28.8 97.6 SP 5933 Using Surveillance 19.1 2.159 1.209 17.0

'40 23.1 80.1 r

Capsule DataM Lower Shell Forging 26.0 2.159 1.209 31.4 10 31.4 72.8

$P 5951

~

Weld Metal, WF 336 68,0 2.159 1.209 56.0 30 82.2 108.2 Using Ssveillancu 61.8 2.159 1.209 28.0-30 74.7 72.7 Capsu'e DataN (a) 2.159 x 10 ' n/cm'(E>l.0 Mev) for 32 EFPY from By in 1 PTS report, WCAP 13881 (Ref. 9).

(b) FF (Fluence Factor) = f"'N'-

g) Calculated using a CF based on surveiDance capsule data per RO l.99, Revision 2, Position 2 (Ref.12).

i 16

,_. - - _ _,..,.. - _. _ _., _ +-

l BYRON UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4

Table 4.6 y

RTns Values for Byron Unit I for 48 EFPY i

CF p

ppm M

RTmnm ARTns RTns Material

('F)

(*F)

('F)

('F)

('F) i Intermediate Shell Forging 23.8 3.238 1.309 31.2 40 31.2 102.4 SP 5933 Using Surveillance 19.1 3.238 1.309 17.0

'40 25.0 82.0 i

Capsule Data'"

Lower Shell Forging 26.0 3.238 1.309 34.0 10 34.0 78.0 5P 59$1 Weld Metal, WF 336 68.0 3.238 1.309 56.0 30 89.0 115.0 Using Surveillance 61.8 3.238 1.309' 28.0

-30 80.9 78.9 Capsule Data (4 i

a) 2.159x10 n/cm'(E>l.0 Mev) for 32 EFPY from Byron 1 PTS report WCAP 13881 (Ref. 9). The following calcuistion provides the 48 EFPY fluence value:

2.159x10" + ((2.159x10" 3.807x10"y37 5.64 EFPY)) * (48 32 EFPY) = 3.238x10" n/cm' b) FF (Fluence Factor) = f""'*'0 c) Calculated using a CF based on surveillance capsule data per RG 1.99, Revision 2, Position 2 (Ref.12).

i

-17 i

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v~w--.*

BYRON. UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References

1. WCAP 14040 NP A, Revision 2," Methodology Used to Develop Cold Overpressure Mitigating Sy6 tem Setpoints and RCS 11eatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.

2. WCAP 14824 Revision 2," Byron Unit 111eatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld MetalIntegration for Byron A Braidwood", November 1997.

3. WCAP 13880," Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", P.A. Peter, et. al., January 1994.

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

5. Westinghouse Letter to Commomvealth Edison Company, CAE-96106," Byron Unit I and 2 LTOPS Setpoints Based on 10 and 12 EFPY P/T Limits", January 17,1996.

6. WCAP 9517," Commonwealth Edisor, Company, Byron Station Unit 1 Reactor Vessel Surveillance Program", J.A. Davidson, July 1979.

7. Westinghouse Letter Report to Commonwealth Edison Company, FDRT/SPRO 009(94)," Byron Unit 1 11eatup and Cooldown Lin,it Curves for Normal Opera. ion", P.A. Peter, January 1994.

8. WCAP 14044," Westinghouse Surveillance Capsule Neutron Fluence Reevaluation", E.P.

Lippencott, April 1994.

9. WCAP-13881," Evaluation of Pressurized Thermal Shock for Byron Unit 1", P.A. Peter, January 1994.

10.10 CFR Part 50, Appendix G. " 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 Thermal Shock Events", May 15,1991. (PTS Rule)

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

Nuclear Regulatory Commission, May 1988.

13. Comed Calculation BRW-96-9061/DYR 96 293, Rev. O " Channel Accuracy for Power Operated Relief Valve (PORV) Fetpoints and Wide Range RCS Temperature Indication (Unit 1 Original Steam Generators and Replacement Steam Generators)".

18

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References (cont.)

14. Comed Nuclear Fuel Services Department NDIT No. 960186, Revision 1 " Maximum Allowable LTOPS PORY Setpoints for Byron Unit I with RSGs".

15. Westinghouse Letter to Comed, CAE 97 211/CCE 97 290," Byron and Braidwood Units 1 and 2 ATmetal Evaluation," November 7,1997.

1 l

h C

e I

4 19

.2...

_. _.. _.., _. _.,. _.. _.. -,._ _ _. _ _. _,, J

1 i

ATTACHMENT WCAP 14824, REVISION 2, BYRON UNIT 1 HEATUP AND COOLDOWN LIMIT CURVES FOR NORMAL OPIRATION AND SURVEILLANCE WELD METAL INTEGRATIC N FOR BYRON AND BRAIDWOOD 1

. - -.