BW210065, Pressure and Temperature Limits Report, Revision 8

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Pressure and Temperature Limits Report, Revision 8
ML21300A076
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Site: Braidwood Constellation icon.png
Issue date: 10/27/2021
From:
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To:
Office of Nuclear Reactor Regulation
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BW210065
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BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Revision 8

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section 1.0 Introduction 2.0 RCS Pressure Temperature Limits 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 3.0 Low Temperature Over Pressure Protection and Boltup 3.1 LTOP System Setpoints (LCO 3.4.12) 3.2 LTOP Enable Temperature Page 1

1 7

7 7

3.3

  • Reactor Vessel Boltup Temperature (Non-Technical Specification) 7 4.0 Reactor Vessel Material Surveillance Program 5.0 Supplemental Data Tables 6.0 References 10 12 20

Figure

. 2.1 2.2 3.1 BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr) Applicable for 57 EFPY (Without Margins for Instrumentation Errors)

Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, and 100°F/hr) Applicable to 57 EFPY (Without Margins for Instrumentation Errors)

Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 57 EFPY (Includes Instrumentation Uncertainty) ii Page 3

4 8

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page

2. la Braidwood Unit 2 Heatup Data Points at 57 EFPY (Without 5

Margins for Instrumentation Errors) 2.1 b Braidwood Unit 2 Cooldown Data Points 57 EFPY (Without 6

Margins for Instrumentation Errors) 3.1 Data Points for Braidwood Unit 2 Nominal PORV 9

Setpoints for the L TOP System Applicable for 57 EFPY

  • (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit 2 Capsule Withdrawal Summary 11 5.1 Braidwood Unit 2 Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5.2 Braidwood Unit 2 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit 2 Adjusted Reference 16 Temperature (ART) Values at l/4T and 3/4T Locations for 57 EFPY 5.4 Braidwood Unit 2 Calculation of Adjusted Reference 17 Temperatures (AR Ts) at 57 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7056 5.5 RTPTS Calculation for Braidwood Unit 2 Beltline Region 18 Materials at EOLE (57 EFPY) iii

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 2 has been prepared in accordance with the requirements of Braidwood Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)". Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

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

2.0 RCS Pressure Temperature Limits The PTLR limits for Braidwood Unit 2 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-14040-A, Revision 4 (Reference 1) was used with the following exception:

a) Elimination of the flange requirements documented in WCAP-16143-P.

b) The initial reference temperatures of the inlet/outlet nozzle forging to shell welds are determined using BA W-2308 in lieu of the ASME NB-2300 requirements.

WCAP-183 70-NP, Revision O (Reference 7), provides the basis for the Braidwood Unit 2 P/T curves, along with the best estimate chemical compositions, fluence projections and adjusted reference temperatures used to determine these limits. The "Master Curve" fracture toughness properties from BAW-2308 Revision 1-A Safety Evaluation (SE) and Revision 2-A SE (Reference 2) are used for the inlet/outlet nozzle to upper shell forgings welds.

WCAP-16143-P, (Reference 8), documents the technical basis for the elimination of the flange requirements. These exceptions to the methods in WCAP-14040-A, Revision 4 have been reviewed and accepted by the NRC in References 9, 10, 11 and 12.

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

2. l. l The RCS temperature rate-of-change limits defined in WCAP-18370-NP, Revision O (Reference 7) are:
a.

A maximum heatup of l 00°F in any I-hour period.

b.

A maximum cooldown of 100°F in any I-hour period, and

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

c.

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

2.1.2 The RCS PIT limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2. la. The RCS PIT limits for cooldown are shown in Figure 2.2 and Table 2.1 b. These limits are in WCAP-18370-NP, Revision 0 (Reference 7) using the limiting material between Braidwood Units 1 and 2. T,his approach is conservative. Consistent with the methodology described in Reference 1, with the exception noted in Section 2.0, the RCS PIT 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 Boiler and Pressure Vessel Code Section XI, Appendix G, 1998 Edition through the 2000 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 PIT 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 PIT curve for heatup and cooldown.

