BYRON 2003-0064, Revision to Reactor Coolant System Pressure & Temperature Limits Report

From kanterella
Jump to navigation Jump to search
Revision to Reactor Coolant System Pressure & Temperature Limits Report
ML032050516
Person / Time
Site: Byron  Constellation icon.png
Issue date: 07/17/2003
From: Kuczynski S
Exelon Nuclear
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
2.01.0700, BYRON 2003-0064
Download: ML032050516 (49)
v • d • e


Text

oil 1W Exeln.M Exelon Generation Byron Generating Station www.exeloncorp.com Nuclear 4450 North German ChuTch Road Byron, IL 61010-9794 Tel 815-234-5441 July 17, 2003 LTR: BYRON 2003-0064 File: 2.01.0700 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 Byron Station, Unit 1 and Unit 2 Facility Operating License Nos.NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Revision to the Reactor Coolant System Pressure and Temperature Limits Report, Byron Station Unit 1 and Unit 2 In accordance with Technical Specification 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)," section c. we are providing a revision to the Unit I and Unit 2 PTLRs. The current PTLRs were revised to reflect a slight reduction in the Effective Full Power Years (EFPY) applicability for figure 2.1, "Reactor Coolant System Heatup Limitations,"

figure 2.2, "Reactor Coolant System Cooldown Limitations," and figure 2.3, "Maximum Allowable Nominal PORV Setpoint for Low Temperature Overpressure Protection System' as a result of our Power Uprate analyses.

Unit 1 EFPY was reduced from the current 16 to 15.6 EFPY and Unit 2 was reduced from the current 16 to 15.5 EFPY. Unit 1 is estimated to reach 15.6 EFPY in November 2004, and Unit 2 is estimated to reach 15.5 EFPY in July 2005.

In addition, several minor editorial and administrative changes were made and are listed in the attachments. Should you have any questions concerning these reports, please contact William Grundmann, Regulatory Assurance Manager, at (815) 406 2800.

Respectfully, 5$# 72iXo Stephen E. Kuczynski Site Vice President Byron Nuclear Generating Station Attachments: 1) Byron Station Unit 1 PTLR, revision 1

2) Byron Station Unit 2 PTLR, revision 1 AoDI

cc: Regional Administrator - NRC Region Il NRC Senior Resident Inspector - Byron Station NRC Project Manager - NRR - Byron Station Office of Nuclear Facility Safety - Illinois Department of Nuclear Safety

bcc: Manager of Energy Practice - Winston & Strawn Site Vice President - Byron Station Vice President - Licensing & Regulatory Affairs Director - Ucensing Manager - Licensing and Compliance - Braidwood & Byron Stations Regulatory Assurance Manager - Byron Station Exelon Document Control Desk Licensing (Hard Copy)

Exelon Document Control Desk Licensing (Electronic Copy)

PWR Supervisor - Nuclear Fuels Management Supervisor - Byron Reactor Engineering

z-ATTACHMENT 1 Byron Station Unitl Reactor Coolant System Pressure and Temperature Limits Report Revision 1

Summary of Unit 1 Changes

  • Cover sheet was revised to reflect new revision date
  • Table of Contents was revised to reflect changes in page number and titles for figures and tables
  • Administrative/editorial changes were made to paragraph 1.0
  • Reference 23. was added to paragraph 2.0
  • A reference to Table 2.3 was revised to Table 2.2, in paragraph 2.2
  • The applicability for Figures 2.1, 2.2, and 2.3, and Tables 2.1 and 2.2 was revised from 16 to 15.6 EFPY. The applicable period is being revised due to power uprate impacts and the increased neutron flux.
  • In Table 3.1, the removal times for capsules Z, V, and Y were revised to properly reflect their status as standby capsules and footnotes were revised or deleted, as necessary. Other administrative changes were made to Table 3.1
  • In paragraph 4.0, revisions to table titles are reflected, as well as the elimination of Table 4.5 from the previous revision
  • Table 4.1: reference information was added to the Materials column, minor corrections were made to data in the FF, FF2, FF*delta RTNDT columns, and footnotes were revised or deleted, as necessary.
  • Table 4.2: footnotes were revised or deleted, as necessary.
  • Table 4.3: minor revisions were made to ART values for Circumferential Weld WF-336, the applicable EFPY was changed and footnotes were revised or deleted, as necessary.
  • Table 4.4: The title was changed to reflect the revision to 15.6 EFPY, and footnotes were revised or deleted, as necessary. The applicable period is being revised due to power uprate impacts and the increased neutron flux.
  • The original Table 4.5 was no longer applicable to Byron Unit 1 and was deleted
  • Table 4.5 (as titled in 2003 revision): The title was changed from Table 4.6, and footnotes were revised or deleted, as necessary.
  • Reference 23 was added to the References paragraph, 5.0

BYRON UNIT 1 PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

(April 2003)

BYRON - UNIT 1 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 TS-LCO 3.4.3 1 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints TS-LCO 3.4.12. 2 2.3 LTOP Enable Temperature 2 2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification) 3 2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification) 3 3.0 Reactor Vessel Material Surveillance Program 10 4.0 Supplemental Data Tables 11 5.0 References 17 List of Figures Figure Page 2.1 Byron Unit I Reactor Coolant System Heatup Limitations 4 (Heatup Rates up to 100 0F/hr) Applicable for the First 15.6 EFPY (Without Margins for Instrumentation Errors and Using the 1996Appendix G Methodology) 2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations 5 (Cooldown Rates up to 1000F/hr) Applicable for the First 15.6 EFPY (Without Margins for Instrumentation Errors and Using the 1996Appendix G Methodology) 2.3 Byron Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for the First 15.6 EFPY ii

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 Byron Unit 1 Heatup and Cooldown Data Points at 15.6 EFPY 6 (Without Margins for Instrumentation Errors and Using the 996Appendix G Methodology) 2.2 Data Points from Byron Unit 1 Maximum Allowable Setpoints for the LTOP System Applicable for the First 15.6 EFPY 9 3.1 Byron Unit 1 Capsule Withdrawal Schedule 10 4.1 Calculation of Chemistry Factors Using Surveillance Capsule Data 12 4.2 Reactor Vessel Beltline Material Unirradiated Toughness Properties 13 4.3 Summary of Adjusted Reference Temperature (ART) at the 1/4T 14 and 3/4T Location for 15.6 EFPY 4.4 Byron Unit 1 Calculation of Adjusted Reference Temperature (ART) at 15 15.6 EFPY at the Limiting Reactor Vessel Material, Intermediate Shell Forging 5P-5933 (Conservatively Based on Surveillance Capsule Data) 4.5 RTpn for Byron Unit 1 Beltline Region Materials at Life Extension (48 EFPY) 16 iii

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

This PTLR for Unit 1 has been prepared in accordance with the requirements of TS 5.6.6 (RCS Pressure and Temperature Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

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

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

a) Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, and b) Use of RELAP computer code for calculation of LTOP setpoints for Byron Unit 1 replacement steam generators.

