Byron Station, Units 1 and 2, Pressure and Temperature ... · Limits Report). Revisions to the PTLR shall be provided to the NRC after issuance. The Technical Specifications addressed

Byron Generating Station

Exelon Generation®4.450 North German Church RoadByron, IL 61010-9794www.exeloncorp.com

March 27, 2014

LTR: BYRON 2014-0040File: 2.01 .0300

1.10.0101

United States Nuclear Regulatory CommissionAttention: Document Control DeskWashington, DC 20555-0001

Subject:

Byron Station, Units 1 and 2Facility Operating License Nos, NPF-37 and NPF-66NRC Docket No, STN 50-454 and 50-455

Pressure and Temperature Limits Report (PTLR) Revised for Negative PressureApplication during Reactor Coolant System Vacuum FillByron Station, Units 1 and 2

In accordance with Technical Specification 5.66, “Reactor Coolant System (RCS) Pressure andTemperature Limits Report (PTLR),” we are submitting the March 2014 revisions to the ByronStation Units 1 and 2 PTLR documents. The PTLRs are revised to change the lowest pressurevalue for Figure 2.1, “Reactor Coolant System Heatup Limitations” and Figure 2.2, “ReactorCoolant System Cooldown Limitations” from 0 psig to minus(-)14 psig, which is applicableduring vacuum fill of the Reactor Coolant System.

In addition, minor editorial corrections were made to the Unit 1 PTLR text to document thecorrect supporting references.

Should you have any questions concerning these reports, please contact Steven Gackstetter,Regulatory Assurance Manager, at (815) 406-2800.

Respectfully,

FAK/GC/sg

Attachments: 1. Byron Unit 1 Pressure and Temperature Limits Report, March 20142. Byron Unit 2 Pressure and Temperature Limits Report, March 2014

cc: Regional Administrator — NRC Region IIINRC Senior Resident Inspector — Byron Station

Site Vice PresidentByron Nuclear Generating Station

BYRON UNIT 1

PRESSURE AND TEMPERATURELIMITS REPORT

(PTLR)

(March 2014)

BYRON - UNIT 1PRESSURE AN]) TEMPERATURE LIMITS REPORT

Table of Contents

Section Page

1.0 Introduction 1

2.0 RCS Pressure and Temperature Limits 1

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 1

3.0 Low Temperature Over Pressure Protection and Boltup 7

3.1 LTOP System Setpoints (LCO 3.4.12) 7

3.2 LTOP Enable Temperature 7

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) 7

4.0 Reactor Vessel Material Surveillance Program 10

5.0 Supplemental Data Tables 12

6.0 References 17

BYRON - UNIT 1PRESSURE AND TEMPERATURE LIMITS REPORT

List of Figures

Figure Page

2.1 Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup 3Rates of 100°FIhr) Applicable for 32 EFPY (Without Margins forInstrumentation Errors)

2.2 Byron Unit 1 Reactor Coolant System Cooldown Limitations 4(Cooldown Rates of 0, 25, 50, and 100°F/hr) Applicable for 32 EFPY(Without Margins for Instrumentation Errors)

3.1 Byron Unit 1 Nominal PORV Setpoints for the Low Temperature 8Overpressure Protection (LTOP) System Applicable for 32 EFPY(Includes Instrumentation Uncertainty)

11


List of Tables

Table Page

2.la Byron Unit 1 Heatup Data Points at 32 EFPY (Without Margins 5for Instrumentation Errors)

2. lb Byron Unit 1 Cooldown Data Points at 32 EFPY (Without 6Margins for Instrumentation Errors)

3.1 Data Points for Byron Unit 1 Nominal PORV Setpoints for the 9LTOP System Applicable for 32 EFPY (Includes InstrumentationUncertainty)

4.1 Byron Unit 1 Surveillance Capsule Withdrawal Summary 11

5.1 Byron Unit 1 Calculation of Chemistry Factors Using 13Surveillance Capsule Data

5.2 Byron Unit 1 Reactor Vessel Material Properties 14

5.3 Summary of Byron Unit 1 Adjusted Reference Temperature 15(ART) Values at l/4T and 3/4T Locations for 32 EFPY

5.4 RTPTS Calculation for Byron Unit 1 Beitline Region Materials at 16EOL (32 EFPY)

111


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

This Pressure and Temperature Limits Report (PTLR) for Byron Unit 1 has been prepared inaccordance with the requirements of Byron TS 5.6.6 (RCS Pressure and Temperature LimitsReport). 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 (PIT) Limits; andTS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.

2.0 RCS Pressure and Temperature Limits

This section provides the Byron Unit 1 Heatup and Cooldown Limitations.

The PTLR limits for Byron Unit 1 were developed using a methodology specified in theTechnical Specifications. The methodology listed in WCAP- 14040-NP-A, Revision 2(Reference 1) was used with the following exceptions:

a) Optional use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda,b) Use of ASME Code Case N-640, “Alternative Reference Fracture Toughness for

Development of P-T Limit Curves, Section XI, Division 1”,c) Use of ASME Code Case N-588, “Alternative to Reference Flaw Orientation of

Appendix G for Circumferential Welds in Reactor Vessel, Section XI, Division 1”, andd) Elimination of the flange requirements documented in WCAP-16 143-P.