2

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Material Property Basis Limiting Material: BRAIDWOOD UNIT 2 Nozzle Shell Forging SP-7056 Limiting ART Values at 57 EFPY l/4T 75°F 3/4T61°F 2500 I,

I '

IOperlimAnalysis Version:5.4 Run:4069 Operlim.xlsm I Version: 5.4.1.

2250 2000 1750 S' 1500 en CL -

! 1250

s 0 0 !

CL "0 1000 s

ftS :i u

ca 750 0

500 250 0

-250 I

i

~ Leak Test Limit I.

I I

I~

I I l H

Unacceptable I

' Operation I

,11 I

HHeatup Rate I

100°/hr

"')

!Critical Limitl 100°/hr

/

r--",

/

I Acceptable i Operation

./

A/

Criticality Limit based on inservice hydrostatic test

!lt temperature (135°F) for the jBoltup service period up to 57 EFPY iTemo. 1 i

i

-~1~

i I

I 1j Lower limit for RCS pressure is O psia I

l i

l i

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

Moderator Temperature (Deg. F)

Figure 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°F/hr)

Applicable to 57 EFPY (Without Margins for Instrumentation Errors) 3

2500 2250 2000 1750 C) 1500 en ll. -

! 1250

s II)

II) ll.

"C 1000 G) -

. ::s u ii 750 0

500 250 0

-250 BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Material Property Basis Limiting Material: BRAIDWOOD UNIT 2 Nozzle Shell Forging SP-7056 Limiting ART Values at 57 EFPY l/4T 75°F 3/4T 61 °F I

IOperlimAnalysis Version:5.4 Run:4069 Operlim.xlsm Version: 5.4.1 I

-1 u~:;~!~le --

j I

I I

I l

I I

j L

Cooldown mu

(

Rates

,t I

Acceptable I steady-state Ooeration*

25°F/hr

-50°F/hr 100°F/hr I'-="~--

I i

I I

I I

Lower limit for RCS pressure is O psia I I

I I

I 0

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

Figure 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 25, 50, and 100°F/hr) Applicable to 57 EFPY (Without Margins for Instrumentation Errors) 4

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.la Braidwood Unit 2 Heatup Data Points at 57 EFPY (Without Margins for Instrumentation Errors)

Heatup Curve 100 f Heatup Criticality Limit Leak Test Limit T (0 f)

P (osig)

T (°F)

P (osig)

T (0f)

P (osig) 60 Note 1 135 Note 1 118 2000 60 879 135 940 135 2485 65 912 140 957 70 921 145 978 75 921 150 1004 80 921 155 1035 85 923 160 1071 90 929 165 1113 95 940 170 1161 100 957 175 1215 105

-978 180 1277 110 1004 185 1345 115 1035 190 1422 120 1071 195 1508 125 1113 200 1604 130 1161 205 1710 135 1215 210 1827 140 1277 215 1957 145 1345 220 2102 150 1422 225 2261 155 1508 230 2437 160 1604 165 1710 170 1827 175 1957 180 2102 185 2261 190 2437 Note 1:

The Minimum acceptable pressure is O psia 5

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.lb Braidwood Unit 2 Cooldown Data Points at 57 EFPY (Without Margins for Instrumentation Errors)

Cooldown Curves Steady State 25 °P Cooldown 50 °P Cooldown 100 °P Cooldown T (°F)

P (psig)

T (°F)

P (osig)

T (°F)

P (osig)

T(OP)

P (osig) 60 Note l 60 Note l 60 Note l 60 Note 1 60 882 60 854 60 828 60 788 65 912 65 886 65 864 65 835 70 944 70 923 70 905 70 887 75 980 75 963 75 950 75 944 80 1020 80 1007 80 1000 80 1000 85 1063 85 1056 85 1055 85 1055 90 1112 90 1110 90 1110 90 lllO 95 1165 95 1165 95 1165 95 1165 100 1224 100 1224 100 1224 100

