ML20207H762

Pressure Temp Limits Rept (Ptlr)

ML20207H762
Person / Time
Site:	Byron
Issue date:	06/28/1999
From:	COMMONWEALTH EDISON CO.
To:
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ML20207H753	List: ML20207H750 ML20207H756 ML20207H762 ML20207H767 ML20207H773 ML20207H777 ML20207H785 ML20207H794 ML20207H801 ML20207H807
References
NUDOCS 9907210204
Download: ML20207H762 (40)

Text

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BYRON UNIT 2 PRESSURE TEMPERATURE LIMITS REPORT (PTLR) l (June 281999)

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F BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT L

l Table of Contents Section Page l

1.0 Reactor Coolant System (RCS) Pre:<ure and Temperature Limits Report (PTLR) 1 2.0 Operating Limits 1

2.1 RCS Pressure and Temperature (P/T) Limits 1

l 2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints 2

2.3 LTOP Enable Temperature 2

2.4 Reactor Vessel Boltup Temperature 2

2.5 Reactor Vessel Minimum Pressurization Temperature 3

3.0 Reactor Vessel Material Surveillance Program 9

4.0 Supplemental Data Tables 11 5.0 References 18 l

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BYRON - UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page

.2.1 Byron Unit 2 Reactor Coolant System Heatup Limitations (Heatup Rates up to 4

- 100 F/hr) Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) 2.2 Byron Unit 2 Reactor Coolant System Cooldown Limitations (Cooldown Rates 5

up to 100 F/hr) Applicable for the First 16 EFPY (Without Margins for

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Instrumentation Errors; Using 1996 Appendix G Methodology) 2.3 Byron Unit 2 Maximum Allowable Nominal PORV Setpoints for the Low 7

i Temperature Overpressure Protection (LTOP) System Applicable for the First 16 EFPY iii

BYRON - UNIT 2 PRESSURE 'AND TEMPERATURE LIMITS REPORT List of Tables Table Page 2.1 Byron Unit 2 Heatup and Cooldown Data Points at 16 EFPY 6

(Without Margins for Instrumentation Errors) 2.2 Data Points from Byron Unit 2 PORV Setpoints for the LTOP System 8

3.1 Byron Unit 2 Capsule Withdrawal Schedule 10 4.1 Byron Unit 2 Calculation of Chemistry Factors Using Surveillance Capsule Data 12 4.2 Byron Unit 2 Reactor Vessel Material Properties 13 4.3 Summary of Byron Unit 2 Adjusted Reference Temperatures (ARTS) at the 1/4T 14 and 3/4T Locations for 16 EFPY 4.4 Byron Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 15 16 EFPY at the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data) 4.5 RTers Values for Byron Unit 2 for 32 EFPY 16 4.6 RTers Values for Byron Unit 2 for 48 EFPY 17 is

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

This PTLR for Unit 2 has been prepared in accordance with the requirements of TS-5.6.6. Revisions to the PTLR shall be provided to the NRC after issuance.

iThe Technical Specifications addressed in this report are listed below.

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

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

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

- WCAP-15178, Reference 14, provides the basis for the Byron Unit 2 P/T curves, along with the best estimate chemical compositions, fluence projections, and adjusted reference temperatures used to determine these limits. The weld metal data integration for Byron and Braidwood Units 1 and 2 is documented in Reference 2.

2.1 RCS Pressure and Teuperature (P/T) Limits TS-LCO 3.4.3 2.1.l The RCS temperature rate-of-change limits defined in Reference 14 are:

a. A rnaximum heatup of 100 F in any 1-hour period,

b. A maximum cooldown of 100'F in any 1-hour period, and I

A maximum temperature change ofless than or equal to 10 F in any 1-hour period c.

during inservice hydrostatic and leak testing operations above the heatup and cooldown limit curves.

2.1.2 The RCS P/T limits for heatup, inservice hydrostatic and leak testing, and criticality are specified by Figure 2.1 and Table 2.1. The RCS P/T limits for cooldown are shown in Figure 2.2 and Table 2.1. These limits are defined in Reference 14. Consistent with the i

methodology described in Reference 1, the RCS P/T limits for heatup and cooldown shown in Figures 2.1 and 2.2 are provided without margins for instrument error. These limits were developed using ASME Code Section XI, Appendix G, Article G-2000,1996 i

Addenda. The criticality limit curve specifies pressure-temperature limits for core 1

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!1 BYRON - UNIT 2 L

E

' PRESSURE AND TEMPERATURE LIMITS REPORT

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_ Operating Limits (continued) operation to provide additional margin during actual power production as specified in 10 CFR 50i Appendix'G.