These exceptions to the methodology in WCAP-14040-NP-A have been reviewed and accepted by the NRC in Reference 16.

WCAP-15 124, Reference 17, provides the basis for the Byron Unit 1 P/T curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for Byron and Braidwood Units 1 and 2 is documented in Reference 2. Reference 23 evaluated the effect of higher fluence from 5%

uprate on the existing PT curves.

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

a) A maximum heatup of 100F in any l-hour period.

b) A maximum cooldown of 100 0F in any 1-hour period, and c) A maximum temperature change of less than or equal to 100F in any 1-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

I

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2.0 Operating Limits (continued) 2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defined in WCAP-15124, Rev. 0 (Reference 17).

Consistent with the methodology described in Reference 1, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints TS-LCO 3A.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 Residual Heat Removal (RH) Suction Relief Valves are also analyzed to individually provide low temperature overpressure protection. This analysis for the RH Suction Relief Valves remains valid with the current Appendix G limits contained in this PTLR document and will be reevaluated in the future as the Appendix G limits are revised.

The LTOP setpoints are based on P/T limits that were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error. The LTOP setpoints were developed using the methodology described in Reference 1. The LTOP PORV maximum lift settings shown in Figure 2.3 and Table 2.2 account for appropriate instrument error.

2.3 LTOP Enable Temperature The as analyzed LTOP enable temperature is 200'F (Reference 15 andl7).

The required enable temperature for the PORVs shall be < 350'F RCS temperature.

(Byron Unit procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350'F and below and disarming of LTOP for RCS temperature above 350'F).

Note that the last LTOP PORV segment in Table 2.2 extends to 450°F where the pressure setpoint is 2350 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

2

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2A Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be 2 60'F. Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere (Reference 17).

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

Based on the steady-state limits specified in Table 2.1, the minimum temperature at which the Reactor Vessel may be pressurized (i.e., in an unvented condition) shall be Ž 650 F, plus an allowance for the uncertainty of the temperature instrument, determined using a technique consistent with ISA-S67.04-1994.

3

& Z.

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT zo-- - - - - ILATS OFF' 20tW I 1000 2

1750 2250 __

__t _

T3 IN l _} __ _CRM_+A_ LIM BASED ON l l--R\AC 0 f1111:i+z e.WIIII1 UPTO100 F/HR~~~~~~~~~FR 00FAf

-1 CN1 C~ f Y U B ASEON R A E 25D TEST TEW1ERAUFJE (225 F)

HY~~~~~~~~~~~~~-DROSTAIc

- EFPY I1FORT'IESERVICEPEODUP7 0 50 100 150 200 250 300 350 400 450 500 Moderator Temperahn (Deg. F)

Figure 2.1:

Byron Unit 1 Reactor Coolant System IHeatup Limitations (Heatup rates up to 100°F/hr) Applicable for the First 15.6 EFPY (Without margins for instrumentation errors and using 1996 Appendix G Methodology) 4

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2250-1750 - - - - - - _ I 1500- 150O

__ __-_ ___ _ l l ll J f __ __ _. OPERATION I la I

I 1250- - - - -

I UI 1000 --

IRATES F/Hr IIIiII Ir.IItI 750 - 1I11- Il_____IIIII1 500 - 50 100 0 ZOIJI i IMMII~

LI-1 X IDOLTUPI I lii M ff I I IaiIMiM I II ii lII 1111111111111111111111i I . f . . .

I I

I-I .