These exceptions to the methodology in WCAP-14040-NP-A, Revision 2 have beenreviewed and accepted by the NRC in References 6, 10, 11 and 12.

WCAP- 15391, Revision 1, Reference 7, provides the basis for the Byron Unit 1 P/T curves,along with the best estimate chemical compositions, fluence projections, and adjustedreference temperatures used to determine these limits. The weld metal data integration forByron and Braidwood Units 1 and 2 is documented in Reference 2. WCAP-16 143-P.Reference 11, documents the technical basis for the elimination of the flange requirements.

2.1 RCS Pressure and Temperature (PIT) Limits (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 of less than or equal to 10°F in any 1-hourperiod during inservice hydrostatic and leak testing operations above the heatupand cooldown limit curves.


2.1.2 The RCS P/T limits for beatup, inservice hydrostatic and leak testing, and criticalityare specified by Figure 2.1 and Table 2.1 a. The RCS PIT limits for cooldown areshown in Figure 2.2 and Table 2.lb. These limits are defined in WCAP-15391, Rev.1 (Reference 7). Consistent with the methodology described in Reference 1, the RCSPIT limits for heatup and cooldown shown in Figures 2.1 and 2.2 are providedwithout margins for instrument error. These limits were developed using ASMEBoiler and Pressure Vessel Code Section XI, Appendix G, Article G-2000, 1996Addenda. The criticality limit curve specifies pressure-temperature limits for coreoperation to provide additional margin during actual power production as specified in10 CFR 50, Appendix G.

The PIT limits for core operation (except for low power physics testing) are that thereactor vessel must be at a temperature equal to or higher than the minimumtemperature required for the inservice hydrostatic test, and at least 40°F higher thanthe minimum permissible temperature in the corresponding P/T curve for heatup andcooldown.

2


MATERIAL PROPERTY BASIS

LiMITING MATERIAL: INTERMEDIATE SHELL FORGING

LIMITING ART VALUES AT 32 EFPY: 1/4T, 106°F

314T, 97°F

Moderator Temperature (Deg. F)

Figure 2.1Byron Unit 1 Reactor Coolant System Heatup Limitations (Heatup rates of 100°FIhr)

Applicable for 32 EFPY (Without Margins for Instrumentation Errors)

2500

2250

2000

1750

15000

1250

• 1000a)4-’

750

500

250

0

LeakTestLimit f / fI Acceptable

OperationUnacceptable

.__OPeoJ__

Jeatup Rate

Deg. Frjcal Limit

.—-%-__

Temp. I inservice hydrostatic testtemperature (1 66SF) for thejOF

service period up to 32 EFPY

IOltUP

riticalityLtbason

The lower limit for RCSpressure is -14.7 psig

0 50 100 150 200 250 300 350 400 450 500 550

3



LIMITING MATERIAL: INTERMEDIATE SHELL FORGING

LIMITING ART VALUES AT 32 EFPY: 1/4T, 106°F

314T, 97°F

Moderator Temperature (Deg. F)

Figure 2.2Byron Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown rates of 0,25,50 and

100°F/hr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors)

2500

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2000

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-100 -

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—_________

The lower limit for RCS..._- jpressure is -14.7 psig

0 50 100 150 200 250 300 350 400 450 500 550

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3.0 Low Temperature Overpressure Protection and Boltup

This section provides the Byron Unit 1 power operated relief valve lift settings, lowtemperature overpressure protection (LTOP) system arming temperature, andminimum reactor vessel boltup temperature.

3.1 LTOP System Setpoints (LCO 3.4.12)

The power operated relief valves (PORVs) shall each have maximum lift settings inaccordance with Figure 3.1 and Table 3.1. These limits are based on References 3and 5.

The LTOP setpoints are based on PIT limits that were established in accordance with10 CFR 50, Appendix G without allowance for instrumentation error. The LTOPsetpoints were developed using the methodology described in Reference 1. TheLTOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account forappropriate instrument error.

3.2 LTOP Enable Temperature

The required enable temperature for the PORVs shall be 3 50°F RCS temperature.(Byron Unit 1 procedures governing the heatup and cooldown of the RCS require thearming of the LTOP System for RCS temperature of 3 50°F and below and disarmingof LTOP for RCS temperature above 350°F).

Note that the last LTOP PORV segment in Table 3.1 extends to 400°F where thepressure setpoint is 2335 psig. This is intended to prohibit PORV lift for aninadvertent LTOP system arming at power.

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F.Boltup is a condition in which the Reactor Vessel head is installed with tensionapplied to any stud, and with the RCS vented to atmosphere (Reference 7).