The P/T limits for core operation (except for low power physics testing) are that the reactor vessel must be at a temperature equal to or higher than the minimum temperature required for the inservice hydrostatic test, and at least 40 F higher than the minimum i-

permissible temperature in the corresponding P/T curve for heatup and cooldown.

Low Temperature Overpressure Protection (LTOP) System Setpoints TS-LCO 3 4.12.

2.2 '

The power operated relief valves (PORVs) shall each have maximum lift settings in -

accordance with Figure 2.3 and Table 2.2. These limits are based on References 5,13 and 15. The Residual Heat Removal (RH) Suction Relief Valves are also analyzed to 4

l individually provide low temperature overpressure protection. This analysis for the RH Suction Relief Valves remains valid with the current Appendix G limits contained in this

. PTLR document and will be reevaluated in the future as the Appendix G limits are

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I revised.

The LTOP setpoints are based on P/T limits which were established in accordance with j

10 CFR 50, Appendix G without allowance for instrumentation error and in acco'rdance l

with the methodology described in Reference 1. The LTOP PORV maximum lift settings shown in Figure 2.3 and Table 2.3 account for appropriate instrument error.

1 2.3 LTOP Enable Temperature The as-analyzed LTOP enable temperature is 200*F (References 14 and 16).

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

l (Byron Unit 2 procedures governing the heatup and cooldown of the RCS require the L

arming of the LTOP System for RCS temperature of 350 F and below and disarming of L

LTOP for RCS temperature above 350 F).

l l

Note that the last LTOP PORV segment in Table 2.2 extends to 450 F where the i

l pressure setpoint is 2350 psig. This is intended to prohibit PORV lift for an inadvertent l

LTOP system arming at power.

. Reactor Vessel Boltup Temperature (Non-Technical Specification) 2.4 i

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

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

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

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1 CYRON - UNIT 2 -

PRESSURE AND TEMPERATURE LIMITS REPORT i

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F)FORTHE SERVICE PERIOD UP TO 160 0-50 100 150 200 250 300 350 400 450 500 ModeratorTemperature(Deg. F) i l

i Figure 2.1 Eyron Unit 2 Reactor Coolant System Hestup Limitations (Heatup Rates up to 100 F/hr)

Applicable for the First 16 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology).

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

Applicable for the First_16 EFPY (Without Margins for Instrumentation Errors; Using 1996 Appendix G Methodology) -

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BYRON - UNIT 2 Y-PRESSURE'AND TEMPERATURE LIMITS REPORT Table 2.1 Byron Unit 2 Heatup and Cooldown Data Points at 16 EFPY

-(Without Margins for lustrumentation Errors and Using the 1996 Appendix G Methodology)-

1 Hastup Curve Cooldown Cunes 100 F.

Cnticakty LaakTest steady

,25 F 50 F 100 F Hestu)

Umt Unit

-State Cocidan Cooldom Cooldom T.

3 T

P T

P T

P T

P T

P T

P 60 0

219 0

198 2000 60 0

60 0

60 0

60 0

60 621 219 635 219 2485 60 621 60 574 60 523

.60' 418 65 621 219 674 65 621 65 585 65 534 65 431 85 621 219 660 70 621 70 597 70 547 70 446 90 621 219 650 75 621 75 610 75 561 75 462 95 621 219 643 80 621 80 621 80 576 80 480 100 621 219 638 85 621 85 621 85 592 85 498

- 105 621 219 637 90 621 90 621 90 609 90 519 110 621 219 637 95 621 95 621 95 621 95 541 115 621 219 641 100 621 100 621 100 621 100 564 120 621 219 646 105 621 105 621 105 621 105 590 125 621 219 654 110 621 110 621 110 621 110 -618 130 621 219 664 115 621 115 621 115 621 115 621 135 621 219 676 120 621 120 621 120 621 120 621 140 621 219 690 125 621 125 621 125 621 125 621 145 621 219 707 130 621 130 621

-130 621 130 621 150 621 219 725 135 621 135 621 135 621 135 621 150 707 219 746 140 621 140 621 140 621 140 621 155 725 219 770 145 621 145 621 145 621 145 621 160 746 219 796 150 621 150 621 150 621 150 621 165 770 219 824 150 995 150 974 150 957 150 935 170 796 220 855 155 1033 155 1016 155 1002 155 989 175 824 225 889 160 1075 100 1061 160 1051 160 1048 180 855 230 926 165 1119 165 1109 165 1104 165 1112 185 85 235 92 170 1166 170 1161 170 1161 190 926 240 1010 175 1217 175 1217 195 986 245 1057 180 1272 200 1010 250 1107 185 1331 205 1057 255 1182 190 1305

-210 1107 200 1221 195 1463 215 1162 265 1285 200 1536 220 1221 270 1353 205 1615 225 19A'i 275 1427 210 1700 230 1353 280 1506 215 1791 235 1427 285 1591 220 1889 240 1500 290 1683 225 1995 245 1591 295 1781 230 2108 250 1m 300 1887 235 2230 255 1781 305 2001 240 2361 260 1887 310 2123 265 2001 315 2254 270 2123 320 2305 275 2254 zau z.ao Noe 1: lleanup and Cooldown data aneludes the w I flange requirements of I80 F and 62I psig per 10CFRSO. Appenda O Nose 2: For caeh cooldown rate, the steady state pre.aure values shall govern the temperature uhere no allowable preuure values are pronded 4

Wate 3: Temperatures and prenavres are yven in

• F and poig, respectnely.