I'l E=

I I I I

~~~~----

U I I I I I I - I I II 1

I I I II I I 4

ff .

I L-f 0 50 100 1SO 200 250 300 350 400 450 500 Moderator Temperature (Deg F)

Figure 2.2:

Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates up to 100°Flhr)

Applicable'for the First 15.6 EFPY (Without margins for instrumentation errors and using 1996 Appendix G Methodology) 5

ii BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit 1 Heatup and Cooldomn Data Points at 15.6 EFPY (Without Margins for Instrumentation Errors and Using the 1996 Appendix G Methodology)

Heatup Curve Cooldown Curves 100 F Criticality Leak Test Steady 25 F 50 F 100 F Heatup Limit Limit State Cooldown Cooldown Cooldown T P T P T P T P T P T P T P 60 0225 0 204 2000 60 0 60 0 60 0 60 0 60 587 225 587 225 2485 60 613 60 561 60 509 60 402 65 587 225 587 65 621 65 572 65 520 65 414 70 587 225 587 70 621 70 582 70 531 70 427 75 587 225 587 _ 75 621 75 594 75 544 75 442 80 587 225 587 80 621 80 607 80 557 8 458 85 587 225 587 85 621 85 620 85 572 85 475 90 587 225 587 90 621 90 621 90 588 90 494 95 58 225 587 95 621 95 621 95 605 95 514 100 587 225 587 100 621 1001 621 100 621 100 535 105 587 225 587 105 621 105 621 105 621 105 559 110 587 225 587 110 621 110 621 110 621 110 584 115 587 225 587 115 621 115 621 115 621 115 611 120 588 225 587 120 621 120 621 120 621 120 621 125 591 225 587 125 621 125 621 125 621 125 621 130 596 225 587 130 621 130 621 130 621 130 621 135 602 225 587 135 621 135 621 135 621 135 621 140 611 225 587 140 621 140 621 140 621 140 621 145 621 225 588 145 621 145 621 145 621 145 621 150 621 225 591 150 621 150 621 150 621 150 621 155 621 225 596 155 621 155 621 155 621 155 621 160 621 225 602 160 621 160 621 160 621 160 621 165 621 225 611 165 621 165 621 165 621 165 621 170 621 225 622 170 621 170 621 170 621 170 621 175 621 225 634 1751 621 175 621 175 621 _

180 621 225 648 180 621 180 621 180 750 225 665 180 1207 180 1205 185 777 225 683 185 1261 190 806 225 703 190 1319 _ .

195 838 225 725 195 1382 .

200 872 225 750 200 1449 205 910 225 777 205 1521 .

210 950 230 806 - 210 1599 .

215 994 235 838 215 1h3 68 220 1041 240 872 220 1773 _ .

225 1092 245 910 _ 225 1869 . _

230 1147 250 950 230 1973 -

235 1206 255 994 =_=_235 2085 = = =

6

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1 (Continued)

Heatup Curve Cooldown Curves 100 F Criticality Leak Test Steady 25 F 50 F 100 F Heatup Lim It Limit State Cooldown Cooldown Cooldown T I P T P TIPr T P T P T -IP 240 1269 260 1041 - - 240 2205 - - - - -

245 1338 265 1092 - - 245 2334 1 - 1 -

250 1411 270 1147 - - 250 2473 -

255 1490 275 1206 __

260 1575 280 1269 - -

265 1666 285 1338 = =

270 1764 290 1411 = = = __

275 1869 295 1490 = =

280 1982 300 1575 = = -

285 2104 305 1666 = = =

290 2234 310 1764 = = - -r- -

295 2374 3151 1869 320 1982 325 2104 330 2234 335 2374

= = -=-  :--= - -

= - - - - - - ==- -- = - -

Note 1: Heatup and Cooldown data includes the vessel flange requirements of 180 F and 621 psig per IOCFR50, Appendix G.

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

Note 3: Temperatures and pressures are given in ° F and psig, respectively.

7

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 650 600 550 R 500 450 400 0 50 100 150 20o 250 300 350 400 AUCTIONEERED LOW RCS TEMPERATURE (DEG. F)

Figure 2.3 Byron Unit 1 Maximum Allowable Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 15.6 EFPY 8

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Byron Unit 1 Maximum Allowable Setpointsfor the LTOP System Applicable for the First 15.6 EFPY PCV-455A PCV456 (ITY-0413M) (1TY-0413P)

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

RCS TEMP. (DEG. F) (PSIG) 65 497 65 514 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 setpoints for RCS Pressure and RCS Temperatures greater than 3501F, linearly interpolate between the 3501F and 4501F data points shown above. (Setpoints extend to 4500 F to prevent PORV liftoff from an inadvertent LTOP system arming while at power.)

9

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Ref. 6) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME, Section Im,NB-233 1. The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

The third and final reactor vessel material irradiation surveillance specimens have been removed and analyzed to determine changes in the reactor vessel material properties. The surveillance capsule testing has been completed for the original operating period. Other capsules will be removed to avoid excessive fluence accumulation should they be needed to support life extension.

The removal schedule is provided in Table 3.1. The time of specimen withdrawal may be modified to coincide with those refueling outages or reactor shutdowns most closely approaching the withdrawal schedule.

Table 3.1 Byron Unit 1 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time(a) Estimated Capsule (Degrees) Factor (EFPY) Fluence (n/cm 2 )

U 58.50 4.22 1.15 (Removed) 4.04 x 10'"

X 238.50 4.27 5.64 (Removed) 1.57 x 10'9 W 121.50 4.20 9.24 (Removed) 2.43 x 0' 9 (b)

Z 301.50 4.20 Standby(c) 3.27x 0' 9 (c)

V 61.00 3.97 Standby (c)

Y 241.00 3.97 Standby (c) a) Effective Full Power Years (EFPY) from plant startup.

b) Maximum end of license (32 EFPY) inner vessel wall fluence is estimated to be 2.02 x l0'9 n/cm 2.

c) Standby capsule to be used for future license renewal (Derived from WCAP 15132, Rev. 1).

10

BYRON - 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.

Table 4.3 provides a summary of the Byron Unit 1 adjusted reference temperature (ARTs) at the 1/4T and 3/4T locations for 15.6 EFPY.

Table 4.4 shows the calculation of ARTs at 15.6 EFPY for the limiting Byron Unit 1 reactor vessel material (Intermediate Shell Forging 5P-5933).

Table 4.5 provides RTpTs values for Byron Unit 1 for 48 EFPY obtained from Reference 9.

I1I

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Calculation of Chemistry Factors Using Surveillance Capsule Data (a)

Fluence Material Capsule (n/cm 2, FF(a) Measured FF*ARTNDT (FF) 2 E>1.0 Mev), ARTNUT f