7

0

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a>

0a.CoC

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Figure 3.1Byron Unit 1 Nominal PORV Setpoints for the Low TemperatureOverpressure Protection (LTOP) System Applicable for 32 EFPY

(Includes Instrumentation Uncertainty)

0 50 100 150 200 250 300 350 400 450

Auctioneered Low RCS Temperature (DEG. F)

8


Table 3.1Data Points for Byron Unit 1 Nominal PORV Setpoints

for the LTOP System Applicable for 32 EFPY(Includes Instrumentation Uncertainty)

PCV-455A

(1TY-0413M)

AUCTIONEERED LOWRCS TEMP. (DEG. F)

300400

PCV-456

(1TY-0413P)

AUCTIONEERED LOWRCS TEMP. (DEG. F)

300400

Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greaterthan 300°F, linearly interpolate between the 300°F and 400°F data points shown above.(Setpoints extend to 400°F to prevent PORV liftoff from an inadvertent LTOP systemarming while at power.)

RCS PRESSURE(PSIG)

60 541541

2335

RCS PRESSURE(PSIG)

60 595595

2335

9


4.0 Reactor Vessel Material Surveillance Program

The pressure vessel material surveillance program (Reference 12) is in compliance withAppendix H to 10 CFR 50, “Reactor Vessel Radiation Surveillance Program.” Thematerial test requirements and the acceptance standards utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASME Boilerand Pressure Vessel Code, Section III, NB-2331. The empirical relationship betweenRTNDT and the fracture toughness of the reactor vessel steel is developed in accordancewith Appendix G, “Protection Against Non-Ductile Failure,” to Section XI of theASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedulemeets the requirements of ASTM E185-82.

The third and final reactor vessel material irradiation surveillance specimens (Capsule W)have been removed and analyzed to determine changes in the reactor vessel materialproperties. The surveillance capsule testing has been completed for the original operatingperiod. The remaining three capsules, V, Y, and Z, were removed and placed in the spentfuel pool to avoid excessive fluence accumulation should they be needed to support lifeextension. The removal summary is provided in Table 4.1.

10


Table 4.1

Byron Unit 1 Survefflance Capsule Withdrawal Summary

Capsule Capsule Lead Factor Withdrawal EFPY FluenceLocation (n/cm2,E> 1.0 MeV)

U 58.5° 4.05 1.18 0.409 x io’

X 238.5° 4.09 5.67 1.49

W 121.5° 4.08 9.27 2.26 x i019

z 301.5° 4.11 14.59 (EOC 12) 3.34 x 1019

v 61.00 3.89 14.59(EOC12) 3.16x iO’9

y(c) 241.0° 3.85 18.81 (EOC 15) 3.97 x i09

Notes:

(a) Source document is CN-AMLRS- 10-8 (Reference 4), Table 5.7-3.(b) Effective Full Power Years (EFPY) from plant startup.(c) Standby Capsules Z, V, and Y were removed and placed in the spent fuel pool. No testing or

analysis has been performed on these capsules. If license renewal is sought, one of thesestandby capsules may need to be tested to determine the effect of neutron irradiation on thereactor vessel surveillance materials during the period of extended operation.

11


5.0 Supplemental Data Tables

The following tables provide supplemental information on reactor vessel material propertiesand are provided to be consistent with Generic Letter 96-03. Some of the material propertyvalues shown were used as inputs to the PIT limits.

Table 5.1 shows the calculation of the surveillance material chemistry factors usingsurveillance capsule data.

Table 5.2 provides the reactor vessel material properties table.

Table 5.3 provides a summary of the Byron Unit 1 adjusted reference temperature (ART)values at the 114T and 314T locations for 32 EFPY.

Table 5.4 provides the RTfrrs values for Byron Unit 1 for 32 EFPY obtained fromReference 4.

12


Table 5.1

Byron Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data (a)

Capsule f(b)

FF ARTNDTb) FF*LRTT

FF2Material Capsule(n/cm2,E> 1.0 MeV) (°F) (°F)

U 4.09x i’ 0.752 28.55 21.47 0.57Intermediate

X 1.49x i’ 1.110 9.82 10.90 1.23Shell Forging

(Tangential) W 2.26 x 1.22 1 49.20 60.06 1.49

U 0.409 x 1019 0.752 18.52 13.93 0.57Intermediate

X l.49x 1019 1110 53.03 58.89 1.23Shell Forging

(Axial) W 2.26x i0 1.221 29.34 35.82 1.49

Sum: 201.06 6.58

CF IS Forging = XFF * t\RTNDT)÷ (jr2)

= (201.06) ÷ (6.58) = 30.6°F

11.22

U 4.09x iO’9 0.752 (5.61) 8.44 0.57Byron Unit 1

Surveillance 80.22Weld Material X 1.49 x 1.110 (40.11) 89.08 1.23

(Heat #442002)102.68

W 2.26x 1019 1.221 (51.34) 125.34 1.49

16.88

U 0.406 x iO’9 0.750 (8.44) 12.66 0.56Byron Unit 2

Surveillance 57.76Weld Material W 1.20x i0’ 1.051 (28.88) 60.70 1.10(Heat #442002)

108.02X 2.18x iO’9 1.211 (54.01) 130.86 1.47

SUM: 427.08 6.42

CF Weld Metal = (FF * LRTNDT) ÷(J2)

= (427.08) ÷ (6.42) 66.5°F

Notes:

a) Source document is CN-AMLRS- 10-8 (Reference 4), Table 5.2-1.b) f = fluence; L\RTNDT values are the measured 30 ft-lb shift values taken from Reference 13.

