ML14059A124
| ML14059A124 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 02/28/2014 |
| From: | Kanavos M Exelon Generation Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW140001, TAC MF2418, TAC MF2419, TAC MF2420, TAC MF2421 | |
| Download: ML14059A124 (74) | |
Text
February 28, 2014 BW140001 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Units 1 and 2 Facility Operating License Nos. NPF-72 and NPF-77 NRC Docket Nos. STN 50-456 and 50-457
Subject:
Pressure and Temperature Limits Reports (PTLRs), Braidwood Station, Units 1 and 2
References:
(1) Letter from Craig Lambert (Exelon Generation Company, LLC) to u. S.
NRC, "Request for License amendment Regarding Measurement Uncertainty Recapture (MUR) Power Uprate," dated June 23, 2011 (2) Letter from J. S. Wiebe (U. S. NRC) to M. J. Pacilio, "Braidwood Station, Units 1 and 2, and Byron Station, Unit Nos. 1 and 2 -Issuance of Amendments Regarding Measurement Uncertainty Recapture Power Uprate (TAC Nos. MF2418, MF2419, MF2420, and MF2421)," dated February 7, 2014 The purpose of this letter is to transmit the Pressure Temperature Limits Reports (PTLRs) for Braidwood Station, Units 1 and 2 in accordance with Technical Specification (TS) 5.6.6, "Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR}." The Braidwood Unit 1 and Unit 2 PTLRs were recently revised to incorporate analysis to support the Measurement Uncertainty Recapture (MUR) conditions. The methodology for developing the revised PTLRs is consistent with a recently approved License Amendment Request for Braidwood Station (Reference 2). In addition, Unit 1 PTLR, Revision 6 is attached which corrects a typographical error made under Revision 5.
Please ""'uN",,.rl!.n,.... this matter to Raush, Sincerely, Mark Kanavos
February 28, 2014 U. S. Nuclear Regulatory Commission Page 2 Attachments: 1. Braidwood Unit 1 Pressure Temperature Limits Report (PTLR), Revision 5
- 2. Braidwood Unit 1 Pressure Temperature Limits Report (PTLR), Revision 6
- 3. Braidwood Unit 2 Pressure Temperature Limits Report (PTLR), Revision 5
ATTACHMENT 1 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 5
BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
Revision 5
BRAID"VOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 2.0 ReS Pressure and Temperature Limits 2.1 ReS Pressure and Temperature (NT) Limits (LeO 3.4.3) 3.0 Low Temperature Over Pressure Protection and Boltup 7 3.1 LTOP System Setpoints (LeO 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 18
BRAIDWOOD - DNIT 1 PRESSURE AND TEMPERATlJRE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit I Reactor Coolant System Heatup Limitations (Beatup Rate of IOO°F/hr) Applicable for EFPY (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit I 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 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page
- 2. I a Braidwood Unit 1 Heatup Data Points at 32 EFPY (Without 5 Margins for Instrumentation Errors) 2.1 b Braidwood Unit I Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Errors) 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for 9 the L TOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit 1 Surveillance Capsule Withdrawal Summary JI 5.1 Braidwood Unit I Calculation of Chemistry Factors Using Surveillance Capsule Data 5.2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature 15 (ART) Values at lJ4T and 3/4T Locations 32 EFPY 5.4 Braid\vood ] Calculation Adjusted Reference 16 Temperatures (ARTs) at 32 EFPY at Limiting Reactor Vessel 55 Unit) I7
BRAID\VOOD - UNIT 1 PRESSURE AND TEl\;lPERA TURE LIMITS REPORT
] .0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit I has been prepared in accordance with the requircments of Braidwood Technical Specification (TS) 5.6.6, I!Reactor Coolant Sysrem (RCS) Pressure and Temperature Limits Report (PTLR)".
Revisions to the PTLR shall be provided to the NRC after issuance.
The Technical Specifications (TS) addressed in this report are listed below:
LCO 3.4.3 RCS Pressure and Temperature (Pff) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 ReS Pressure and Temperature Limits The PTLR limits for Braidwood Onit 1 were developed using a methodology specified in the Technical 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 ASME Code Case N-588. "Alternative to Reference flaw Orientation of Appendix G Circumferential Welds in Reactor Vessel,Section XI, Division 1",
and d) Elimination the requirements documented in WCAP-16J 2
2.1.1 The RCS temperature rate-of-change limits defined in WCAP-15364, Revision 2 (Reference ] 1) are:
- a. A maximum heatup of 100°F any I-hour period,
BRAIDWOOD - UNIT 1 PRESSlJRE AND TEMPERATURE LIMITS REPORT
- c. A maximum temperature change of less than or equal to 10°F in any I-hour period during inscrvicc hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1.2 RCS Ptf limits hcatup, inservice hydrostatic leak testing, and criticality are specified by Figure 2.1 and Table 2.1 a. The RCS Ptf limits for cool down are shown in Figure 2.2 and Table 2.1 b. These limits are defined in WC AP-15364, Revision 2 (Reference 11). Consistent with the methodology described in Reference I and exceptions noted in Section 2.0, the RCS PIT limits for heatup and cool down 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, Article G2000, 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 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.
BRAIDWOOD* UNIT 1 PRESSURE AND TEl\1PERA TURE LIMITS REPORT 2500
!Leak Test Limit I 2250 Acceptable 2000 Operation 1750 8' Critical Limit
~
1500 1...-_--,100 Deg. F/Hr
~
- l III
£ 1250
~
'3 1000 u
d 750 Criticality Limit based on
_-----iinservice hydrostatic test temperature (108 F) for the 500 service period up to 32 EFPY 250 o 50 100 150 200 250 300 350 400 450 500 550 Moderator Figure Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rate of 100°Flhr)
Applicable f'Or 32 EFPY (Without :Margins for Instrumentation Errors)
BRAID\VOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 Ac(~ptabJe 2000 Operation 1750 G'
~
1500 e
- J l 1250
~
a 1000 d
750 500 250 o
o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg.
Figure 2.2 Braidwood Unit 1 Reactor Coolant System CooIdown Limitations (Cooldown Rates of 0, 25, 50 and 100°F/hr) Applicable for 32 EFPY (\Vithout Margins for Instrumentation Errors)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a Braidwood Unit 1 Heatup Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Heatup Curve 100 F Heatup Criticality Leak Test Limit Limit T (OF) P (psi g) T (,)F) P (psig) TeF) P(Dsig) 60 0 108 0 91 2000 60 1064 108 1114 108 2485 65 1114 110 1166 70 1166 115 1172 75 1172 120 1176 80 1176 125 1188 85 !I 88 130 1207 90 1207 135 1234 95 1234 140 1267 100 1267 145 1308 105 1308 150 1357 110 1357 155 1414 115 1414 160 1479 120 1479 165 1554 125 1554 170 1638 130 1638 175 1732 135 1732 180 1838 140 1838 185 1956 145 1956 190 2088 150 2088 195 2235 155 2235 200 2397 160 2397
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.tb Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Cooldown Curves Steady State 25 OF Cool down 50 of Cooldown 100 of CooldoWll T (OF P (psi g) T ["F) P T (OF) P (psi g) T (oF P (psig) 60 0 60 0 60 0 60 0 60 1082 60 1078 60 1078* 60 1078*
65 ILB 65 1133 65 1133 65 1133 70 JI88 70 1188 70 1188 70 ! 188 75 1250 75 1250 75 1250 75 1250 80 1318 80 1318 80 1318 80 1318 85 1393 85 1393 85 1393 85 1393 90 1476 90 1476 90 1476 90 1476 95 1568 95 1568 95 1568 95 1568 100 1669 100 1669 100 1669 100 1669 105 1781 105 1781 105 1781 105 1781 110 1905 110 1905 110 1.905 110 1905 J 15 2042 J 15 2042 115 2042 115 2042 120 2194 120 2194 120 2194 120 2194 125 2361 125 2361 125 2361 125 2361
- Refer to Reference 13
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This provides the Braidwood Unit I power operated relief valve lift settings. low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor boltup temperature.