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O 50 100 150 200 250 300 350 AUCTIONEERED lJ0W RCS TENPERATURE (DEG F)

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BYRON - UNIT 2 j

PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2 Data Points for Byron Unit 2 Maximum Allowable PORY Setpoints for the LTOP System Applicable for the First 16 EFPYa PCV-455A PCV-456 (2TY-0413M)

(2TY-0413P)

' AUCTIONEERED LOW RCS PRESSURE AUCTIONEERED LOW RCS PRESSURE RCS TEMP. (DEG F)

(PSIG)

RCS TEMP. (DEG. F)

(PSIG) 50 497 50 514 70 497 70 514 100 497 100 514 120 446 120 462

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150 446 150 462 200 446 200 462 250 587 250 604 300 587 300 604 i

350 587 350 604 450 2350 450 2350 Note: To determine maximum allowable lift setpoints for RCS Pressure and RCS Temperatures greater than 350 F, linearly interpolate between the 350 F and 450 F data points shown above. (Setpoints extend to 450 F to prevent PORY liftofT from an inadvertent LTOP system arming while at power.)

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CYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT It 3.0

' Reactor Vessel Material Surveillance Program i

The pressure vessel material surveillance program (Reference 6) is in compliance with Appendix H to 10 CFR 50," Reactor Vessel Radiation Surveillance Program." The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTmi, which is determined in accordance with ASME Section 111, NB-2331. The empirical relationship between RTmr and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G," Protection Against Non-Ductile Failure," to Section XI of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

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

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BYRON - UNIT 2 l

l PRESSURE AND TEMPERATURE LIMITS REPORT Table 3.1 l

Byron Unit 2 Capsule Withdrawal Schedule Capsule Vessel Location Capsule Lead Remova! Timeta)

Estimated Capsule 2

(Degrees)

Factor (EFPY)

Fluence (n/cm )(c)

U 58.5 4.40 1.15 (Removed) 4.05 x 10

I W

121.5 4.25 4.634 (Removed) 1.27 x 10

X 238.5 4.25 8.573 (Removed at 2.30 x 10(b)

EOL Wall)

Z(d) 301.5 4.21 12.8 (c) 3.35 x 10

V(e) 61.0 3.97 13.6 (c) 3.35 x 10

Y(c) 241.0 3.07 13.6 (c) 3.35 x 10

(a) Effective Full Power Years (EFPY) from plant :'.7,nup.

2 (b) Maximum end oflicense (32 EFPY) inner vessel wall fluence is estimated to be 1.99 x 10" n/cm.

(c) Derived from Tables 6-13 and 7-lof WCAP 15176, Rev. 0 (Reference 18).

(d) This Standby capsule, to be used for future life extension, will reach a fluence of approximately 3.35 x 10(54 EFPY Peak Wall Fluence) at approximately 12.8 EFPY.

(c) These Standby capsules, to be used for future life extension, will reach a fluence of approximately 3.35 x 10(54 EFPY Peak Wall Fluence) at approximately 13.6 EFPY.

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

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

Table 4.2 provides the reactor vessel material properties table.

Table 4.3 provides a summary of the Byron Unit 2 adjusted reference temperature (ARTS) at the 1/4T and 3/4T locations for 16 EFPY.

Table 4.4 shows the calculation of ARTS at 16 EFPY for the limiting Byron Unit 2 reactor vessel material, i.e. weld metal HT # 442002, (Based on Surveillance Capsule Data).