.RTNDT values for the surveillance weld data are adjusted by a ratio of 2.0 (pre-adjusted values arelisted in parentheses).

c) FF = fluence factor =

13


Table 5.2

Byron Unit 1 Reactor Vessel Material Properties (a)

InitialMaterial Description Cu (%) Ni (°“) RT NDT oFb

Closure Head Flange 124K358VA1 -- 0.74 60Vessel Flange 123J219VA1 -- 0.73 10

Nozzle Shell Forging 123J218 0.05 0.72 30Intermediate Shell Forging 5P-5933 0.04 0.74 40

Lower Shell Forging 5P-5951 0.04 0.64 10Intermediate to Lower Shell Forging Circ.

0.04 0.63 -30Weld_Seam WF-336_(Heat # 442002)Nozzle Shell to Intermediate Shell Forging

0.03 0.67 10Circ. Weld Seam WF-501 (Heat # 442011)Byron Unit 1 Surveillance Program

0.02 0.69 --Weld_Metal_(Heat # 442002)Byron Unit 2 Surveillance Program

0.02 0.71 --Weld_Metal_(Heat # 442002)Braidwood Units 1 & 2 Surveillance 0.67,

0.03Program Weld Metals (Heat # 442011) 0.71 --

a) Reference 7.

b) The initial RT NDT values for the plates and welds are based on measured data.

14

BYRON - UMT 1PRESSURE AND TEMPERATURE LIMITS REPORT

Table 5.3

Summary of Byron Unit 1 Adjusted Reference Temperature (ART) Values at1/4T and 314T Locations for 32 EFPY (a)

Surface Fluence 32 EFPYReactor Vessel Material

(n/cm2,E> 1.0 MeV) 114T ART (°F) 3/4T ART (°F)

Nozzle Shell Forging 0.598 x 1019 74 59

Intermediate Shell Forging 1.77 iO’9 93 78

* Using noncredible1.77 x 1019 102 85surveillance data

Lower Shell Forging 1.77 b’9 63 48

Nozzle to Intermediate Shell Forging0.598 x i’ 69 49Circ. Weld Seam (Heat #442011)

—+ Using credible Braidwood Units0.598 x 1O’ 47 351 and 2 surveillance data

Intermediate to Lower Shell Forging1.72x 1019 79 49Circ. Weld Seam (Heat # 442002)

* Using credible surveillance data 1.72 x 1019 65 46

Note:

(a) The source document containing detailed calculations is CN-AMLRS- 10-8 (Reference 4),Tables 5.3.1-1 and 5.3.1-2. The ART values summarized in this table utilize the most recentfluence projections and materials data, but were not used in development of the P/T limitcurves. See Figures 2.1 and 2.2 of this PTLR for the ART values used in development ofthe PIT limit curves.

15

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16


6.0 References

1. WCAP-14040-NP-A, Revision 2, “Methodology Used to Develop Cold Overpressure MitigatingSystem Setpoints and RCS Heatup and Cooldown Limit Curves”, J.D. Andrachek, et al., January1996.

2. WCAP-14824, Revision 2, “Byron Unit 1 Heatup and Cooldown Limit Curves for NormalOperation and Surveillance Weld Metal Integration for Byron & Braidwood”, November 1997with Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE-97-23 1/CCE-97-3 14 and CAE-97-233/CCE-97-3 16, dated January 6, 1998.

3. Westinghouse Letter to Exelon Nuclear, CAE-10-MUR-197, Revision 0, “Low TemperatureOverpressure Protection (LTOP) System Evaluation Final Letter Report,” M. P. Rudakewiz,September 8, 2010.

4. Westinghouse Calculation Note CN-AMLRS-10-8, Revision 0, “Byron Units 1 and 2Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations,”A. E. Leicht, September 2010.

5. Byron Station Design Information Transmittal DIT-BYR-06-046, “Transmittal of Byron Unit1 and Unit 2 Temperature and Pressure Uncertainties for Low Temperature OverpressureSystem (LTOPS) Power Operated Relief Valves (PORVS),” David Neidich, August 15, 2006.

6. NRC Letter from R. A. Capra, NRR, to 0. D. Kingsley, Commonwealth Edison Co., “ByronStation, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing ofPressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, andM98802),” January 21, 1998.

7. WCAP- 15391, Revision 1, “Byron Unit 1 Heatup and Cooldown Limit Curves for NormalOperation,” T. J. Laubham, et a!., November 2003.

8. NRC Letter from G. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC,“Issuance of Amendments: Revised Pressure-Temperature Limits Methodology; ByronStation, Units 1 and 2, and Braidwood Station, Units 1 and 2,” dated October 4, 2004.

9. NRC Letter from M. Chawla to O.D. Kingsley, Exelon Generation Company, LLC, “Issuanceof exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G for ByronStation, Units 1 and 2, and Braidwood Stations, Units 1 and 2,” dated August 8, 2001.

17


10. NRC Letter from R. F. Kuntz, NRR, to C. M. Crane, Exelon Generation Company, LLC,“Byron Station, Unit Nos. 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 - Issuance ofAmendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TACNos. MC8693, MC8694, MC8695, and MC8696),” November 27, 2006.