1.1 LTOP System Setpoints (LCO 1.4.12)
The power operated relief valves (PORVs) shall each have maximum lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 4.
The LTOP setpoints are based on Pff 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 nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.
3.2 LTOP Enable Temperature Braidwood Unit J 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 3.1 extends to where the pressure setpoint is 2335 psig. This is intended to prohibit PORV ]ift 'II-'r""", LTOP system arming at power.
Reactor Bohup Temperature (Non-Technical Specification)
- Boltup the Reactor applied to the ReS
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT
~
-; 1500 S
~ ,
Unacceptable Operation
- a. 1250 a:
oa.
- 1000 c
'j§ PCV-456 o
z 750 595 psig 1 500 ============~==~
541 psig o 50 100 150 200 250 300 350 400 450 Auctioneered Low ReS Temperature (DEG. F)
Figure 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (L TOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
PCV-455A PCV-456
( ITY-0413M) (lTY -04I3P)
AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE ReS TEMP. RCS TEMP. (DEG.
60 541 60 595 300 541 300 595 400 2335 400 2335 Note: To detennine 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 LTOP system arming while at power.)
BRAIDWOOD ~ UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillauce Program The pressure material program {Referene.e 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, RT NDT, which is determined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-2331. The empirical relationship between RTNtrr 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 finaJ reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine ehanges in material properties. The surveillance capsule testing has been completed for the original operating period. The remaining three capsules, V, 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.
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERA TURE LIMITS REPORT Table 4.1
.Braidwood Unit 1 Capsule Withdrawal Summary(a)
Capsule Withdrawal EFPy(b) Fluence Capsule Lead Factor Location (nlcm 2, E>1.0 MeV) lJ 58.5° 4.02 1.16 0.388 x IDI9 19 I
X 238.5° 4.06 4.30 1.17 x 10 I
W 121.5° 4.05 7.79 1.98 x 10 19 1
Z(C) 301.5° 4.09 12.01 (EOC 10) 2.79 x 10 19 I
v(c) 61.0° 3.92 17.69 (EOC 14) 3.71 x 10 19 I
2.60 x 10 19 y(c) 241.0° 3.81 12.01 (EOC 10)
I Notes:
Source document is CN-AMLRS-IO-7 (Reference 14), Table 5.7-3.
Effective Full Power Years (EFPY) from Standby Capsules Z. V. and Y were removed and placed in the fuel pool. No testing or analysis has been on these If license renewal is sought, one of these standby may need to be tesled to determine tbe effect of neutron irradiation on Ihe reactor vessel surveillance materials the of extended "'.......<>h'
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERA TllRE l,IMITS 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 PIT 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 1 adjusted reference temperature (ART) values at the 1I4T and 3/4T locations for 32 EFPY.
Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit 1 reactor vessel material, i.e. weld WF-562 ( HT # 442011, Based on Surveillance Capsules U and X Data).
Table 5.5 provides the RT pTs calculation for Braidwood Unit 1 BeHline Region Materials at EOL (32 EFPY), (Reference 7),
BRAID';VOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood IJnit J Calculation of Chemistry Factors Using Surveillance Capsule Data(a)
Material Capsule Capsule rhl (nJcm 2, E> 1.0 MeV)
(bi FF*ART"!JT (oF)
FF2 ,
19 U 0.388 x 10 0.738 5.78 4.26 0.54 Lower Shell Forging X 1.17 x (OJ9 1.044 38.23 39.91 L09 Cl'angentiall 19 W 1.98 x 10 Ll86 24.14 28.64 1.41 U 0.388 x 10 19 0.738 O.Oldl 0.00 0.54 LowerShelJ 19 Forging X 1.17 x 10 1.044 28.75 30.01 1.09 (Axial) 19 W 1.98 x 10 1.186 37.11 44.03 1.41 SUM: 146.85 6.08 I
CF LS F[Jrgin~ LlFF *LlRT NDI) +
19 I(FF 2
) (146.85) + (6.08) =24.1°.F I U 0.388 X 10 0.738 17.06 12.59 0.54 Braidwood Unit )
19 Surveillance Weld X Ll7 x 10 1.()44 30.15 31.47 L09 Material W 1.98 x 1.186 49.68 58.94 1041 19 U 0.388 X [0 0.738 0.00 0.54 Braidwood Unit 2 Surveillance Weld X I 15 x I.tH9 26.3 27.:n 1.08 Material W 2.07 Ll98 23.9 28.63 .44 SUM: 158.96 6.10 I
CF Wt!ll Melal =:L(FF . "m) L<FF2) = ) + (6. =26.]"F I Notes:
r.
Source document CN-AMLRS-10-7 U'::H::m.e Table 5.2-1.
f =: tluc!1('e; the tnf'::l<;!ln"¥1 30 shift values taken ReIerence
=
- H.""' ell "'H~U be lCUUCllO!1 L'UU;-'CI watf
BRAIDWOOD - UNIT 1 PRESSlJRE AND TEMPERATURE LIMITS REPORT Table 5.2 Braidwood Unit 1 Reactor Vessel Material Properties Chemistry Initial Material Description Cu ) Ni RT NIH,(oF)(a)
Factor .
Closure Head Flange 0.11 0.67 -- -20 Heat # 5P738113P6406 Vessel Flange Heat # 122N357V
-- 0.77 -- -10 Nozzle Shell Forging
- 26.0°F(h) 0.04 0.73 10 Hear # 5P-7016 Intermediate Shell Forging
- 31.0oph) 0.05 0.73 -30 Heat # [49D383/49C344]- I 1 Lower Shell Forging
- 31.0oph) 0.05 0.74 24. Jope) -20 Heat # [49D867/49C813 J-l-l Circumferential Weld
- 41.0°F(h)
(Intermediate Shell to Lower Shell) 0.03 0.67 26.1ope) 40 WF-562 (HT# 44201 I)
Upper Circumferential Weld
- Shell to lntennediate OJ)4 0.46 54.0o ph) -25 WF-645 (HT# H4498)
Belfline Materials a) The Initial values for the and welds are based on measured data.
racfor calculated for eu and Ni values per Guide I Rev. 2, Position I I.
faclOr calculated for eu and Ni values per Guide 1.99. Rev. 2. Posifion 2. L
BRAID\VOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature (ART) Values at 1I4T and 3/4T Locations for 32 EFPy(al 32EFPY Surface :Fluence Reactor Vessel Material 1/4T ART 3/4T ART (nlcm2, E>1.0 MeV)
(OF) (OF)
Nozzle Shell Forging 0.586 x Wit) 47 34 Intermediate Shell Forging 1.76 x Wit) 33 15 Lower Shell Forging 1.76 x WIt) 43 25 19
--Using credible surveillance data 1.76 x 10 2] 15 Nozzle to Intermediate Shell Forging 19 Circ. Weld Seam 0.586x 10 52 25 (Heat # H4498)
Intermediate to Lower Shell Forging 19 eirc. Weld Seam 1 x 10 122 99 (Heat # I)
,19 iSIng 1.70 X 93 78 Notes:
la) The source document detailed calculations is CN-AMLRS-IO-7 Tables 53.1-1 and 5.3. -2. The ART values summarized in this table utilize the most recent iluence OW'leCf1011s and materials but were not used in de\'eJo'DJIltenl of the prr limit curves. See 2.1 and 2.2 of this PTLR for the ART values used in of I.he prr limit curves.