Inter. Shell U 4.04x1018 0.748 28.55 21.36 0.560 Forging 5P-5933 X 1.57x10' 9 1.124 9.82 11.04 1.263 (Tangential) W 2.43x10 9 1.239 49.20 60.96 1.535 Inter. Shell U 4.04x10' 8 0.748 18.52 13.85 0.560 Forging 5P-5933 X 1.57xlO' 9 1.124 53.03 59.61 1.263 (Axial) W 2.43xl0' 9 1.239 29.34 36.35 1.535 Sum: 203.17 6.716 Chemistry Factor = 203.17 --6.716= 30.3 0F Byron 1 Weld U 4.04xl0' 8 0.749 11.22 (5.61) 8.40 0.561 Metal WF-336 X 1.57x10' 9 1.125 80.22 40.11) 90.25 1.266 (Heat #442002) W 2.43xl0' 9 1.239 102.68 (51.34) 127.22 1.535 Byron 2 Weld U 4.05xl0' 8 0.749 16.88 (8.44) (D) 12.64 0.561 Metal WF-447 W 1.27 x10' 9 1.067 57.76 (28.88) (b) 61.63 1.138 (Heat #442002) X 2.30 x 1iO9 1.225 108.02 (54.01) lb) 132.32 1.500 Sum: 432.46 6.561 Chemistry Factor = 432.46.6.561= 65.90F a) Reference 17, Table 4-8 b) Adjusted ARTNm per Ratio Procedure of Regulatory Guide 1.99, Rev. 2 (Ref. 12). Ratio = 2.0. See Table 4.8 of WCAP 15178, Rev. 0. (Ref 22).

12

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2:

Reactor Vessel Beltline Material Unirradiated Toughness Properties (a)

Material Description Cu (%) Ni ()Initial RTNDT(

Closure Head Flange 124K358VA1 0.74 60 Vessel Flange 123J219VAI 0.73 10 Nozzle Shell Forging 123J218 0.05 0.72 30 Intermediate Shell Forging 5P-5933 0.04 0.74 40 Lower Shell Forging 5P-5951 0.04 0.64 10 Intermediate to Lower Shell Forging Circ. Weld 0.04 0.63 -30 Seam WF-336 (Heat # 442002)

Nozzle Shell to Intermediate Shell Forging Circ. 0.03 0.67 10 Weld Seam WF-501 (Heat # 442011)

Byron Unit I Surveillance Program 0.02 0.69 ---

Weld Metal (Heat # 442002)

Byron Unit 2 Surveillance Program 0.02 0.71 ---

Weld Metal (Heat # 442002)

Braidwood Units I & 2 Surveillance Program 0.03 0.67, 0.71 Weld Metals (Heat # 442011) a) Reference 17 13

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3 Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4T Location for 15.6 EFPY (a)

Material 15.6 EFPY 1/4T ART 314T ART Intermediate Shell Forging 5P-5933 84 70

- Using Surveillance Data(b) 100(c) 92(c)

Lower Shell Forging 5P-5951 54 40 Circumferential Weld WF-336 62 33

- Using Credible Surveillance Datad) 47 32 Circumferential Weld WF-501 54 37

- Using Credible Surveillance Data 28 21 form Braidwood 1 and 2 Nozzle Shell Forging 123J218 64 51 (a) Fluence, f is based upon ff(E>l.0 MeV) = 9.85x10' 8 at 15.6 EFPY (Ref. 23).

(b) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Revision 2, Position 2 along with a full margin since it was determined that this data was not credible and the Table chemistry factor was non conservative (Ref. 19).

(c) These ART values were used to generate the Byron Unit 1 15.6 EFPY heatup and cooldown curves (Ref 17).

(d) Calculated using the chemistry factor from the Byron Unit I and 2 integrated surveillance data as reported in WCAP-15178 (Reference 22) 14

A .

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4 Byron Unit 1 Calculation of Adjusted Reference Temperature (ART) at 15.6 EFPY at the Limiting Reactor Vessel Material, Intermediate Shell Forging 5P-5933 (Conservatively Based on Surveillance Capsule Data) (c)

Parameter Values Operating Time 15.6 EFPY Location(b) 1/4T ART 3/4T ART Chemistry Factor, CF (F) 30.3 30.3 Fluence(f), n/cm 2 5.9lxlO'l2.13x10 8 (E>1.0 Mev))(a)

Fluence Factor, FF 0.853 0.585 ARTNDry= CFxFF(-F) 25.8 17.7 Initial RT NDT, If) 40 40 Margin, M( 0F) _4 34 ART= I+(CF*FF)+M, F 100 92 per RG 1.99, Revision 2 a) Fluence, f, is based upon ff (E>1.0 Mev) 9.85xlO' 0 at 15.6 EFPY (Ref. 23).

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

c) WCAP 15123 15

II BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5:

RTpTS for Byron Unit 1 Beltline Region Materials at Life Extension (48 EFPY) (f) (g)

Material Fluence FF (h) CF (OF) ARTrs 5 c) Margin RTNDT(WJ() RTm(b)

(n/cm2, E>1.0 (of) (F) (OF) (OF)

MeV)

Intermediate Shell Forging 5P-5933 2.91 x 1019 28 26.0 33.3 33.3 40 107 Intermediate Shell Forging 5P-5933 2.91 x 1019 1.28 30.3 38.8 34 40 113 using S/C Data(d)

Lower shell Forging 5P-5951 2.91 x 109 1.28 26.0 33.3 33.3 10 77 Inter. To Lower Shell Circ. Weld 2.91 x IO'9 1.28 54.0 69.1 56 -30 95 Metal WF-336 (442002)

Inter. To Lower Shell Circ. Weld 2.91 x 109 1.28 65.9 84.4 28 -30 82 Metal (442002) using S/C Data(')

Nozzle Shell Forging 123J218 8.70 x 1018 0.961 31.0 29.8 29.8 30 90 Nozzle Shell to Inter. Shell Circ. 8.70 x loll 0.961 41.0 39.4 39.4 10 89 Weld Metal WF-501 (442011)

Nozzle Shell to Inter. Shell Circ.

Weld Metal (442011) using S/C 8.70x lO's 0.961 16.7 16.0 16.0 10 42 Data Notes:

a) Initial RTmT values are measured values (See Table 4.2) b) RTprs = RTNDT(u) + ARTpm + Margin (F) c) ARTpTs = CF

  • FF d) Surveillance data is considered not credible, however, since the chemistry factor (CF) from the Reg. Guide Tables (Pos.

1.1) is lower (i.e. CF via Pos. 2.1 > CF via Pos. 1.1), then the Pos. 2.1 CF is used to determine PTS with a fill a margin term, i.e. 17 'F.

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

f) The fluence for 48 EFPY (Ref. 9) did not incorporate the 5% increase. However, this fluence value is greater than the end-of-life fluence (32 EFPY).

g) Limiting RTpTs is significantly less than the PTS Screening Criteria of 270 'F.

h) FF (Fluence Factor) = o2 - o.0 lo bg 16

BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References

1. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D. Andrachek, et. al., January 1996.
2. 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 with Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE-97-23 1/CCE-97-314 and CAE-97-233/CCE-97-316, dated January 6, 1998.
3. WCAP-13880, "Analysis of Capsule X from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", P.A. Peter, et. al., January 1994.
4. WCAP-12685, "Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program", E. Terek, et. al., August 1990.
5. Westinghouse Letter to Commonwealth Edison Company, CAE-96-106, "Byron Unit 1 and 2 LTOPS Setpoints Based on 10 and 12 EFPY P/T Limits", January 17, 1996.
6. WCAP-9517, "Commonwealth Edison Company, Byron Station Unit 1 Reactor Vessel Surveillance Program", J.A. Davidson, July 1979.
7. Westinghouse Letter Report to Commonwealth Edison Company, FDRT/SPRO-009(94), "Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation", P.A. Peter, January 1994.