\\

,. 1224 105 1290 105 1290 105 1290 105 1290 110 1362 110 1362 flO 1362 110 1362 115 1442 115 1442 115 1442 115 1442 120 1530 120 1530 120 1530 120 1530 125 1627 125 1627 125 1627 125 1627 130 1735 130 1735 130 1735 130 1735 135 1854 135 1854 135 1854 135 1854 140 1986 140 1986 140 1986 140 1986 145 2131 145 2131 145 2131 145 2131 150 2292 150 2292 150 2292 150 2292 155 2469 155 2469 155 2469 155 2469

  • Note 1:

The Minimum acceptable pressure is O psia 6

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Braidwood Unit 2 low temperature overpressure protection (LTOP) system pressurizer power operated relief valve (PORV) lift.*

settings, LTOP system arming temperature, and minimum reactor vessel boltup temperature.

3; 1 LTOP System Setpoints (LCO 3.4.12).

Two PORVs shall each have nominal lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on Reference 3.

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 L TOP PORV nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.

3.2 L TOP Enable Temperature Braidwood Unit 2 procedures governing the heatup and cooldown of the RCS require the arming of the LTOP system for RCS temperature less than 350°Fand disarming of L TOP for RCS temperature of 350°F and above.

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

3.3 Reactor VesselBoltup 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 tension applied to any stud, and with the RCS vented to atmosphere.

7

~

~

ill BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2300.00 + --********

1800.00

-- PCV 456 PCV 455A Unacceptable Operation

£ 1300.00 +*------+-------j----+----+----+------+-----J"******c-;:::,::_::_::_-:._-:._t_-::._-::._--_--~---+

Ill u

a:

Acceptable Operation 800.QQ *>---------;---'-- -~-----+*****-****--* *c------+ l-------**----t--****************"

....J -

_i_ -

300.00.. >-------~---+-----+------+-----..------~>---------j 0

so 100 150 200 250 300 350 400 Auctioneered Low RCS Temperature (°F)

Figure 3.1 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 57 EFPY (Includes Instrumentation Uncertainty) 8 450

PCV-455A BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Braidwood Unit 2 Nominal PORV Setpoints for the L TOP System Applicable for 57 EFPY (Includes Instrumentation Uncertainty)

PCV-456 RCS TEMP.

RCS Pressure RCS TEMP.

RCS Pressure

{DEG. F}

{PSIG}_

{DEG. F}

{PSIG}

60 620 60 695 300 620 300 695 400 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS

, Temperatures greater than 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 L TOP system arming while at power).

9

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 4) 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 Boiler and Pressure Vessel Code,Section III, NB-2331.

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 El85-82.

The fourth reactor vessel material irradiation surveillance specimens (Capsule V) have been analyzed to determine changes in material properties (Reference 5).

The surveillance capsule testing has been completed for the original operating period. The remaining two capsules, Y and Z, were removed and placed in the spent fuel pool to avoid excessive fluence accumulation should they be needed to support life extension. The removal summary is provided in Table 4.1.

10

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 2 Capsule Withdrawal Summary<al Capsule Capsule Lead Factor Withdrawal EFPY(bl Fluence Location (n/cm2, E>l.0 MeV) u 58.5° 4.08 1.18 (EOC l) 0.387 X 1019 X

238.5° 4.03 4.24 (EOC 4A) 1.15 X 1019 w

121.5° 4.06 8.56 (EOC 7) 2.07 X 1019 z(c) 301.5° 4.15 12.78 (EOC 10) 2.83 X 1019 y(c) 241.0° 3.90 12.78 (EOC l 0) 2.66x 1019 V

61.0° 3.92 18.41 (EOC 14) 3.73 X 1019 Notes:

(a)

Source document is WCAP-18107-NP (Reference 5), Table 7-1.

(b)

Effective Full Power Years (EFPY) from plant startup.

(c)

Standby Capsules Zand Y were removed and placed in the spent fuel pool. No testing or analysis has been performed on these capsules.