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 Braidwood Unit 1 Calculation of Adjusted Reference Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P*7016 Parameter Values Operating Time 32 EFPY Location(3) 114T ART(OF) 3/4T ART(OF)
Chemistry Factor, CF (OF) 26.0 26.0 FJuence(t), nlcm 2 3.65 x 10 Ix 1.32 x 101x (E> 1.0 Mev)(b)
Fluence Factor, FF 0.772 0.475 ARTNDT= CFxFF(OF) 18.8 12.4 Initial RT NDT., I(OF) 10 10 Margin, M eF) 18.8 12.4 ART= J+(CF*FF)+M,oF 48 35 II per RG 1.99, Revision 2 The Braidwood Unit I reactor vessel wall thickness is 8.5 inches at the heltline (b) Fluenee L is based upon > 1.0 = 6.08 x I at 32 EFPY
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 RTpTS Calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPy)(3,h)
R.G,1.99, (el G (el (dl CF Fluence IRTN1)'l' ARTNIH \I G.\ Margin RT pTS Reactor Vessel Matt~rial Rev. 2 FF
(°F') (nkm2, E>1.0 MeV) (°F) (OF) (°F) (OF) (OF) (OF)
Position I*
Nozzle Shell llrging LI 26 0.586 x IOI~ 0.8504 10 22.1 0 11.1 22.1 54
..~ ....
19 Intermediate Shell rgmg 1.1 .~ 1 1.76 X 10 1.1554 -30 35.8 () 17 34 40 19 Lower Shell orging LI 31 176 x 10 1.1554 -20 35.8 0 17 34 50
~ .. ...
~
19 ill1g eredihle I 24.1 1.76 x 10 LJ554 -20 27.8 0 8.5 17 25 Nozzle to Intermediate 19 Cire. Weld I I 54 O.586x 10 0.8504 -25 45.9 0 23.0 45.9 67 (Heat #
Intermediate to 19 Cire. Weld 1.1 41 1.70 x 10 I 1461 40 47.0 0 23.5 47.0 134 (Heat # 4420 II) f***..****
19 Ig credible data I 26.1 1.70 x 10 1.1461 40 29.9 0 14 28 98 The 10 CFR was utilized in the calculation of I.he RT vrs values.
detallt:d calculations IS CN-AMLRS .. lO-7 (Reference 14). Tahle 5.5-1.
on measured daHL Hence, O'u ::: OaF.
o CFH 5(L61. thc hase mctal 0',\ ITOP for Position 1.1 (without surveillance data) and with credihle surveillance data
- XS'F for I: t.he wdd metal crt, 28°P for Position 1.1 (without surveillance data) and with credihle surveillance data a J\
110weVCL all need not exceed 17
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE IJMITS REPORT 6.0 References J. WCAP-14040-NP-A, Revision "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," J.D. Andrachek.
e1 aL January 1996.
- 2. NRC Letter from R. A Capra to 0.0. Kingsley, Commonwealth Edison Company, "Byron Station Units 1 and 2 and Braidwood Station Units I and 2, Acceptance for refelTing of pressure temperature limits report. (M98799, M98800, M9880 I, and M98802)," January 21 1998.
- 3. Westinghouse Letter to Exelon Nuclear, CAE-I0-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report," M.P. Rudakewiz, September 8,2010.
- 4. Byron & Braidwood Design Information Transmittal DIT-BRW-2006-0051, "Transmittal of Braidwood Unit I and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System (LTOPS) Power Operated Relief Valves (PORVS)," Nathan (Joe)
Wolff Jr., July 18,2006.
- 5. WCAP-9807, "Commonwealth Edison Company, Braidwood Station Unit I Reactor Vessel Radiation Surveillance Program," S.E. Yanichko, et aI., February] 981.
WCAP-] 5316, Revision I, "Analysis Capsule W from CommonweaJth Company Braidwood 1 Vesse] Radiation etaL
,,,,,,(>"",1'"',,,,,, 1 WCAP-l I."J.H.
"Byron Station, Unit Nos. I and 2, and Braidwood Station, Unit Nos. 1 and 2 - Issuance of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC Nos. MC8693, MC8694, MC8695, and MC8696)," 2006.
BRAIDWOOD - UNIT 1 PRESSURE AND TEl\fPERATURE 1,IMITS REPORT II. WCAP-15364, Revision "Braidwood Unit I Heatup and Cool down Limit Curves for Operation," TJ. Laubham, November 2003.
J 2. WCAP- i 6143-P, Revision 0, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units 1 and 2," W. Bamford, et at, November 200:1 j 3. Westinghouse Letter to Exelon Nuclear, CCE-07-24, "Braidwood Unit J and 2 RCS HU/CD Limit Curve Table Values," dated February 15,2007.
Westinghouse Calculation Note CN-AMLRS-I 0-7, Revision 0, "Braidwood Units 1 and 2 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations."
A.E. Leicht, September 2010.
ATTACHMENT 2 Braidwood Unit 1 Pressure and Temperature Limits Report (PTLR), Revision 6
BRAIDWOOD UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
Revision 6
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Seetion Page 1.0 Introduction 1 2.0 ReS Pressure and Temperature Limits 1 2.1 ReS Pressure and Temperature (PIT) Limits (LeO 3.4.3) 1 3.0 Low Temperature Over Pressure Protection and Boltup 7 3.1 LTOP System Setpoints (LeO 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 18
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit 1 Reactor Coolant System Heatup Limitations (Heatup 3 Rate of 100°Flhr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0,25,50, and 100°Flhr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (hlcludes Instrumentation Uncertainty)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1a Braidwood Unit 1 Heatup Data Points at 32 EFPY (Without 5 Margins for Instrumentation Errors) 2.1b Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without 6 Margins for Instrumentation Errors) 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for 9 the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty) 4.1 Braidwood Unit 1 Surveillance Capsule Withdrawal Summary 11 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using 13 Surveillance Capsule Data 5.2 Braidwood Unit 1 Reactor Vessel Material Properties 14 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature 15 (ART) Values at 1I4T and 3/4T Locations for 32 EFPY 5.4 Braidwood Unit 1 Calculation of Adjusted Reference 16 Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P-7016 RTprs Calculation for Braidwood Unit 1 Beltline Region 17 at EOL (32 EFPY)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 1 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 (TS) addressed in this report are listed below:
LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 ReS Pressure and Temperature Limits The PTLR limits for Braidwood Unit 1 were developed using a methodology specified in the Technical Specifications. The methodology listed in WCAP-I4040-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 I",
and d) Elimination of the flange requirements documented in WCAP-16143-P.