8. WCAP-14044, "Westinghouse Surveillance Capsule Neutron Fluence Reevaluation", E.P. Lippencott, April 1994.
9. WCAP-15125, "Evaluation of Pressurized Thermal Shock for Byron Unit 1", Revision 0, T. J.

Laubham et al., November 1998.

10. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements", Federal Register, Volume 60, No.

243, dated December 19, 1995.

11. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal 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. CoinEd Calculation BRW-96-906I/BYR 96-293, Rev. 0 "Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 1 Original Steam Generators and Replacement Steam Generators)".

17

%.9 BYRON - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References (continued)

14. ComEd Nuclear Fuel Services Department NDIT No. 960186, Revision I "Maximum Allowable LTOPS PORV Setpoints for Byron Unit 1 with RSGs".
15. Westinghouse Letter to ComEd, CAE-97-21 1/CCE-97-290, "Byron and Braidwood Units 1 and 2 AT Metal Evaluation," November 7, 1997.
16. NRC Letter from R. A. Capra, NRR, to 0. D. Kingsley, Commonwealth Edison Co., "Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802)," January 21, 1998.
17. WCAP- 15124, Revision 0, "Byron Unit 1 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, et al., November 1998.
18. WCAP-15 123, Revision 1, "Analysis of Capsule W from Commonwealth Edison Company Byron Unit 1 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et al., January 1999.
19. WCAP-15 183, Revision 0, "Commonwealth Edison Company Byron Unit 1 Surveillance Program Credibility Evaluation," T. J. Laubham, et al., June 1999.
20. WCAP-15176, Revision 0, "Analysis of Capsule X from Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et al., March 1999.
21. WCAP-15180, Revision 0, "Commonwealth Edison Company Byron Unit 2 Surveillance Program Credibility Evaluation," T. J. Laubham, et al., June 1999.
22. WCAP- 15178, Revision 0, "Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, et al., June 1999
23. Westinghouse Calculation CN-EMT-01-8, "Braidwood Unit 1 and 2, Development of New Pressure Temperature Limit Curves and Evaluation of Byron Units 1 and 2 PT Curve EFPY 18 Final

ATTACHMENT 2 Byron Station Unit 2 Reactor Coolant System Pressure and Temperature Limits Report Revision

Summary of Unit 2 Changes

  • Cover sheet was revised to reflect new revision date
  • Table of Contents was revised to reflect changes in page number and titles for figures and tables
  • Administrative/editorial changes were made to paragraph 1.0
  • Reference 20 was added to paragraph 2.0
  • The applicability for Figures 2.1, 2.2, and 2.3, and Tables 2.1 and 2.2 was revised from 16 to 15.5 EFPY. The applicable period is being revised due to power uprate impacts and the increased neutron flux.
  • In Table 3.1, the removal times for capsules Z, V, and Y were revised to properly reflect their status as standby capsules and footnotes were revised or deleted, as necessary. Other administrative changes were made to Table 3.1
  • In paragraph 4.0, revisions to the titles for Tables 4.3 and 4.4 are reflected
  • Table 4.1: An error in the table number was corrected, administrative changes were made, and footnotes were revised or deleted, as necessary.
  • Table 4.2: The title was revised to match Unit 1, the Chemistry Factor column was removed to match Unit 1, additional weld materials and their Cu and Ni data were added, minor revisions to Cu and Ni data was made, and footnotes were revised or deleted, as necessary.
  • Table 4.3: The title was changed to match Unit 1, the applicable EFPY was changed and footnotes were revised or deleted, as necessary.
  • Table 4.4: The title was changed to reflect the revision to 15.5 EFPY, and footnotes were revised or deleted, as necessary.
  • Table 4.5: The title was changed to match Unit 1, the title was changed to reflect the revision to 15.5 EFPY, and minor changes were made to data in the Fluence and FF columns.
  • Table 4.6: The title was changed to match Unit 1,the title was changed to reflect the revision to 15.5 EFPY, and administrative changes were made.
  • References 2, 8, and 19 were deleted, and Reference 20 was added to Paragraph 5.0.

BYRON UNIT 2 PRESSURE TEMPERATURE LIMITS REPORT (PTLR)

(April 2003)

411.

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) 1 2.0 Operating Limits 1 2.1 RCS Pressure and Temperature (PfI) Limits (TS-LCO 3.4.3) 1 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints TS-LCO 3.4.12. 2 2.3 LTOP Enable Temperature 2 2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification) 2 2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification) 3 3.0 Reactor Vessel Material Surveillance Program 9 4.0 Supplemental Data Tables 10 5.0 References 17 List of Figures Figure Page 2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 1000 F/hr) 4 Applicable for 15.5 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates 5 up to 100 'F/hr) Applicable for 15.5 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) 2.3 Byron Unit 2 Maximum Allowable Nominal PORV Setpoints for the Low 7 Temperature Overpressure Protection (LTOP) System Applicable for 15.5 EFPY 11

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 Byron Unit 2 Heatup and Cooldown Data Points at 15.5 EFPY (Without Margins 6 for Instrumentation Errors and Using the 1996 Appendix G Methodology) 2.2 Data Points for Byron Unit 2 Maximum Allowable PORV Setpoints for the LTOP 8 System Applicable for 15.5 EFPY 3.1 Byron Unit 2 Capsule Withdrawal Schedule 9 4.1 Calculation of Chemistry Factors for Byron Unit 2 Using Surveillance Capsule Data 11 4.2 Reactor Vessel Beltline Material Unirradiated Toughness Properties 12 4.3 Summary of Adjusted Reference Temperature (ART) at the 1/4T and 3/4T Locations 13 for 15.5 EFPY 4.4 Byron Unit 2 Calculation of Adjusted Reference Temperature (ART) at 15.5 EFPY at 14 the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data) 4.5 Table 4.5: RTPTS for Byron Unit 2 Beltline Region Materials - 32 EFPY 15 4.6 Table 4.6: RTPTS for Byron Unit 2 Beltline Region Materials at Life 16 Extension - 48 EFPY iii

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