11

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 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 5.1 shows the calculation of the surveillance material chemistry factors using surveillance capsule data.

Table 5.2 provides the reactor vessel material properties table.

Table 5.3 provides a summary of the Braidwood Unit 2 adjusted reference temperature (ART) values at the 1/4T and 3/4T locations for 57 EFPY.

Table 5.4 shows the calculation of ARTs at 57 EFPY for the limiting Braidwood Unit 2 reactor vessel material, i.e. Nozzle Shell Forging 5P-7056.

Table 5.5 provides the RTPTs Calculation for Braidwood Unit 2 Beltline and Extended Beltline Regions Materials at EOLE (57 EFPY), (Reference 7).

12

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data<a>

Material Capsule Capsule f1bl FF<c>

ARTNoy!b)

FF*ARTNDT (n/cm2, E > 1.0 MeV)

(OF)

(OF) u 0.387 X 1019 0.737 o.o<d) 0.00 Lower Shell X

1.15 X 1019 1.039 o.o<d>

0.00 Forging (Tangential) w

  • 2.07 X 10 19 1.198 4.6 5.51 V

3.73 X 1019 1.341 28.4 38.08 u

0.387 X 1Q19 0.737 o.o<d) 0.00 Lower Shell X

1.15 X 10 19 1.039 33.8 35.12 Forging (Axial) w 2.07 X 10 19 1.198

. 33.1 39.66 V

3.73 X 10 19 1.341 63.3 84.88 SUM:

203.25 CF LS Forging= L(FF *ARTNDT) + L(FF2) = (203.25) + (9.71) = 20.9°F u

0.388 X 1019 0.738 17.4 12.84 Braidwood Unit 1 X

l.17x 10 19 1.044 29.8 31.11 Surveillance Weld Material w

1.98 X 1019 1.186 49.0 58.14 V

3.71 X 1019 1.340 62.8 84.13 u

0.387 X 1019 0.737 o.o<d) 0.00 Braidwood Unit 2 Surveillance Weld X

1.15 X 1019 1.039 26.l 27.12 Material w

2.07 X 10 19

  • 1.198 23.7 28.39 V

3.73 X 10 19 1.341 45.6

  • 61.14

)

SUM:

302.87 CF Weld Metal= L(FF

  • ARTNDT) + L(FF2) = (302.87) + (9.69) = 31.2°F Notlls:

(a)

Source document is WCAP-18370-NP (Reference 7), Table 5-2 and Table 5-3.

(b) f= fluence; ARTNDT values are the measured 30 ft-lb shift values taken from Reference 5.

( c)

FF = fluence factor ~ f<0*28 - 0* 10*10s f)

(d)

Measured ARTNDT values were determined to be negative, but physically a reduction should not occur; therefore, conservative values of zero are used.

13 FF2 0.54 1.08 1.44 1.80 0.54 1.08 1.44 1.80 9.71 0.54 1.09 1.41

  • 1.79 0.54 1.08 1.44 1.80 9.69

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Braidwood Unit 2 Reactor Vessel Material Properties Material Description Cu(%)

Ni(%)

Chemistry Initial Factor RT NDT (°F)<a)

Closure Head Flange Heat# 3P6566/5P7547/4P6986 0.75 20 Serial # 2031-V-l Vessel Flange 0.07 0.70 20 Heat# 124P455 Inlet Nozzle O 1-001 0.07 0.83 44(b)

-10 Heat#41-5414 Inlet Nozzle O 1-002 0.07 0.85 44(b)

-10 Heat# 41-5414 Inlet Nozzle 02-001 0.09 0.88 58(b)

-10 Heat # 42-5417 Inlet Nozzle 02-002 0.09 0.89 58(b)

-10 Heat # 42-5417 Outlet Nozzle O 1-002 0.09 0.86 58(b) 10 Heat # 11-5266 Outlet Nozzle O 1-003 0.09 0.88 58(b)

-10 Heat # 11-5226 Outlet Nozzle 02-001 0.07 0.84 44(b)