These exceptions to the methodology in "VCAP I4040-NP-A, Revision 2 have been reviewed and accepted by the NRC in References 2, 8, 9 and lO.
WCAP 15364, Revision 2 (Reference 11), provides the basis for the Braidwood Unit 1 PIT curves, along with best estimate chemical compositions, fluence projections and to these uu,,~,,>.
2.1.1 The RCS temperature rate-of-change limits defined in WCAP-I5364, Revision 2 (Reference 11) are:
- a. A maximum heatup of lOO°F in any I-hour period,
BRAIDWOOD* UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT
- c. A maximum temperature change of less than or equal to 10°F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
2.1 The RCS PIT limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table lao The RCS PIT limit" for cooldown are shown in Figure 2.2 and Table 2.1 b. These limits are defined in WCAP-15364, Revision 2 (Reference 11). Consistent with the methodology described in Reference 1 and exceptions 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 instmment enOL These limits were developed using ASME Boiler and Pressure Vessel Code Section XI, Appendix G, Article G2000, 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 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 pennissible temperature in the conesponding PIT curve for heatup and cooldown.
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2250 2000 1750 8'
--f£
~
- J 1500 i
l 1250 a10 1000 a
750 500 250 o +-~~~~++~~~~+-~~~~+r~~~~+r~~~~~~M o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)
Figure 2.1 Braidwood 1 Reactor Coolant System Heatup Limitations (Heatup Rate of lOO°Flhr)
Applicable for 32 EFPY (Without Margins for Instrumentation Errors)
BRAIDWOOD* UNIT 1 PRESSURE AND TEl\1PERATURE LIMITS REPORT 2500 2250 2000 1750 6
--~ 1500 I
0..
j 1250
~(J 1000 a steady-state,
-25,
-50, and 750 -100 500 250 o~~~~~~~~~~~~~~~~~~~~~~~~~~
o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)
Figure 2.2 Braidwood Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0,25,50 and 100°Flhr) Applicable for 32 EFPY (Without Margins for Instrumentation Errors) 4
BRAIDWOOD* UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a Braidwood Unit 1 Heatup Data Points at 32 EFPY (Witbout Margins for Instrumentation Errors)
Heatup Curve 100F Heatup Criticality Leak Test Limit Limit T(°F) P(psig) TeF) P (psig) T(°F) P (Pliig) 60 0 108 0 91 2000 60 1064 108 1114 108 2485 65 1114 110 1166 70 1166 115 1172 75 1172 120 1176 80 1176 125 1188 85 1188 130 1207 90 1207 135 1234 95 1234 140 1267 100 1267 145 1308 105 1308 150 1357 110 1357 155 1414 115 1414 160 1479 120 1479 165 1554 125 1554 170 1638 130 1638 175 1732 135 1732 180 1838 140 1838 185 1956 145 1956 190 2088 150 2088 195 2235 155 2235 200 2397 160 2397
BRAIDWOOD - UNIT 1 PRESSURE AND TKMPERATURE LIMITS REPORT Table 2.1b Braidwood Unit 1 Cooldown Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Cooldown Curves Steady State 25 of Cooldown 50 OF Cooldown 100 of Cooldown TCF) P (psig) TCF) P (psig) T (OF) P (psi g) T (OF) P (psig) 60 0 60 0 60 0 60 0 60 1082 60 1078 60 1078* 60 1078*
65 1133 65 I133 65 1133 65 1133 70 1188 70 1188 70 1188 70 1188 75 1250 75 1250 75 1250 75 1250 80 1318 80 1318 80 1318 80 1318 85 1393 85 1393 85 1393 85 1393 90 1476 90 1476 90 1476 90 1476 95 1568 95 1568 95 1568 95 1568 100 1669 100 1669 100 1669 100 1669 105 1781 105 1781 105 1781 105 1781 110 1905 110 1905 110 1905 110 1905 115 2042 115 2042 115 2042 115 2042 120 2194 120 2194 120 2194 120 2194 125 2361 125 2361 125 2361 125 2361
- Refer to Reference 13
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Braidwood Unit 1 power operated relief valve lift settings, low temperature overpressure protection (LTOP) system arming temperature, and minimum 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 in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 4.
The LTOP setpoints are based on PIT 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 nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.
3.2 LTOP Enable Temperature Braidwood Unit 1 procedures governing the heatup and cool down 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 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 Vessel Boltup Temperature (Non-Technical Specification)
The minimum boltup temperature for the Reactor Vesse} 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.
BRAIDWOOD* UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 2335 psig 2250 2000 1750 (j
iii D.
E1500
- J Unacceptable Operation E
D. 1250 a:
0 D.
iii 1000 c:
'E PCV-456 0
z 750 500 541 psig PCV-455A 250 0
0 50 100 150 200 250 300 350 400 450 Auctioneered low ReS Temperature (OEG. F)
Figure 3.1 Braidwood Unit 1 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Braidwood Unit 1 Nominal PORV Setpoints for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
PCV-455A PCV-456 (1TY-0413P)
AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG. RCS TEMP. (DEG. F) (PSIG) 60 541 60 595 300 541 300 595 400 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 LTOP system arming while at power.)
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference 5) 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, RT NDT, which is detennined in accordance with ASME Boiler and Pressure Vessel Code Section III, NB-233L 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 EI85-82.
The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in material properties. The surveillance capsule testing has been completed for the original operating period. The remaining three capsules, V, 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.
BRAIDWOOD* UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 1 Capsule \Vithdrawal Summary(a)
Capsule Withdrawal EFPy(b) Fluence Capsule Lead Factor Location (n/cm2, E>1.0 MeV)
U 58.5° 4.02 1.16 0.388 x 10 19 X 238.5° 4.06 4.30 1.17 x 10 19 W 121.5° 4.05 7.79 1.98 x 10 19 Z(c) 301.5° 4.09 12.01 (EOe 10) 2.79 x 10\9 Vee) 61.0° 3.92 17.69 (EOe 14) 3.71 x 10 19 y(c) 241.0° 3.81 12.01 (EOe 10) 2.60 x 10\9 Notes:
(a) Source document is CN-AMLRS-IO-7 (Reference 14), 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 these standby capsules may need to be tested to determine the effect of neutron irradiation on the reactor vessel surveillance materials during the period of extended operation.
BRAIDWOOD - UNIT 1 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 Pff 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 1 adjusted reference temperature (ART) values at the 1I4T and 3/4T locations for 32 EFPY.
Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit 1 reactor vessel material, i.e. weld WF-562 ( HT # 442011, Based on Surveillance Capsules U and X Data).
Table 5.5 provides the RTPTs calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPY), (Reference 7).