11. WCAP-16143-P, Revision 0, “Reactor Vessel Closure Head/Vessel Flange RequirementsEvaluation for Byron/Braidwood Units 1 and 2,” W. Bamford, et al., November 2003.

12. WCAP-9517, “Commonwealth Edison Company, Byron Station Unit 1 Reactor VesselSurveillance Program”, J.A. Davidson, July 1979.

13. WCAP-15123, Revision 1, “Analysis of Capsule W from Commonwealth Edison CompanyByron Unit 1 Reactor Vessel Radiation Surveillance Program,” T.J. Laubham, et al, January1999.

18

BYRON UNIT 2

PRESSURE AND TEMPERATURELIMITS REPORT

(PTLR)

(March 2014)


Table of Contents

Section Page

1.0 Introduction 1

2.0 RCS Pressure and Temperature Limits 1

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 1

3.0 Low Temperature Over Pressure Protection and Boltup 7

3.1 LTOP System Setpoints (LCO 3.4.12) 7

3.2 LTOP Enable Temperature 7

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification) 7

4.0 Reactor Vessel Material Surveillance Program 10

5.0 Supplemental Data Tables 12

6.0 References 17


List of Figures

Figure Page

2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup 3Rates of 100°FIhr) Applicable for 30.5 EFPY (Without Margins forInstrumentation Errors)

2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations 4(Cooldown Rates of 0, 25, 50, and 100°F/br) Applicable for 30.5 EFPY(Without Margins for Instrumentation Errors)

3.1 Byron Unit 2 Nominal PORV Setpoints for the Low Temperature 8Overpressure Protection (LTOP) System Applicable for 30.5 EFPY(Includes Instrumentation Uncertainty)

II


List of Tables

Table Page

2.la Byron Unit 2 Heatup Data Points at 30.5 EFPY (Without Margins 5for Instrumentation Errors)

2. lb Byron Unit 2 Cooldown Data Points at 30.5 EFPY (Without 6Margins for Instrumentation Errors)

3.1 Data Points for Byron Unit 2 Nominal PORV Setpoints for the 9LTOP System Applicable for 30.5 EFPY (IncludesInstrumentation Uncertainty)

4.1 Byron Unit 2 Surveillance Capsule Withdrawal Summary 11

5.1 Byron Unit 2 Calculation of Chemistry Factors Using 13Surveillance Capsule Data

5.2 Byron Unit 2 Reactor Vessel Material Properties 14

5.3 Summary of Byron Unit 2 Adjusted Reference Temperatures 15(ART) Values at 1/4T and 3/4T Locations for 32 EFPY

5.4 RTPTS Calculation for Byron Unit 2 Beitline Region Materials at 16EOL (32 EFPY)

111


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

This Pressure and Temperature Limits Report (PTLR) for Byron Unit 2 has beenprepared in accordance with the requirements of Byron TS-5.6.6 (RCS Pressure andTemperature Limits Report). Revisions to the PTLR shall be provided to the NRC afterissuance.

The Technical Specifications addressed in this report are listed below:

TS-LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; andTS-LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.

2.0 RCS Pressure and Temperature LimitsThis section provides the Byron Unit 2 Heatup and Cooldown Limitations.

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

a) Use of ASME Code Section XI, Appendix G, Article G-2000, 1996 Addenda,b) Use of ASME Code Case N-640, “Alternative Reference Fracture Toughness for

Development of P-T Limit Curves, Section XI, Division 1”,c) Use of ASME Code Case N-588, “Alternative to Reference Flaw Orientation of

Appendix G for Circumferential Welds in Reactor Vessels, Section XI, Division 1”,and

d) Elimination of the flange requirements documented in WCAP-16 143-P.

This exception to the methodology in WCAP-14040-NP-A, Revision 2 has beenreviewed and accepted by the NRC in References 8, 10, 11 and 12.

WCAP-15392, Revision 2 (Reference 7), provides the basis for the Byron Unit 2 P/Tcurves, along with the best estimate chemical compositions, fluence projections, andadjusted reference temperatures used to detenriine these limits. The weld metal dataintegration for Byron and Braidwood Units 1 and 2 is documented in Reference 2.WCAP- 16143 -P, Reference 13, documents the technical basis for the elimination of theflange requirements.

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3)

2.1.1 The RCS temperature rate-of-change limits defmed 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 of less than or equal to 10°F in any 1-hourperiod during inservice hydrostatic and leak testing operations above theheatup and cooldown limit curves.


2.1.2 The RCS PIT limits for heatup, inservice hydrostatic and leak testing, andcriticality are specified by Figure 2.1 and Table 2. la. The RCS PIT limits forcooldown are shown in Figure 2.2 and Table 2.lb. These limits are defined inWCAP-15392, Revision 2 (Reference 7). Consistent with the methodologydescribed in Reference 1, the RCS P/T limits for heatup and cooldown shown inFigures 2.1 and 2.2 are provided without margins for instrument error. Theselimits were developed using ASME Boiler and Pressure Vessel Code Section XI,Appendix G, Article G-2000, 1996 Addenda. The criticality limit curve specifiespressure-temperature limits for core operation to provide additional margin duringactual power production as specified in 10 CFR 50, Appendix G.