-10 Heat# 4~3481 Outlet Nozzle 02-002 0.09 0.78 58(b)

-10 Heat# 4-3502 Nozzle Shell Forging

  • 0.04 0.90 26.0°F(b) 30 Heat # 5-P-7056 Intermediate Shell Forging
  • 0.03 0.71 20.0°F(b)

-30 Heat# f 490963/49C904 l-l-l Lower Shell Forging

  • 0.06 0.76 37.0°F(b)

-30 Heat# f50D102/50C971-l-l 20.9°F(c)

Circumferential Weld*

(Intermediate Shell to Lower Shell) 0.03 0.67 41.0F(b) 40 Weld Seam WF-562 3 l.2F(c)

Heat# 442011 Circumferential Weld*

(Nozzle Shell to Intermediate Shell) 0.04 0.46 54.0°F(b)

-25 Weld Seam WF-645 Heat# H4498 Inlet Nozzle to Nozzle Shell Forging**

Circumferential Weld Seams 0.18 0.52 167(d)

-48.6(e)

WF-654 (HT# 41404)

Outlet Nozzle to Nozzle Shell Forging**

Circumferential Weld Seams 0.18 0.52 167(d)

-48.6(e)

WF-654 (HT# 41404)

Notes contained on the following page.

14

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Notes: Source document is WCAP-18370-NP (Reference 7), Table 3-2 and Table 5-5

  • Beltline Region Materials
    • Extended Beltline Region Materials a) The Initial RTNoTvalues for the plates and welds are based on measured data.

b) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position 1.1.

c) Chemistry Factor calculated for Cu and Ni values per Regulatory Guide 1.99, Rev. 2, Position 2.1 d) Value is the required minimum per condition from BA W-2308 (Reference 2).

e) Generic value taken from BAW-2308 (Reference 2) 15

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Braidwood Unit 2 Adjusted Reference Temperature (ART) Values at 1/4T and 3/4T Locations for 57 EFPY<a>

Surface Fluence 57 EFPY Reactor Vessel Material (n/cm2, E>l.O MeV) 1/4T ART (°F) 3/4T ART (°F)

Inlet Nozzle 0 1-001 1.06 X 1017 0.0(b)

Inlet Nozzle 0 1-002 1.06 X 1017 0.0(b)

Inlet Nozzle 02-001 1.06 X 1017 3.2(b)

Inlet Nozzle 02-002 1.06 X 1017 3.2(b)

Outlet Nozzle 0 1-002 7.97x 1016 20.9(b)

Outlet Nozzle 0 1-003 7.97 X 1016 0.9(b)

Outlet Nozzle 02-001 7.97 X 1016

-l.7(b)

Outlet Nozzle 02-002 7.97 X 1016 0.9(b)

Nozzle Shell Forging 0.994 X 1019 74.5 60.5 Intermediate Shell Forging 2.95x 1019 16.3 5.0 Lower Shell Forging 3.03 X 1019 47.1 35.3

-Using non-credible surveillance data 3.03 X 1019 18.7 6.9 Nozzle to Intermediate Shell Forging Circ. Weld Seam 0.994x 1019 67.4 38.4 (Heat # H4498)

Intermediate to Lower Shell Forging Circ. Weld Seam 2.90 X 1019 134.5 111.3 (Heat# 442011)

-Using credible surveillance data 2.90 X 1019 104.0 94.3 Inlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams 1.06 X 10 17 37.0(b)

WF-654 (HT# 41404)

Outlet Nozzle to Nozzle Shell Forging Circumferential Weld Seams 7.97 X 1016 33.6(b)(c)

WF-654 (HT# 41404)

Notes:

(a)

The source document containing detailed calculations is WCAP-18370-NP (Reference 7), Table 7-2, Table 7-6, Table 7-7, and Table 7-8.

(b)

The ART value.s for the extended beltline materials are conservatively calculated at the surface, i.e.,

without attenuation of the fluence.

( c)

The outlet nozzle materials do not exceed the l x 10 17 n/cm2 fluence threshold at 57 EFPY; therefore, neutron irradiation embrittlement need not be considered for the nozzle materials.