BRAIDWOOD ~ UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood Unit 1 Calculation of Chemistry Factors Using Surveillance Capsule Data(a)
(b)
Capsule fb) FF(c) ARTNDT FF*ART NDT Material Capsule FF2 (nlcm2, E > 1.0 MeV) (OF) (OF)
U 0.388 x 10 19 0.738 5.78 4.26 0.54 Lower Shell Forging X 1.17x 10 19 1.044 38.23 39.91 1.09 (Tangential)
W 1.98 x 10 19 1.186 24.14 28.64 1.41 19 U 0.388 x 10 0.738 O.O(d) 0.00 0.54 Lower Shell Forging X 1.17 x 10 19 1.044 28.75 30.01 1.09 (Axial) 19 W 1.98 x 10 1.186 37.11 44.03 1.41 SUM: 146.85 6.08 CF LSForging = I:(FF *ARTNDT) + L(FF2 ) = 046.85) + (6.08) =24.1°F 19 U 0.388 X 10 0.738 17.06 12.59 0.54 Braidwood Unit 1 Surveillance Weld X 1.17 x 10 19 1.044 30.15 31.47 1.09 Material 19 W 1.98 x 10 1.186 49.68 58.94 1.41 19 U 0.388 X 10 0.738 O.O(d) 0.00 0.54 Braidwood Unit 2 Surveillance Weld X 1.15 x 10 19 1.039 26.3 27.33 1.08 Material 19 W 2.07 x 10 1.198 23.9 28.63 1.44 SUM: 158.96 6.10 CF Weld Metal =L(FF
- lNDT) + L(FF2 ) = (158.96) + (6.10) = 2ci.1°F Notes:
Source document is CN-AMLRS-1O-7 (Reference I Table 5.2-1 f= values are the measured 30 ft-Ib shift values taken from Reference 6.
FF =fluence factor =
Measured values were determined to be but IJU ,. ' " ' ' ' ' ' ' a reduction should not occur; meret!)re, conservative values of zero are used.
BRAIDWOOD UNIT 1 M PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.2 Braidwood Unit 1 Reactor Vessel Material Properties Chemistry Initial Material Description Cu(%) Ni (%)
Closure Head Flange 0.11 0.67 -- -20 Heat # 5P738113P6406 Vessel Flange
-- 0.77 -- -10 Heat # 122N357V Nozzle Shell Forging
- 26.0°P<b) 0.04 0.73 10 Heat # 5P-7016 Intermediate Shell Forging
- 31.0°F(b) 0.05 0.73 -30 Heat # [49D383/49C344]-I-1 Lower Shell Forging
- 31.0°p<b) 0.05 0.74 24.1°p<e) -20 Heat # [49D867 /49C813 ]-1-1 Circumferential Weld
- 41.0°F(b)
(Intermediate Shell to Lower Shell) 0.03 0.67 26.1°p<e) 40 WF-562 (HT# 442011)
Upper Circumferential Weld *
(Nozzle Shell to Intermediate Shell) 0.04 0.46 54.0°P<b) -25 WF-645 (HT# H4498)
- Beltline Region Materials a) The Initial RTNDT values for the plates and welds are based on measured data.
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.
14
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.3 Summary of Braidwood Unit 1 Adjusted Reference Temperature (ART) Values at 1I4T and 3/4T Locations for 32 EFPy(a) 32EFPY Surface Fluence Reactor Vessel Material 1I4T ART 3/4T ART (nlcm2, E>1.0 MeV)
(OF) (OF)
Nozzle Shell Forging 0.586 x 10 19 47 34 Intermediate Shell Forging 1.76 x 10 19 33 15 Lower Shell Forging 1.76 x 10 19 43 25
--Using credible surveillance data 1.76 x 10 19 21 15 Nozzle to Intermediate Shell Forging Circ. Weld Seam 0.586x 10 19 52 25 (Heat # H4498)
Intermediate to Lower Shell Forging Circ. Weld Seam 1.70 x 10 19 122 99 (Heat # 442011)
--Using credible surveillance data 1.70 x 10 19 93 78 Notes:
The source document containing detailed calculations is CN-AMLRS-1O-7 (Reference 14),
Tables 5.3.1-1 and 5.3. I -2. The ART values summarized in this table utilize the most recent f1uence projections and materials but were not used in development of the PIT limit curves. See rl~ILl'" 2.1 and 2.2 of this PTLR for the ART values used in development of the PIT limit curves.
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 Braidwood Unit 1 Calculation of Adjusted Reference Temperatures (ARTs) at 32 EFPY at the Limiting Reactor Vessel Material, Nozzle Shell Forging 5P*7016 Parameter Values Operating Time 32EFPY Location(a) 1I4T ART(OF) 3/4T ART(OF)
Chemistry Factor, CF (OF) 26.0 26.0 Fluence(f), nlcm 2 3.65 x 10 18 1.32 x10 18 (E>1.0 Mevi b)
Fluence Factor, FF 0.772 0.475 L1RTNDT= CFxFFCOF) 18.8 12.4 Initial RT NDT, 1(°F) 10 10 Margin, M (OF) 18.8 12.4 ART= 1+(CF*FF)+M,oF 48 35 per RG 1.99, Revision 2 (a) The Braidwood Unit I reactor vessel wall thickness is 8.5 inches at the beltline region.
(b) Fluence f, is based upon > 1.0 Mev) = 6.08 x 10 18 at 32 EFPY (Reference II).
16
BRAIDWOOD* UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.5 Calculation for Braidwood Unit 1 Beltline Region Materials at EOL (32 EFPYia,b)
R.G.l.99, (e) au(c) a}d)
CF Fluence IRTNDT ..1RTNDT Margin RTPTS Reactor Vessel Material Rev. 2 FF (OF) (n/cm2, E>1.0 MeV) (OF) ("F) (OF) (OF) (OF)
Position Nozzle Shell Forging 1.1 26 0.586 x 10 19 0.8504 10 22.1 0 11.1 22.1 54 Intermediate Shell Forging 1.1 31 1.76 x 10 19 1.1554 -30 35.8 0 17 34 40 Lower Shell FV15ll!5 L1 31 1.76 x 10 19 1.1554 -20 35.8 0 17 34 50 Ising credible data 2.1 24.1 1.76 x 1019 1.1554 -20 27.8 0 8.5 17 25 to Intermediate Forging Cire. 1.1 54 0.586x 10 19 0.8504 -25 45.9 0 23.0 45.9 67 (Heat #
Intermediate to Forging Cire. Weld 1.1 41 1.70 x 10 19 1.1461 40 47.0 0 23.5 47.0 134 (Heat # 442011)
Tsing credible illance 2.1 26.1 1.70 x 10 19 1.1461 40 29.9 0 14 28 98 The 10 CFR 50.61 was utilized in the calculation of the RTPTS values.
The source document containing detailed calculations is CN-AMLRS-1O-7 (Reference 14), Table 5.5-1.
Initial RTNDT values are based on measured data. Hence, au =O°F.
=
Per the guidance of 10 CFR 50.61, the base metal aL\ 17°F for Position 1.1 (without surveillance data) and with credible surveillance data
=
aL\ :::: 8.5°F for Position 1, the weld metal aL\ 28°F for Position 1.1 (without surveillance data) and with credible surveillance data aLl ::::
14°F for Position L However. aL\ need not exceed O.5*..1RTNDT.
17
BRAIDWOOD - UNIT 1 PRESSURE Al~D TEMPERATURE LIMITS REPORT 6.0 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. NRC Letter from R. A. Capra to 0.0. Kingsley, Commonwealth Edison Company, "Byron Station Units 1 and 2 and Braidwood Station Units 1 and 2, Acceptance for referring of pressure temperature limits report, (M98799, M98800, M98801, and M98802)," January 21 1998.
- 3. Westinghouse Letter to Exelon Nuclear, CAE-1O-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report," M.P. Rudakewiz, September 8, 2010.
- 4. Byron & Braidwood Design Information Transmittal DIT-BRW-2006-0051, "Transmittal of Braidwood Unit 1 and Unit 2 Temperature and Pressure Uncertainties for Low Temperature Overpressure System (LTOPS) Power Operated Relief Valves (PORVS)," Nathan (Joe)
Wolff Jr., July 18, 2006.