The P/T limits for core operation (except for low power physics testing) are thatthe reactor vessel must be at a temperature equal to or higher than the minimumtemperature required for the inservice hydrostatic test, and at least 40°F higherthan the minimum permissible temperature in the corresponding P/T curve forheatup and cooldown.

2



LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING

LIMITING ART VALUES AT 30.5 EFPY: 1/4T, 107°F (N-588) & 52°F (‘96 App. 0)

3/4T, 89°F (N-588) & 37°F (‘96 App. 0)

,Fiicceptable Ij Operation I FceptabIe

jH ICritical Limit

1OODe.FIHr1 /100 Deg FN/

—

* CritiImitb: onlvice hydi static test‘ Htemperature(112F)torthe

‘service period up to 305 EFP

+

HTernp.6F60°F

ILi_

low mitforRCSressure is -14.7 psig

,t., ,, I rn

Figure 2.1Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup rates of 100°FIhr)

Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors)

2500

2250

2000

1750

1500

12500

a. 1000

C.,

C)

750

500

250

0

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

400 450 500 550

3



LIMITING MATERIAL: CIRCUMFERENTIAL WELD WF-447 & NOZZLE SHELL FORGING

LIMITING ART VALUES AT 30.5 EFPY: 114T, 107°F (N-588) & 52°F (‘96 App. G)

3/4T, 89°F (N-588) & 37°F (‘96 App. G)

2500

2250

2000

1750

1000

750

500

250

0

Figure 2.2Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50and 100°FIhr) Applicable for 30.5 EFPY (Without Margins for Instrumentation Errors)

Cl) 15000

1250

1 1

*-

i— —jsteady-state,

1-25,I-SO, and

L

J BoltupTemp. 6OF

The lower limit for RCS— ._1jpressure is -14.7 psig

I 333333)3313 TT’1 71r

0 50 100 150 200 250 300 350 400 450 500 550Moderator Temperature (Deg. F)

4

C

I3

t’J

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Table 2.lbByron Unit 2 Cooldown Data Points at 30.5 EFPY

(Without Margins for Instrumentation Errors)

Cooldown Curves

Steady State 25 °F Cooldown 50 °F Cooldown 100 °F Cooldown

T (°F) P (psig) T (°F) P (psig) T (°F) P (psig) T (°F) P (psig)60 -14.7 60 -14.7 60 -14.7 60 -14.760 1045 60 1036 60 103365 1092 65 108870 114375 120080 126385 133290 140995 1494100 1587105 1691110 1805115 1932120 2071125 2226130 2396

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

6


3.0 Low Temperature Overpressure Protection and Boltup

This section provides the Byron Unit 2 power operated relief valve lift settings,low temperature overpressure protection (LTOP) system arming temperature, andminimum reactor vessel boltup temperature.

3.1 LTOP System Setpoints (LCO 3.4.12)

The power operated relief valves (PORVs) shall each have maximum lift settings inaccordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 6.

The LTOP setpoints are based on PIT limits which were established in accordance with10 CFR 50, Appendix G without allowance for instrumentation error and in accordancewith the methodology described in Reference 1. The LTOP PORV nominal lift settingsshown in Figure 3.1 and Table 3.1 account for appropriate instrument error.

3.2 LTOP Enable Temperature

The required enable temperature for the PORVs shall be 350°F RCS temperature.(Byron Unit 2 procedures governing the heatup and cooldown of the RCS require thearming of the LTOP System for RCS temperature of 3 50°F and below and disarming ofLTOP for RCS temperature above 350°F).

Note that the last LTOP PORV segment in Table 3.1 extends to 400°F where thepressure setpoint is 2335 psig. This is intended to prohibit PORV lift for aninadvertent LTOP system arming at power.

3.3 Reactor Vessel Boltup Temperature (Non-Technical Specification)

The minimum boltup temperature for the Reactor Vessel Flange shall be 60°F. Boltupis a condition in which the Reactor Vessel head is installed with tension applied to anystud, and with the RCS vented to atmosphere (Reference 7).

7


2500

2335 psig2250

2000

175000

e 1500

Unacceptable Operation

1250>

00.

• 1000PCV 456

750 639 psig 4,500

PCV 455A

250

00 50 100 150 200 250 300 350 400 450

Auctioneered Low RCS Temperature (DEG. F)

Figure 3.1Byron Unit 2 Nominal PORV Setpoints for the Low Temperature

Overpressure Protection (LTOP) System Applicable for the 30.5 EFPY(Includes Instrumentation Uncertainty)

8


Table 3.1Data Points for Byron Unit 2 Nominal PORV Setpoints

for the LTOP System Applicable for 30.5 EFPY(Includes Instrumentation Uncertainty)

PCV-455A(2TY-041 3M)

AUCTIONEERED LOW RCS PRESSURERCS TEMP. (DEG. F) (PSIG)

50 599300400

5992335

PCV-456(2TY-041 3P)

AUCTIONEERED LOW RCS PRESSURERCS TEMP. (DEG. F) (PSIG)

50 639300400

6392335

Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperaturesgreater than 300°F, linearly interpolate between the 300°F and 400°F data points shownabove. (Setpoints extend to 400°F to prevent PORV liftoff from an inadvertent LTOPsystem arming while at power.)