This PTLR for Unit 2 has been prepared in accordance with the requirements of TS-5.6.6 (RCS Pressure and Temperature Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

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

2.0 Operating Limits The PTLR limits for Byron Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-NP-A (Reference 1)was used with the following exception:

a) Use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda, This exception to the methodology in WCAP-14040-NP-A has been reviewed and accepted by the NRC in Reference 17.

WCAP-15178, Reference 14, provides the basis for the Byron Unit 2 P/T curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. Reference 20 evaluated the effect of higher fluence from the 5%

uprate on the existing PT curves. The weld metal data integration for Byron and Braidwood Units 1 and 2 is documented in Reference 2.

2.1 RCS Pressure and Temperature (P/T) Limits (TS-LCO 3A.3) 2.1.1 The RCS temperature rate-of-change limits defined in Reference 14 are:

a. A maximum heatup of 1000 F in any 1-hour period,
b. A maximum cooldown of 100F in any l-hour period, and
c. A maximum temperature change of less than or equal to 10F in any l-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defined in Reference 14. Consistent with the methodology described in Reference 1, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda. The criticality limit curve specifies pressure-temperature limits for core I

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2.0 Operating Limits (continued) operation to provide additional margin during actual power production as specified in 10 CFR 50, Appendix G.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40'F higher than the minimum permissible temperature in the corresponding P/T curve for heatup and cooldown.

2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints TS-LCO 3.4.12.

The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 2.3 and Table 2.2. These limits are based on References 5, 13 and 15. The Residual Heat Removal (RH) Suction Relief Valves are also analyzed to individually provide low temperature overpressure protection. This analysis for the RH Suction Relief Valves remains valid with the current Appendix G limits contained in this PTLR document and will be reevaluated in the future as the Appendix G limits are revised. I The LTOP setpoints are based on P/T limits which were established in accordance with 10 CFR 50, Appendix G without allowance for instrumentation error and in accordance with the methodology described in Reference 1. The LTOP PORV 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 (References 14 and 16).

The required enable temperature for the PORVs shall be

  • 3500 F RCS temperature.

(Byron Unit 2 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP System for RCS temperature of 350'F and below and disarming of LTOP for RCS temperature above 350'F).

Note that the last LTOP PORV segment in Table 2.2 extends to 4500 F where the pressure setpoint is 2350 psig. This is intended to prohibit PORV lift for an inadvertent LTOP system arming at power.

2.4 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be 2 601F.

Boltup is a condition in which the Reactor Vessel head is installed with tension applied to any stud, and with the RCS vented to atmosphere (Reference 2).

2

I BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2.5 Reactor Vessel Minimum Pressurization Temperature (Non-Technical Specification)

Based on the steady-state limits specified in Table 2.1, the minimum temperature at which the Reactor Vessel may be pressurized (i.e., in an unvented condition) shall be 2 60'F, plus an allowance for the uncertainty of the temperature instrument, determined using a technique consistent with ISA-S67.04-1994.

3

I BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT C_ I I I I I I I III I II IIITTT I 11I I I I II M-I F~i II l I

A L

I-I I-so 0-ft ;f I I I I4 1.. an'r4 o IaPMC I 0-L -

I la-a WARPME~~~~~~

UDO n H!H Rj~~~

aCU W rrr . . . . .l l I

III 0 :LFFJ AQDPM

,,, ,, I I M13 I- X~~~~~~~~~~ I IaLlI IL

.C I E- Ui 5 GUGUMW~~~~~~~~~~i'N'D~~~~~~~~~IN3RAM~~~~~~~

0- I 0 O 5 0 1 5) 23 ) 303 3 4 )) 4 3 5 Figure 2.1:

Byron Unit 2 Reactor Coolant System Heatup Limitations (eatup rates up to 100 F/hr) Applicable for 15.5 EFPY (Without margins for instrumentation errors and using 1996 Appendix G Methodology) 4

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2250-WXXOTAELE CFEmnCN gn - . . . . . . -- -

LIIZIII 4-AM I IIII II -If C I0I I II I I I I I I II I I 1/ 1J1 IELE 2 _0_ _ _ _ _ _ _ I I I 1X1V 1:1X1 11CI I II1 a

's a

E 0- I- i 0 9D 100 im 20 30o3D 40D MotrxTaepw(=gF)

Figure 2.2:

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

Applicable for 15.5 EFPY (Without Margins for Instrumentation Errors and using 1996 Appendix G Methodology) 5

. - I BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1: Byron Unit 2 Heatup and Cooldown Data Points at 15.5 EFPY (Without Margins for Instrumentation Errors and Using the 1996 Appendix G Methodology) tip~uP Cooldown Curves 100 F Onticafity Leak Test Steady 25 F 5 F 100 F Heatup Unit Unit State Cooidom Cooidod n CooIdkami T P T P T P T P TTIP T P T P 60 219 0 198 2000 60 60 0 60 60 60 621 219 635 219 2485 60 621 60 574 60 523 60 418 65 621 219 6741 65 621 65 585 65 534 65 431 85 621 219 660 70 621 70 597 70 54 7 446 90 621 219 6501 75 621 75 610 75 561 75 462 95 621 219 643 80 621 80 621 80 576 80 100 621 219 638 85 621 85 621 85 592 85 498 105 621 21 637 90 621 90 621 90 609 90 519 110 621 219 637 95 621 95 621 95 621 95 541 115 621 219 641 - 100 621 100 621 100 621 100 564 120 621 219 646 105 621 105 621 105 621 105 590 125 621 219 654 - 11 621 110 621 110 621 110 618 130 621 219 664 115 621 115 621 115 621 115 621 1 621219 9 67 120 621 120 621 120 621 120 621 1 621 219 6- 125 621 125 621 125 621 125 621 14 621 219 707 130 621 130 621 130 621 130 621 150 621 219 725 135 621 135 621 135 621 135 621 150 707 219 746 = 140621 140 621 140 621 140 621 155 725 219 770 = 145 621 145 621 145 621 145 621 160 746 219 796 150 62 150 621 150 621 150 621 165 770 219 8241 150 995 150 974 150 957 1 935 170 796 220 855 = = 155 1033 155 1016 155 1002 15 989 175 824 225 889 1601075 1601061 160 1051 16 1 1EC 855 230 926 == 165 1119 165 1109 165 1104 165 1112 17L 889125 9661 =170 1166 1701161 170 1161 190 926 240 1010 175 1217 175 1217 19 66 245 1057 = 180 1272 200 101C 250 1107 = 185 1331 205 1057 255 1162 1990 15 210 1107 260 1221 = = 195 1463 215 1162 265 1285 - 200 1536 220 1221 270 1353 205 1615 225 1285 275 14271 210 1700 230 1353 280 1506 215 1791 235 1427 285 1591 - 220 1889 I 240 1506 290 16831 225 1995 _