- 5. WCAP-9807, "Commonwealth Edison Company, Braidwood Station Unit 1 Reactor Vessel Radiation Surveillance Program," S.E. Yanichko, et at, February 1981.
- 6. WCAP-15316, Revision 1, "Analysis of Capsule W from Commonwealth Edison Company Braidwood Unit 1 Reactor Vessel Radiation Surveillance Program," E. Terek, et aI.,
December 1999.
WCAP-15365, Revision 1, "Evaluation of Pressurized Thermal Shock Braidwood Unit 1," J.H. Ledger, January 2002.
- 8. NRC Letter from G. Dick, Jr., NRR, to C. Crane, Exelon Generation Company, LLC, "J~U.UlA'--' of Amendments: Revised Pressure-Temperature Limits Methodology; Byron 1 and Braidwood 1 " dated October
- 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 of Amendments Re: Reactor Coolant System Pressure and Temperature Limits Report (TAC MC8693, MC8694, MC8695, and MC8696)," November 27,2006.
BRAIDWOOD - UNIT 1 PRESSURE AND TEMPERATURE LIMITS REPORT
- 11. WCAP-15364, Revision 2, "Braidwood Unit 1 Heatup and Cool down Limit Curves for Normal Operation," T.1. Laubham, November 2003.
- 12. WCAP-16143-P, Revision 0, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for ByronlBraidwood Units 1 and 2," W. Bamford, et aI., November 2003.
- 13. Westinghouse Letter to Exelon Nuclear, CCE-07-24, "Braidwood Unit 1 and 2 RCS HU/CD Limit Curve Table Values," dated February 15,2007.
- 14. Westinghouse Calculation Note CN-AMLRS-1O-7, Revision 0, "Braidwood Units 1 and 2 Measurement Uncertainty Recapture (MUR) Uprate: Reactor Vessel Integrity Evaluations,"
A.E. Leicht, September 2010.
ATTACHMENT 3 Braidwood Unit 2 Pressure and Temperature Limits Report (PTLR), Revision 5
BRAIDWOOD UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)
Revision 5
BRAID\VOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Introduction 2.0 RCS Pressure Temperature Limits 2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 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 18
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of lOO°F/hr) Applicable for 32 EFPY (Without Margins for Instmmentation Errors) 2.2 Braidwood Unit 2 Reactor Coolant System Cooldown Limitations 4 (Cooldown Rates of 0, 25,50 and IOO°F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors) 3.1 Braidwood Unit 2 Nominal PORV Setpoints ft)r the Low Temperature 8 Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
BRAID\VOOD
- UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.] a Braidwood Unit 2 Heatup Data at 32 EFPY (Without 5 Margins for Instrumentation Errors) 2.1h Braidwood Unit 2 Cooldown Data Points 32 EFPY (Without 6 Margins for Instrumentation Errors) 3.1 Data Points for Braidwood Unit 2 Nominal PORV 9 Setpoints for the LTOP System Applicable for 32 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 Summary Braidwood Unit 2 Adjusted Reference Temperature (ART) Values at 1 and 3/4T Locations for EFPY 16 2
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIl\UTS REPORT 1.0 Introduction This Pressure and Temperature Limits Report (PTLR) for Braidwood Unit 2 has heen 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 (Pff) Limits; and LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System.
2.0 ReS 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-NP-A. Revision 2 (Reference 1) was used with the following exception:
a) Optional use of ASME Code Section Xl, 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", and c) of ASME Code Case N-588, "AJt.ernative to Reference Flaw Orientation of Appendix G for Circumferential Welds Reactor Vessel, Sect.ion XI.
Division I", and Elimination the flange requirement'> in WCAP-16] 43-P.
This exc:eVlum 2 heen 2.1 RCS nlits 2.1 The RCS temperature rate-of-change limits defined in Reference 11 are:
- a. 111
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERi\' TlJRE LIMITS REPORT
- c. A maximum lemperature change less th:m or equal to 10°F in any I-hour period during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.
1.2 The RCS prr limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1 a. The RCS PiT limits for cooldown are shown in Figure 2.2 and Table 2.1 b. These limits are defined in WCAP-15373, Revision 2 (Reference 11). Consistent with the methodology described in Reference 1, with the exception noted in Section 2.0, the RCS prr 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, Article G2000, 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 prr 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 lest, and at least 40°F higher than the minimum permissible temperature in the corresponding PIT curve for heatup and cooldoWJ1.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 2500 2250 Acceptable 2000 Operation Unacceptable Operation 1750 6'
(j)
-e c.. 1500
- J Critical Limit 100 Oeg. F/Hr
~
c..
1250
"~
"3
(,) 1000 iJ 750 Criticality Limit based on inservice hydrostatic test
'------1 temperature (127 F) for the 500 service period up to 32 EFPV 250 o
o 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Oeg. F)
Figure 2.1 Braidwood Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rate of' lOO°Flhr)
Applicable to EFPY (\VithoutMargins Instrumentation
BRAIDWOOD
- UNIT 2 PRESSURE AND TE~tPERA TURE LIMITS REPORT Material: Circumferential Weld WF-562 & NonIe Shell ART Values at 32 EFPY 1/4T 93"P & 6TF 3/4T 79 n F & 54 DF 2500 2250 Unacceptable Acceptable 2000 Operation Operation 1750 (3'
~
~
- J 1500
~
a.
1250 i
16 Cooldown
- 1000 RatesFlHr C
c5 steady.state
- 25
- 50 750 *100 500 250 o
o 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°Flhr) Applicable to 32 EFPY (Without Margins of Instrumentation Errors)
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1a Braidwood Unit 2 Heatup Data Points at 32 EFPY (Without Margins for Instrumentation Errors)
Heatup Curve 100 F Heatup Criticality Limit Leak Test Limit T ("F) P (psig) T (OF) P (psig) T('F) P (psig) 60 0 127 0 110 2000 60 924 127 965 127 2485 65 965 127 977*
70 977 ]17
-, 977 75 977 127 981 80 977 130 990 85 981 135 1005 90 990 140 1025 95 1005 145 1051 100 1025 150 1081 105 1051 155 1118 110 1081 160 1161 115 1118 165 1210 120 1161 170 1266 125 1210 175 1329 130 1266 180 1400 135 1329 185 1480 140 1400 190 1569 145 1480 195 1668 150 I 1569 200 i778 155 1668 205 1901 160 1778 210 2036 165 1901 215 2186 170 2036 220 2353 175 2186 180 2353
BRAIDWOOD - lJNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1b Braidwood Unit 2 Cooldown Data at 32 EFPY (Without Margins for Instrumentation Errors)
Cooldown Curves Steady State 25 OF Cool down 50 "F Cooldown 100 OF Cooldown T (oF) P <psig} TC'F) P (psjg) T ("F) P (psig) T (OF) P (psig) 60 0 60 0 60 0 60+ 0 60 931 60 908 60 889 60 866 65 965 65 946 65 932 65 921 70 1003 70 989 70 980 70 980 75 1045 75 1036 75 1033 75 1033 80 1092 80 1088 80 1088 80 1088 85 1143 85 1143 85 1143 85 1143 90 1200 90 1200 90 1200 90 1200 95 1263 95 1263 95 1263 95 1263 100 1332 100 1332 100 1332 100 1332 105 1409 105 1409 105 1409 105 1409 110 1494 110 1494 110 1494 110 1494 115 1587 115 1587 115 1587 115 1587 120 1691 120 1691 120 1691 120 1691 125 1805 125 1805 125 1805 125 1805 130 1932 130 1932 130 1932 130 1932 135 2071 135 2071 135 2071 135 2071 140 2226 140 2226 J40 2226 140 2226 145 2396 145 2396 145 2396 145 2396
BRAIDWOOD - IJNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 Low Temperature Overpressure Protection and Boltup This section provides the Braidwood Unit 2 power operated relief valve lift low temperature overpressure protection (LTOP) system arming temperature, and minimum reactor vessel boltup temperature.