9


4.0 Reactor Vessel Material Surveillance Program

The pressure vessel material surveillance program (Reference 4) is in compliance withAppendix H to 10 CFR 50, “Reactor Vessel Radiation Surveillance Program.” Thematerial test requirements and the acceptance standard utilize the reference nil-ductilitytemperature, RTNDT, which is determined in accordance with ASME Boiler and PressureVessel Code Section III, NB-233 1. The empirical relationship between RTNDT and thefracture toughness of the reactor vessel steel is developed in accordance with AppendixG, “Protection Against Non-Ductile Failure,” to Section XI of the ASME Boiler andPressure Vessel Code. The surveillance capsule removal schedule meets therequirements of ASTM E185-82.

The third and final reactor vessel material irradiation surveillance specimens (Capsule W)have been removed and analyzed to determine changes in the reactor vessel materialproperties. The surveillance capsule testing has been completed for the original operatingperiod. The remaining three capsules, V, Y and Z, were removed and placed in the spentfuel pool to avoid excessive fluence accumulation should they be needed to support lifeextension. The removal summary is provided in Table 4.1.

10


Table 4.1

Byron Unit 2 Surveillance Capsule Withdrawal Summary

Capsule Capsule Lead Factor Withdrawal EFPY FluenceLocation (n/cm2,E> 1.0 MeV)

U 58.5° 4.02 1.19 0.406 x iO’9

W 121.5° 4.07 4.67 1.20 x 1019

X 238.5° 4.14 8.63 2.18 x i019

z 301.5° 4.11 14.28 (EOC 11) 3.25 x

v 61.0° 3.88 14.28 (EOC 11) 3.07 x

y{c) 241.0° 3.88 20.05 (EOC 15) 4.19 x

Notes:

(a) Source document is CN-AMLRS- 10-8 (Reference 5), Table 5.7-4.(b) Effective Full Power Years (EFPY) from plant startup.(c) Standby Capsules Z, V, and Y were removed and placed in the spent fuel pool. No

testing or analysis has been performed on these capsules. If license renewal is sought,one of these standby capsules may need to be tested to determine the effect of neutronirradiation on the reactor vessel surveillance materials during the period of extendedoperation.

11

BYRON - UMT 2PRESSURE AND TEMPERATURE LIMITS REPORT

5.0 Supplemental Data Tables

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

Table 5.1 shows the calculation of the surveillance material chemistry factors usingsurveillance capsule data.

Table 5.2 provides the reactor vessel material properties table.

Table 5.3 provides a summary of the Byron Unit 2 adjusted reference temperature(ART) values at the 1/4T and 3/4T locations for 32 EFPY.

Table 5.4 provides the RTprs values for Byron Unit 2 for 32 EFPY obtained fromReference 5.

12

BYRON - UNIT 2

PRESSURE AND TEMPERATURE LIMITS REPORT

Table 5.1

Byron Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data(a)

Capsule f (b)

FF ARTNDTb) Jl?*RTNDT

FF2Material Capsule(n/cm2,E> 1.0 MeV) (°F) (°F)

U 4.06 x i0’ 0.750 00(d) 0.00 0.56

Lower Shell Forging W 1.20 , iO’9 1.051 3.65 3.84 1.10(Tangential)

X 2.18x i0’ 1.211 15.75 19.08 1.47

Lower Shell U 04.06 x i0 0.750 19.76 14.82 0.56

Forging W 1.20x 1019 1.051 31.88 33.50 1.10(Axial)

X 2.18x 1019 1.211 38.91 47.14 1.47

SUM: 118.38 6.27

CFLs Forging = (FF *t\RTNDT) ÷ ( FF2) = (118.38) ÷ (6.27) = 18.9°F

11.22U 0.409 x iO’9 0.752 (5.61) 8.44 0.57

Byron Unit 1Surveillance Weld 80.22

X 1.49x 1019 1.110 (40.11) 89.08 1.23Material

102.68(Heat #442002) W 2.26x iO’9 1.221 (51.34) 125.34 1.49

16.88U 0.406 x iO’9 0.750 (844) 12.66 0.56

Byron Unit 2Surveillance Weld 57.76

W 1.20x 1019 1.051 (28.88) 60.70 1.10Material

108.02(Heat#442002) X 2.18x l0’ 1.211 (5401) 130.86 1.47

SUM: 427.08 6.42

CFwe1dMet = (FF * ARTNDT) + ( Ff2) = (427.08) ÷ (6.42) = 66.5°F

Notes:

a) Source document is CN-AMLRS-10-8 (Reference 5), Table 5.2-2.b) f = fluence; ARTNDT values are the measured 30 ft-lb shift values taken from Reference 9.