However, the results are included for information.

16

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTs) at 57 EFPY at the Limiting Reactor Vessel Material, Nozzle

. Shell Forging 5P-7056 Parameter Values Operating Time 57 EFPY Location<a) l/4T ART (°F) 3/4T ART(°F)

Chemistrv Factor, CF (°F) 26.0 26.0 Fluence(t), n/cm2 5.97x1018

  • 2.15x1018 (E> 1.0 Mev)<h)

Fluence Factor, FF 0.856 0.587

~TNoT=:. CFxFF(°F) 22.2 15.3 Initial RT NDT,, l(°F) 30 30 Margin, M(°F) 22.2 15.3 ART= I+(CF*FF)+M, °F 74.5 60.5 per RG I.99, Revision 2 Notes: Source document is WCAP-18370-NP (Reference 7), Table 7-6, Table 7-7 and Table 7-9 a) The Braidwood Unit 2 reactor vessel wall thickness is 8.5 inches at the beltline region.

b) Fluence, f, is the calculated peak clad/base metal interface fluence (E>l.O Mev) = 9.94xl0 18 n/cm2 at 57 EFPY (Reference 7).

17

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTPTS Calculation for Braidwood Unit 2 Beltline and Extended Beltline Regions Materials at EOLE (57 EFPY)<a,hJ R.G.

Fluence 1.99, CF I

IRTNo'r(c)

ARTNDT

<lu(c)

<J,1(d)

Margin RTPTs Reactor Vessel Material Rev.2 (OF)

(n/cm2, E>l.0 FF

(°F)

(OF)

(°F)

(°F)

(°F)

(OF)

Position MeV)

Reactor Vessel Extended Beltline Materials Inlet Nozzle O l -00 I 1.1 44.0 1.06 X 1017 0.114

-10 5.0 0

2.5 5.0 0.0 Inlet Nozzle O 1-002 1.1 44.0 1.06 X 1017 i

0.114

-10 5.0 0

2.5 5.0 0.0 Inlet Nozzle 02-001 1.1 58.0 1.06 X 1017 0.114

-10 6.6 0

3.3 6.6 3.2 I

Inlet Nozzle 02-002 1.1 58.0 1.06 X 1017 0.114

-10 6.6 0

3.3 6.6 3.2 Outlet Nozzle 01-002 1.1 58.0 7.97 X 1016 0.094 IO 5.4(e) 0 2.7 5.4 20.9 Outlet Nozzle 01-003 1.1 58.0 7.97 X 1016 0.094

-10 5.4(e) 0 2.7 5.4 0.9 Outlet Nozzle 02-001 1.1 44.0 7.97 X 1016 0.094

-10 4.I<e>

0 2.1 4.l

-1.7 Outlet Nozzle 02-002 I.I 58.0 7.97 X 1016 0.094

-10 5.4(e) 0 2.7 5.4 0.9 Inlet Nozzle to Nozzle Shell Forging 1.1 167.0(t) 1.06 X 1017 0.114

-48.6(t) 19.0 I 8<t) 28.0(t) 66.6 37.0

  • Circumferential Weld Seams Outlet Nozzle to Nozzle Shell Forging 1.1 167.0(t) 7.97 X 1016 0.094

-48.6(t)

J5_7(e)

I8<t) 28.0(t) 66.6 33.6 Circumferential Weld Seams 18

BRAIDWOOD - UNIT 2.

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 (continued)

Reactor Vessel Beltline Materials Nozzle Shell Forging 1.1 26 0.994 X 1019 0.998

  • 30 Intermediate Shell Forging 1.1 20 2.95 X 1019 1.287

-30 Lower Shell Forging 1.1 37 3.03 X 1019 1.293

. -30

-using non-credible surveillance 2.1 20.9 3.03 X 1019 1.293

  • -30 data Nozzle to Intermediate Shell Forging Circ. Weld Seam 1.1 54 0.994x 1019 0.998

-25 (Heat # H4498)

Intermediate to Lower Shell Forging Circ. Weld Seam 1.1 41 2.90 X 1019 1.283 40 (Heat# 442011)

-using credible surveillance data 2.1 31.2 2.90 X 1019 1.283 40 Notes:

(a) The 10 CFR 50.61 methodology was utilized in the calculation of the RTrTs values.