3.1 LTOP System Setpoinls (LeO 3.4.12).
The power operated relief valves (PORVs) shall each have nominal lift settings in accordance with Figure 3.1 and Table 3.1. These limits are based on References 3 and 8.
The LTOP setpoints are based on Pff 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 nominal lift settings shown in Figure 3.1 and Table 3.1 account for appropriate instrument error.
3.2 LTOP Enable Temperature Braidwood Unit 2 procedures governing the heatup and cooldown of the ReS require the arming of the LTOP System for ReS temperature of 350°F and below and disarming of LTOP for ReS temperature above 350°F.
Note that the LTOP PORV segment in Table 1 extends to 400°F the pressure selpoint is psig. is intended to prohibit PORV lift an nQfnU>.-TC"n. LTOP system arming at "",Up,"
Boltup
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT 8'
in e
$ Unacceptable Operation e'"
Q. 1250 a:
oQ.
pev 456 pev 455A o 50 wo 150 200 250 300 350 400 450 Auctioneered Low ReS Temperature (OEG. F)
Figure 3.1 Braidwood Unit 2 Nominal PORV Setpoints for the Low Temperature Overpressure Protection (LTOP) System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 Data Points for Braidwood Unit 2 Nominal PORV Set points for the LTOP System Applicable for 32 EFPY (Includes Instrumentation Uncertainty)
PCV-455A PCV-456 RCS TEMP. RCS Pressure RCSTEMP. ReS Pressure (DEG. F) (PSKi) 60 599 300 599 300 639 400 2335 400 2335 Note: To determine nominal lift setpoints for RCS Pressure and RCS Temperatures greater than 300 0 E 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 system arming while at power).
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE l,,1MITS REPORT 4.0 Reactor Vessel Material Surveillance Program The pressure vessel material surveillance program (Reference is in compliance with Appendix H to 10 CFR 50, "Reactor Vessel Radiation Surveillance Program:' The material test requirement') 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 E 185-82.
The third and final reactor vessel material irradiation surveillance specimens (Capsule W) have been removed and analyzed to determine changes in material properties. The surveillance capsule testing has been completed for the original operating period. The remaining three capsules, V, 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.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.1 Braidwood Unit 2 Capsule 'Vithdrawal Summary(a)
Capsule 'Vithdrawal EFPy(b) Fluence Capsule Lead Factor 2 Location (nlcm , E>1.0 MeV)
U 58.5° 4.08 1.18 0.388 x 1O J9 X 238.5° 4.03 4.24 1.15 x 10 19 I
W 121.5° 4.06 8.56 2.07 x 10 19 I
Z(C) 301.5° 4.14 12.78 (EOC 10) 2.83 x 10 19 I
V(C) 61.0° 3.92 18.42 (EOC 14) 3.73 x 10 19 I
y(c) 241.0° 3.89 12.78 (EOC 10) 2.66 X 10 19 I
Notes:
Source document is CN-AMLRS-IO-7 (Reference I Table 5.7-4.
(b) Effective Full Power Years (EFPY) from plant startup.
Standby Z, V. and Y were removed and in the fuel pool. No or has been pcft"cJlHlcd on these capsules. lflicense renewal is one of these
"'41"""Y may need to be tested to determine the effect of neutron irradiation on the reactor vessel surveillance materials the of extended VP"'I "UVB.
BRAIDWOOD - UNIT 2 PRK"tSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables The rahles provide suppIemenral information on reactor material properties and are provided to he consistent with Generic Letter 96-03. Some of the material property values shown were used as inputs to the prr limits.
Table 5.] 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 J/4T and 3/4T locations for 32 EFPY.
Table 5.4 shows the calculation of ARTs at 32 EFPY for the limiting Braidwood Unit 2 reactor vessel material.
Table 5.5 provides the RTpTs Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPY), (Reference 6).
BRAIDWOOD UNIT 2
- PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.1 Braidwood lJnit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data(a)
Capsule t,b) FF(C) .1RT"IlH (h)
.FF* .1RT'iDT Material Capsule FF2 (n/cnl, E > 1.0 MeV) (OF) (OF) 19 U 0.388 x 10 0.738 0.0(<1' 0.00 0.54 Lower Shell Forging X U5 x 10 10 1.tB9 O.O(d( 0.00 1.08 (Tangential)
W 2.07 x IO 'Y 1.198 4.53 5.43 1.44 U 0.388 x IO!'! 0.738 O.O'd! 0.00 0.54 Lower Shell Forging X U5 x 101') 1.039 33.94 35.26 1.08 (Axial) 19 W 2.07 x 10 1.198 33.2 39.78 1.44 SUM: 80.47 6.12 CF LS lorgjn~
2
- I(FF *.1RTNDr) + I(FF ) (80.47) + (6.12) 13.2°F U 0.388 X 10)') 0.738 17.06 12.59 0.54 Braidwood Unit 1 19 Surveillance Weld X 1.17 x 10 1.044 30.15 31.47 1.09 Material W 1.98 x Ll86 49.68 58.94 1.41 U 0.388 x 0.738 0.00 0.54 Braidwood Unit 2 Surveillance Weld X U5 x 1.039 26.3 27.33 l.O8 Material W 2.07 x I 198 23.9 28.63 1.44 SUM: 158.96 6.10 CF WddMelal:::; I(FF l "'in) + I(f'"F::>> = .96) (6. ) ::::: 26.1".1"
BRAIDWOOD - lJNIT 2 PRESSURE AND TEMPERATURE LI1\UTS REPORT Table 5.2 Braidwood Unit 2 Reactor Vessel Material Properties Chemistry Initial Material Description Cu(%) Ni (%)
Closure Head Flange Heat # 3P6566/5P7547/4P6986 0.75 -- 20 Serial # 2031-V -1 Vessel Flange 0.07 0.70 -- 20 Heat # 124P455 Nozzle Shell Forging
- 26.0cp b) 0.04 0.90 30 Heat # 5P-7056 Intermediate Shell Forging
- 20.0cpb)
Heat # [49D963/49C904]-1-1 0.03 0.71 -30 Lower Shell Forging
- 37.0°F(b) 0.06 0.76 13.2°pc) -30
, Heat # [50D102/50C97}-1-1 Circumferential Weld *
(Intermediate Shell to Lower Shell) 41.0pb)
Weld Seam WF-562 0.03 0.67 26.1 pC) 40 Heat # 442011 Circumferential Weld *
(Nozzie Shell to Intermediate 0.04 0.46 54.0 o po) -25 Weld Seam WF-645 Heat#H4498
'" Beltline Region Materials a) The Initial for the and welds are based on rnelL'iured data.