LRTNOT values for the surveillance weld data are adjusted by a ratio of 2.0 (pre-adjusted values arelisted in parentheses).

c) FF = fluence factor = O.28 - O.1O*log 1)

d) Measured .RTNDT value was determined to be negative, but physically a reduction should not occur;therefore a conservative value of zero is used.

13


Table 5.2

Byron Unit 2 Reactor Vessel Material Properties (a)

. . . . InitialMaterial Description Cu (%) Ni (%) , (b)ici. NDT(. r)Closure Head Flange 5P7382 / 3P6407 -- 0.71 0

Vessel Flange 124L556VA1 -- 0.70 30Nozzle Shell Forging 4P-6 107 0.05 0.74 10

Inter. Shell Forging [49D329/49C297}-l-l 0.01 0.70 -20Lower Shell Forging [49D330149C298]-1-1 0.06 0.73 -20

Circumferential Weld WF-447 (HT# 442002) 0.04 0.63 10Upper Circumferential Weld WF-562 (HT# 442011) 0.03 0.67 40

Byron Unit 1 Surveillance Program0 02 0 69 --Weld_Metal_(Heat_# 442002)

Byron Unit 2 Surveillance Program002 071 --Weld_Metal_(Heat # 442002)

Braidwood Units 1 & 2 Surveillance Program0 03

0.67, --

Weld Metal (Heat # 442011) 0.71

a) Reference 7.b) Initial RTNDT values are based on measured data.

14


Table 5.3

Summary of Byron Unit 2 Adjusted Reference Temperatures (ART) Values at1/4T and 314T Locations for 32 EFPY (a)

Surface Fluence 32 EFPYReactor Vessel Material 2(n/cm , E> 1.0 MeV) 1/4T ART (°F) 314T ART (°F)

Nozzle Shell Forging 0.549 io’ 53 38

Intermediate Shell Forging 1.76 21 9

Lower Shell Forging 1.76 x 52 34

Using credible surveillance data 1.76 > lO’ 16 8

Nozzle to Intermediate Shell ForgingCirc. Weld Seam 0.549 x l0’ 97 77(Heat # 44201 1)

*Using credible Braidwood Units 1 0.549 io’ 76 64and 2 surveillance data

Intermediate to Lower Shell ForgingCirc. Weld Seam 1.70 x 119 88(Heat # 442002)

+Using credible surveillance data 1.70 X iO’9 105 86

Note:

(a) The source document containing detailed calculations is CN-AMLRS- 10-8 (Reference 5),Tables 5.3.1-3 and 5.3.1-4. The ART values summarized in this table utilize the mostrecent fluence projections and materials data, but were not used in development of theP/T limit curves. See Figures 2.1 and 2.2 of this PTLR for the ART values used indevelopment of the PIT limit curves.

15

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6.0 References

1. WCAP-14040-NP-A, Revision 2, “Methodology Used to Develop Cold OverpressureMitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,” Andrachek, J.D.,et a!., January 1996.

2. WCAP-14824, Revision 2, “Byron Unit 1 Heatup and Cooldown Limit Curves for NormalOperation and Surveillance Weld Metal Integration for Byron & Braidwood”, November1997 with Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE-97-23 1ICCE-97-3 14 and CAE-97-233/CCE-97-316, dated January 6, 1998.

3. Westinghouse Letter to Exelon Nuclear, CAE-10-MUR-197, Revision 0, “Low TemperatureOverpressure Protection (LTOP) System Evaluation Final Letter Report,” M. P. Rudakewiz,September 8, 2010.

4. WCAP-10398, “Conunonwealth Edison Company, Byron Station Unit 2 Reactor VesselRadiation Surveillance Program,” Singer, L.R., December 1983.

5. Westinghouse Calculation Note CN-AMLRS-10-8, Revision 0, “Byron Units 1 and 2Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations,”A. E. Leicht, September 2010.

6. Byron Station Design Information Transmittal DIT-BYR-06-046, “Transmittal of Byron Unit 1and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System(LTOPS) Power Operated Relief Valves (PORVS),” David Neidich, August 15, 2006.

7. WCAP- 15392, Revision 2, “Byron Unit 2 Heatup and Cooldown Limit Curves for NormalOperation,” T. J. Laubham, et al., November 2003.

8. NRC Letter from R. A. Capra, NRR, to 0. D. Kingsley, Commonwealth Edison Co., “ByronStation, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing ofPressure Temperature Limits Report (TAC Numbers M98799, M98800, M98801, andM98802),” January 21, 1998.

9. WCAP-15 176, Revision 0, “Analysis of Capsule X from Commonwealth Edison CompanyByron Unit 2 Reactor Vessel Radiation Surveillance Program,” T. J. Laubham, et al., March1999.

10. NRC Letter from 0. F. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC,“Issuance of Amendments: Revised Pressure-Temperature Limits Methodology; Byron Station,Units 1 and 2, and Braidwood Station, Units 1 and 2,” dated October 4, 2004.

11. NRC Letter from M. Chawla to O.D. Kingsley, Exelon Generation Company, LLC,“Issuance of exemption from the Requirements of 10 CFR 50 Part 60 and Appendix G forByron Station, Units 1 and 2, and Braidwood Stations, Units 1 and 2,” dated August 8,2001.

17

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