(b) The source document containing detailed calculations is WCAP-18370-NP (Reference 7), Table E-2.

(c) Initial RTNoT values are based on measured data. Hence, cru = 0°F..

26.0 0

13.0 25.7 0

12.9 47.8 0

17 27.0 0

13.5 53.9 0

27.0 52.6 0

26.3 40.0 0

14 (d) Per the guidance of 10 CFR 50.61, the base metal cr,-. = I 7°F for Position 1.1 (without surveillance data) and for Position 2.1 with non-credible surveillance data; the weld metal cr,-. = 28°F for Position 1.1 (without surveillance data) and with credible surveillance data cr,-. = 14°F for Position 2.1. However, cr,-. need not exceed 0.5*~RTNoT.

( e) The outlet nozzle materials do not exceed the I x 10 17 n/cm2 fluence threshold at 57 EFPY; therefore, neutron irradiation embrittlement need not be considered for the nozzle materials. However, the results are included for information.

(f) The IRTNoT values, are based on BA W-2308 (Reference 2). Use of BA W-2308 as an exemption to the 10 CFR 50.61 methodology was approved in Reference 11. BA W-2308 requires the use of an cr1 = l 8°F, cr,-. = 28°F, and a minimum CF of 167°F.

19 26.0 81.9 25.7 21.5 34 51.8 27.0 24.1 53.9 82.8 52.6 145.2 28 108

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 6.0 References

l.

WCAP-14040-A, Revision 4, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves", J.D.

Andrachek, et al., May 2004.

2.

AREY A Document, BA W-2308, Revision 1-A and 2-A, "Initial RT NOT of Linde 80 Weld Materials," August 2005 and March 2008.

3.

L TR-SCS-19-14, Revision 1, "Braidwood Units I and 2 Low Temperature Overpressure Protection System (L TOPS) Analysis for 57 EFPY," September 19, 2019.

4.

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

5.

WCAP-18107-NP, Revision 0, "Analysis of Capsule V from the Exelon Generation Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program,"

May 2016.

6.

Not Used.

7.

WCAP-18370.:NP, Revision 0, "Braidwood Units 1 and 2 Heatup and Cooldown Limits for Normal Operation," June 2019.

8.

WCAP~16143-P, Revision 1, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," W. Bamford, et al.,

October 2014.

9.

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 -

Exemption from the Requirements of 10 CFR Part 50, Appendix G (TAC Nos.

MC8697, MC8698, MC8699, and MC8700)," November 22, 2006. [ADAMS Accession Number ML061890003]

10. NRC Letter from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 -

Issuance of Amendments to Utilize WCAP-16143-P, Revision 1 "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," Dated October 16, 2014 (CAC Nos. MF5033, MF5034, MF5035 and MF5036)," October 28, 2015. [ADAMS Accession Number ML15232A441]

11. NRC Letter from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2-Exemption from the Requirements of 1 0CFR50.61 and 1 0FR50, Appendix G (EPID L-2019-LLE-0022)," August 31, 2020. [ADAMS Accession Number ML20022A336]

20

BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT

12. Issuance of Amendment Nos. 217, 217, 221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-20 l 9-LLA-0215)," September 18, 2020. [ADAMS Accession Number ML20163A046].

NRC Letter from J. S. Wiebe, NRR, to B.C. Hanson, Exelon Generation Company, LLC, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2-Issuance of Amendment Nos. 217, 217, 221, and 221 Regarding Reactor Coolant System Pressure and Temperature Limits Report Technical Specifications (EPID L-20 l 9-LLA-0215),"

September 18, 2020. [ADAMS Accession Number ML20163A046].

21

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