245 1591 295 1781 - 230 2108 250 163 30 18 235 2230= = = = = _

255 1781 305 2001 - - 240 2361 260 1887 310 212 = = = = = = = = = =

265.2001. 315 2254

~270 21231 320= 2395= -_======= = = = = = = = =

275 22594 =H 0

Note 1: Heatup and Cooldown data includes the vessel flange requirements of 180 F and 621 psig per IOCFR50, Appendix G.

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

Note 3: Temperatures and pressures are given in F and psig, respectively.

6

's I BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 650 600

.0 550 0 500 9

C3 450 400 0 50 100 150 200 250 300 350 AUCTIONEERED LOW RCS TEMPERATURE (DEG. F)

Figure 2.3:Byron Unit 2 Maximum Allowable Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for the First 15.5 EFPY 7

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2:Data Points for Byron Unit 2 Maximum Allowable PORVSetpoints for the LTOP System Applicable for 15.5 EFPY PCV455A PCV456 (2TY-0413M) (2TY-0413P)

AUCTIONEERED RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE LOW RCS TEMP. (DEG. F) (PSIG) RCS TEMEP. (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 setpoints for RCS Pressure and RCS Temperatures greater than 3501F, linearly interpolate between the 3501F and 4501F data points shown above. (Setpoints extend to 4501F to prevent PORV liftoff from an inadvertent LTOP system arming while at power.)

8

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 6) is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME Section m, NB-233 1. The empirical relationship between RTND and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G, "Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

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

Table 3.1: Byron Unit 2 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Removal Time(a) Estimated Capsule (Degrees) Factor (EFPY) Fluence (n/cm 2 ) (c)

U 58.50 4.40 1.15 (Removed) 4.05 x loll W 121.50 4.25 4.634 (Removed) 1.27 x 10'9 X 238.50 4.25 8.573 (Removed at 2.30 x IO"l (b)

EOL Wall)

Z 301.50 4.21 Standby (c) 3.35 x 10'9 (c)

V 61.00 3.97 Standby (c)

Y 241.00 3.97 Standby (c) a) Effective Full Power Years (EFPY) from plant startup.

b) Maximum end of license (32 EFPY) inner vessel wall fluence is estimated to be 2.06 x 1019 n/cm 2.

c) Standby capsule to be used for future license renewal (derived from Table 7-1 of WCAP 15176, Ref. 18).

9

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Supplemental Data Tables The following tables provide supplemental information on reactor vessel material properties and are provided to be consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the P/T limits.

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

Table 4.2 provides the reactor vessel material properties table.

Table 4.3 provides a summary of the Byron Unit 2 adjusted reference temperature (ARTs) at the I/4T and 3/4T locations for 15.5 EFPY.

Table 4.4 shows the calculation of ARTs at 15.5 EFPY for the limiting Byron Unit 2 reactor vessel material, i.e. weld metal HT # 442002, (Based on Surveillance Capsule Data).

Table 4.5 provides RTpTs values for Byron Unit 2 for 32 EFPY obtained from Reference 9.

Table 4.6 provides RTprs values for Byron Unit 2 for 48 EFPY obtained from Reference 9.

10

r -.

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4-1: Calculation of Chemistry Factors Using Surveillance Capsule Data (a)

Material Capsule Fluence FF(A) Measured FF*ARTNDT FF2 (n/cm2 , &RTNDTSb)

__ E>l.OMeV)

Lower Shell Forging U 4.05*10"8 0.749 0.0 0 0.561 49D330/49C298-1-1 W 1.27*109 1.067 3.65 3.89 1.138 (Tangential) X 2.30*109 1.225 15.75 19.29 1.500 Lower Shell U 4.05*1018 0.749 19.76 14.80 0.561 Forging 49D330/ W 1.27*1019 1.067 31.88 34.02 1.138 49C298-1-1 X 2.30*1019 1.225 38.91 47.66 1.500 SUM: 119.66 6.398 CFForgp8 =ff

  • RTw) + 7( FF2) = (119.66) + (6.398) 18.7 0F Byron Unit I Surv. Weld 4 04 *1 01o 0.749 11.22 (5.61)0) 8.40 0.561 Material (Heat # 442002) X 1.57*1019 1.125 8 0. 2 2 (4 0 .11 )(b) 90.25 1.266 W 2.43*1019 1.239 102.68 (51.34)m° 127.22 1.535 Byron Unit 2 Surv. Weld U 4.05*1028 0.749 16.88 ( 8 .4 4 )) 12.64 0.561 Material (Heat # 442002) W 1.27*1019 1.067 57.76 (28.88)M) 61.63 1.138 X 2.30*10'9 1.225 108.02 (54.01)0') 132.32 1.500 SUM: 432.46 6.561 CFsu.Weld. 442002= (FF*RTNmg) + X(FF2 )=(432.46)+ (6.561)= 65.9oF;{e a) Reference 14, Table 4-8 b) ARTNm values are the measured 30 ft-lb shift values taken from Ref 18.

11

': -ri - - . I - -

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2: Reactor Vessel Beltline Material Unirradiated Toughness Properties (a)

Material Description Cu (%) Ni (%) Initial RT NDT ( 0F)

Closure Head Flange 5P7382 / 3P6407 _ 0.71 0 Vessel Flange 124L556VA1 0.70 30 Nozzle Shell Forging 0.05 0.74 10 4P-6107 Inter. Shell Forging 49D329-1-1/49C297-1-1 0.01 0.70 -20 Lower Shell Forging 49D330-1-1/49C298-1-1 0.06 0.73 -20 Circumferential Weld WF-447 (HT# 442002) 0.04 0.63 10 Upper Circumferential Weld WF-562 (HT# 442011) 0.03 0.67 40 Byron Unit 1 Surveillance Program 0.02 0.69 Weld Metal (Heat # 442002)

Byron Unit 2 Surveillance Program 0.02 0.71 Weld Metal (Heat # 442002)

Braidwood Units 1 & 2 Surveillance Program 0.03 0.67, Weld Metal (Heat # 442002) 0.71 a) Reference 14.

12

!- , 4. -.

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.3:Summary of Adjusted Reference Temperature (ART) at 1/4T and 3/4TLocation for 15.5 EFPY (a) 15.5 EFPY Material Description 1/4T ART(-F) 314T ART(-F)

Intermediate Shell Forging 49D329- 14 4 1/49C297-1 (RG Position 1 )_

Lower Shell Forging 49D330-1/49C298-1 43 24 (RG Position l)

Using capsule data 12 2 (RG Position 2(b)

Circumferential Weld 102 73 WF-447 (HT# 442002)

(RG Position et(I Using credible surveillance capsule data 94(C) 77(c)