'hpr,,,,,,b""r Factor calculated Cli and Ni values per Guide 1.99. Rev. 2. Position 1.1 Factor calculated for Cu and Ni values per Guide 1 Rev. Position I
BRAIDWOOD - lJNIT 2 PRESSURE AND TEMPERA TORE LIMITS REPORT Table 5.3 Summary of Braidwood Unit 2 Adjusted Reference Temperature (ART) Values at 1I4T and 3/4T Locatious for 32 EFPy(al Surface Fluence 32EFPY Reactor Vessel Material (nlcm 2, E> 1.0 Me V) 1I4T ART (OF) 3/4T ART (o}<')
19 Nozzle Shell Forgi ng 0.559 x 10 66 54 19 Intermediate Shell Forging J.73 x 10 JO -1 Lower Shell Forging 1.73 x 10 11) 41 24 19
-Using non-credible surveillance data 1.73 x 10 -3 II Nozzle to Intermediate Shell Forging eire. Weld Seam 0.559x 10 19 51 24 (Heat # H4498)
Intermediate to Lower Shell Forging eire. Weld Seam 1.67 x 10 19 122 99 (Heat # 442011) ing credible surveillance 1.67 x 92 78 The source document detailed calculations CN-AMLRS-I 0-7 Tables 5.3.1-3 and 5.3.] -4. The ART values summarized in this table utilize the most recent fluence pro'ICCU0I1S and materials data. but were not used in de'lelclprnicnt of the Pff limit curves. See 2.1 and 2.2 of this PTLR for Ihe ART values used in of the Pff limit curves.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.4 Braidwood Unit 2 Calculation of Adjusted Reference Temperatures (ARTs) at 32 EFPY at the Umiting Reactor Vessel Material, Nozzle Shell Forging 5P-7056 Parameter Values Operating Time 32 EFPY Location(a) 1I4T ART (OF) 3/4T ART(OF)
Chemistry Factor, CF (OF) 26.0 26.0 Fluence(f), n/cm-3.40x 1011i 1.23x 10 1/\
(E>l.O Mev)(b)
Fluence Factor, FF 0.703 0.460 8RTNDT= CFxFF(OF) 18.3 12.0 Initial RT NDT., I(OF) 30 30 Margin, M (OF) 18.3 12.0 ART= J+(CF*FF)+M, OF 67 54 per RG 1.99, Revision 2 a) The Braidwood Unit 2 reactor vessel waH thickness is 8.5 inches at the beltline b) cladlbase metal interface t1uence .0 at32
BRAIDWOOD* UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT
,-"'~,,~
Table 5.5 RT l'TS Calculation for Braidwood Unit 2 Beltline Region Materials at EOL (32 EFPy)~a,b)
R.G.t.99, (c) (e) (iiI CF Fluence IRTNIH ARTNJ)T Gu (JA Margin RTl'rs Reactor Vessel Material Rev. ,2 (OF) FF (OF) (OF) (OF) (OF) (OF)
(nlcm2, 1';>1.0 MeV) (OF)
Position Nozzle Shell Forging 11 26 0,559 x ]{)19 0,8373 30 21.8 () 10,9 21.8 74 19 lntermediate Shell Fnrgi:lg LI 20 L73 X 10 1.1508 -30 23.0 0 IL'i 23.0 16 Lower Shell lrgi I I 37 L73 x IOI'! 1.1508 -30 42.6 0 17 34 47
""""~~'~M 19 ling non-credihle data 2,1 13.2 1.73 x 10 1.1508 -30 15.2 () 7,6 15.2 0 Noule to Intermediat<,'
eire. Weld I I 54 0.559x 10 19 0,8373 -25 45.2 0 22,6 45.2 65 (Heal # H449):;)
Intermediate 10 Circ, Weld 1.1 41 1.67 x 10 1') 1.1413 40 46J~ 0 23.4 46.8 134 (I-kat # 11 JSll1g credible data 2.1 26.1 1.67 x 10 19 1.1413 40 29.8 0 14 28 98 The 10 CFR was utilized in trw calculation of the RTI'TS values.
The source document detailed calculations is CN-AMLRS-10-7 (Reference 14). Table 5.5-2.
haSt~d on ITIl~asured data, Hence. 0" O"F.
10 eFR 50.61. the base metal 1T'F for Position 1.1 (without surveillance data) and for Position 2.1 with non~erediblc lTletal 28"F ror Position 1.1 (without surveillance data) and with credihle surveillance data v.~::: 14F for not e)(et'ed 17
BRAIDWOOD - lJNIT 2 PRESSURE AND TEMPERATVRE LIMITS REPORT 6.0 References I WCAP-14040-NP-A, Revision 2, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoint<; and RCS Heatup and Cooldown Limit Curves", J.D.
Andrachek, et aL January 1996.
- 2. NRC Lener from R. A. Capra to 0.0. Kingsley, Commonwealth Edison Company, "Byron Station Units I and 2 and Braidwood Station Units 1 and Acceptance for referring of pressure temperature limits report, (M98799, M98800, M9880 Land 1\198802)," January 21, 1998.
- 3. Westinghouse Letter to Exelon Nuclear, CAE-I0-MUR-197, Revision 0, "Low Temperature Overpressure Protection (LTOP) System Evaluation Final Letter Report."
M.P. Rudakewiz, September 8,2010.
- 4. WCAP-11188, "Commonwealth Edison Company, Braidwood Station Unit 2 Reactor Vessel Surveillance Program," December 1986.
- 5. WCAP-I5369, "Analysis of Capsule W from the Commonwealth Edison Company Braidwood Unit 2 Reactor Vessel Radiation Surveillance Program," March 2000.
- 6. WCAP-15381, "Evaluation of PressUlized Thermal Shock for Braidwood Unit 2", T.1.
Laubham, September 2000.
NRC Letter from G. F. Dick, Jr., NRR. to C. Crane, Exelon Generation Company. LLC, "Issuance of Amendments: Revised Pressure-Temperature Limits Methodology~ Byron Station, ] 2, Braidwood Station, Units I 2," dated October 4, 2004.
Byron Braidwood Design Information TransmittaJ DIT-BRW-2006-005l, Braidwood Unit 1 cmd Unit 2 Temperature and Pressure Uncertainties (LTOPS) Operated Relief (TAC Nos. MC8693, MC8694, MC8695, 1 I. WCAP-15373, Revision 2, "Braidwood Unit 2 Heatup and Cooldown Limits for Normal Operation," TJ. Laubham et aL November 2003.
BRAIDWOOD - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT WCAP-16143-P, Revision 0, "Reactor Vessel Closure Head/Vessel Flange Requirements Evaluation for Byron/Braidwood Units I and 2:' W. Bamford, et aL November 2003.
- 13. Westinghouse Letter to Exelon Nuclear, CCE-07-24. "Braidwood Unit 1 and 2 RCS HU/CD Limit Curve Table Values," dated February I 2007.
Calculation Note CN-AMLRS-1O-7. 0, Units I and 2 Measurement Uncertainty Recapture (MUR) Upratc: Reactor VesseJ Integrity Evaluations," A.E. Leicht. September 2010, and Westinghouse evaluation MCOE-LTR-13-102 Rev. 0, "Byron and Braidwood Closure Head/Vessel Flange Region: MUR Uprate Assessment," November 2013.