(RG Position 2(b))

Nozzle Shell Forging 41 29 4P-6107 (RG Position 1 (b_)

Nozzle Shell to Intermediate Shell Weld 82 65 WF-562 (HT # 442011)

Using credible surveillance capsule data 57 50 (RG Position 2°b)

(a) Fluence, f, is based upon f,,f(E>1.0 Mev) = 9.86xlOl' at 15.5 EFPY, Reference 20.

(b) Calculated using a chemistry factor based on Regulatory Guide (RG) 1.99, Positions 1 and 2, Reference 12, as reported in WCAP-15178, Reference 14.

(c) These ART values were used to generate the Byron Unit 2 Heatup and Cooldown Curves, WCAP-15178 (Reference 14).

13

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.4:Byron Unit 2 Calculation of Adjusted Reference Temperature (ART) at 15.5 EFPY at the Limiting Reactor Vessel MaterialWeld Metal (Based on Surveillance Capsule Data) (c)

Parameter Values Operating Time 15.5 EFPY Location(b) 1/4T ART 3/4T ART Chemistry Factor, CF ( 0F) 65.9 65.9 Fluence(f), n/cm2 5.92x10'8 2.14x10 8 (E>1.0 Mev)(a)

Fluence Factor, FF 0.853 0.586 ARTNDT= CFxFF(0 F) 56.2 38.6 Initial RTNDT., I(F) 10 10 Margin, M (F) 28.0 28.0 ART= I+(CF*FF)+M,OF 94 77 per RG_1.99,_Revision 2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

(a) Fluence, f, is based upon f..f (E>1.0 Mev) = 9.86xlO' at 15.5 EFPY, Reference 20.

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

(c) WCAP 15178 14

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.5: RTpTS for Byron Unit 2 Beltline Region Materials - 32 EFPY Material Fluence(s) FFb) CF ARTPrs(d) Margin ( 0f) RTNDTuJ) RTprsm0 wncM2, ( OF) (OF) ( (F)

OF E>1.O MeV)

Intermediate Shell Forging 2.06

  • 10'9 1.20 20 23.8 23.8 -20 28 Lower Shell Forging 2.06
  • i0'9 1.20 37 44.0 34 -20 58 Lower Shell Forging Using S/C 2.06
  • 10'9 1.20 18.7 22.3 17 -20 19 DataWO Nozzle Shell Forging 5.22 *108 0.818 31 25.0 25 10 60 Inter. to Lower Shell Circ. Weld 2.03
  • 10'9 1.19 54 63.7 56 10 130 Inter. to Lower Shell Circ. Weld 2.03
  • 10'9 1.19 65.9 77.8 28 10 116W Using S/C Data(')

Nozzle Shell to Inter. Shell Circ. 5.22

  • 108 0.818 41 33.1 33.1 40 106 Weld Nozzle Shell to Inter. Shell Circ. 5.22
  • 10's 0.818 16.7 13.5 13.5 40 67 Weld Using S/C Data(')

(a) Fluence projections for 32 EFPY from Byron 2 PTS report, WCAP-157177 (Reference 9)

(b) FF (Fluence Factor) = f0.2O.IO*Inf (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 12).

(d) ARTpTs = CF

  • FF (e) Initial RTmT values are measured values (See Table 4.2)

(I) RTyrs = RTNDTQJ) + ARTm + Margin (fF)

(g) Limiting RTprs is significantly less than the PTS Screening Criteria of 300 'F.

15

b j -

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6: RTprs for Byron Unit 2 Beltline Region Materials at Life Extension (48 EFPY) (a) (g)

Material Fluence(') FF) CF ART(d Margin (F) RTNDTve RTn~')

(n/cml, (OF) (OF) (0F) (OF)

E>1.O MeV)

Intermediate Shell Forging 2.98

  • 10' 1.29 20 25.8 25.8 -20 32 Lower Shell Forging 2.98
  • 10'9 1.29 37 47.7 34 -20 62 Lower Shell Forging Using S/C 2.98
  • 10'9 1.29 18.7 24.1 17 -20 21 Data (c)

Nozzle Shell Forging 7.53*101s 0.920 31 28.5 28.5 10 67 Inter. to Lower Shell Circ. Weld 2.93

  • 10'9 1.29 54 69.7 56 10 136 Inter. to Lower Shell Circ. Weld 2.93
  • 10'9 1.29 65.9 85 28 10 123 Using S/C Data~c _ _ _ __ _ _ _

Nozzle Shell to Inter. Shell Circ. 7.53*108 0.920 41 37.7 37.7 40 115 Weld Nozzle Shell to Inter. Shell Circ. 7.53*108 0.920 16.7 15.4 15.4 40 71 Weld Using S/C Data(')

(a) The fluence for 48 EFPY (Ref. 9) did not incorporate the 5% increase. However, this fluence value is greater than the end-of-life fluence (32 EFPY).

(b) FF (Fluence Factor) = f- 2 S"O0.O* f)

(c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 12).

(d) ART~rs = CF

  • FF (e) Initial RTNDT values are measured values (See Table 4.2)

(f) RTpTs = RTNDTQJ) + ARTrs + Margin (F)

(g) Limiting RTpTs is significantly less than the PTS Screening Criteria of 300 'F.

16

'3- ,-

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References

1. WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Andrachek, J.D.,

et. al., January 1996.

2. Deleted
3. WCAP-14064, "Analysis of Capsule W from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," Malone, M.J., et al., July 1994.
4. WCAP-12431, "Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," Terek, E., et al., October 1989.
5. Westinghouse Letter to Commonwealth Edison Company, CAE-96-106, "Byron Unit 1 and 2 LTOPS Setpoints Based on 10 and 12 EFPY P/T Limits," January 17, 1996.
6. WCAP-10398, "Commonwealth Edison Company, Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program," Singer, L.R., 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. Deleted
9. WCAP-15 177, "Evaluation of Pressurized Thermal Shock for Byron Unit 2," Revision 0, T. J.

Laubham, et al., September 2000.

10. 10 CFR Part 50, Appendix G, "Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, dated December 19, 1995.
11. 10 CFR 50.61, "Fracture Toughness Requirements for Protection Against Pressurized Thermal 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.
13. ComEd Calculation BRW-96-907IBYR 96-294, "Channel Accuracy for Power Operated Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 2 Original Steam Generators)" Revision 0.
14. WCAP- 15178, Revision 0, "Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," T. J. Laubham, et al., June 1999.

17

BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 References (continued)

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-97-21 1/CCE-97-290, "Byron and Braidwood Units I and 2 OTmetal Evaluation," November 7, 1997.
17. NRC Letter from R. A. Capra, NRR, to 0. D. Kingsley, Commonwealth Edison Co.,

"Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, and M98802)," January 21, 1998.

18. WCAP-15176, Revision 0, "Analysis of Capsule X from Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et al., March 1999.
19. Deleted
20. Westinghouse Calculation CN-EMT-01-8, "Braidwood Units 1 and 2, Development of New Pressure-Temperature Limit Curves and Evaluation of Byron Units 1 and 2 P-T Curve EFPY 18 Final
Retrieved from "https://kanterella.com/w/index.php?title=BYRON_2003-0064,_Revision_to_Reactor_Coolant_System_Pressure_%26_Temperature_Limits_Report&oldid=2650878"