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

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

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 4-8 Calculation of Chemistry Factors for Byron Unit 2 using Surveillance Capsule Data 2

Material Capsule Capsule f*'

FF**

ARTw1"'

FF*ARTwr FF Lower Shell Forging U

0.405 0749 0 0

o 0 561 49D330/49C298-1 1 W

l.27 1.067 3 65 3 89 1.138 (Tangential)

X 2.30 1.225 15 75 19 29 1500 Lower Shell U

0405 0.749 19.76 14 80 0 561 Forging 49D330/

W l.27 1.067 31.88 34 02 1.138 49C2981-1 X

2.30 1.225 38.91 47 66 1.500 SUM:

119.66 6.398 CF,,m = Z(FF

• RTm7) + E( FF;) = (l19.66) + (6 398) = 18.7*F ByTon Unit i Sun. Weld U

0.404 0.749 11.22 (5.61)'d' 8.40 0.561 Material (Heat # 442002)

X 1.57' l.125 80.22 (40.I1)

• 90.25 1.266 W

2.43 1.239 102.68 (51.34)'d' 127.22 1.535 ByTon Unit 2 Sun. Weld U

0.405 0.749 16.88 (8.44)'d' 12.64 0.561 Material (Heat # 442002)

W l.27 1.067 57.76 (28.88)*

61.63 1.138 X

2.30 1.225 108.02 (54.0!)*d' 132.32 1.500 SUM:

432.46 6.561 CFs wa, moo: = E(FF

• RTm7) + E( FF ) = (432 46) + (6.561) = 65.9*F

2 N:tes:

(a) Byron Unit I and 2 capsule fluences were updated as a part of the capsule X dosimetry analysis results (Ref.18), (x 10" n/cm, E > 1.0 MeV).

2 (b) FF = fluence factor = f*

2 ' '* 0 (c) ARTm7 values are the measured 30 ft-lb shift values taken from Ref.18.

(d) The Byron I & 2 surveillance weld metal ARTm1 values have been adjusted by a ratio factor of 2 00.

No temperature adjustment are required.

(c) Actual value of ARTer is -3.8. This physically should not occur, therefore for conservatism (i.e. higher chemistry factor) a value of zero will be used.

(f) : Surveillance data credibility assessment determined data from both Units I and 2 produced the limiting CF, WCAP-15180, Reference 19.

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.2 Byron Unit 2 Reactor Vessel Material Properties Material Description Cu (%)"'

Ni (%)"'

Ch aistry Initial F:ator*'

RT w ('F)"'

Closure Head Flange Not 5P7382 / 3P6407 Reported 0.74 0'd' Vessel Flange Not 124L556 val Reported 0.73 3 0'd' Nozzle Shell Forging 4" 6107(')

0.05 0.74 31 0 10 Inter. Shell Forgir.g 49D329-1/49C297-1 0.01 0.70 20.0

-20 Lower Shell Forging 49D330/49C298-1-1 0.05 0.71 32.2

-20 Circumferential Weld WF-447 (HT# 442002) 0.04 0.63 68.0 10 Upper Circumferential Weld WF-562 (HT# 442011) 0.03 0.67 41.0 40 a) Best estimate chemistry values from WCAP-15178, Reference 14.

b) Chemistry Factors are calculated from Cu and Ni values per Regulatory Guide 1.99. Rev. 2 (Reference 12).

c) Initial RTwrvalues are measured, WCAP-15178 (Reference 14) d) Closure head and vessel flange Initial RTervalues are used for considenng flange requirements for the heatup/cooldown curves, WCAP-15178 (Reference 14).

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BYRON - UNIT 2

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PRESSURE AND TEMPERATURE LIMITS REPORT l

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Table 4.3

' Summary of Byron Unit 2 Adjusted Reference Temperatures (ARTS) at 1/4T and 3/4T Locations for 16 EFPY 16 EFPY i

Material Description 1/4T ART (*F) 3/4T ART ( F)

Intermediate Shell Forging 14 4

49D329-1/49C297-1 (RG Position 1 ))

Lower Shell Forging

'43 24 49D330-1/49C298-1 (RG Position l))

Using capsule data 12-2 (RG Position 2)

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

(RG Position 1(*))

Using credible surveillance 94*'

77**

capsule data (RG Position 2('))-

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

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

-Using credible surveillance 57 50

' capsule data (RG Position 2(*))

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

(b) These ART values were used to generate the Byron Unit 2 Heatup and Cooldown Curves WCAP.

15178 (Reference 14).

14

q DYRON - UNIT-2 PRESSURE AND TEMPERATURE LIMITS REPORT l

Table 4.4 L

Byron Unit 2 Calculation of Adjusted Reference Temperatures (ARTS) at 16 EFPY at the Limiting Reactor Vessel Material Weld Metal (Based on Surveillance Capsule Data) l Parameter.

Values Operating Time 16 EFPY Location *)

1/4T ART 3/4T ART Chemistry Factor, CF ( F) 65.9 65.9 Fluence (f), n/cm' 5.92y. ' O

2.14x10"

- (E>l.0 Mev)")

Fluence Factor, FF 0.853 0.586 ARTer= CFxFF(*F) 56.2 38.6 Initial RT mT., I("F) 10 10 I

Margin, M ("F) 28.0 28.0 j

ART = I+(CF*FF)+M, F 94 77 per RG 1.99, Revision 2 (a) Fluence, f, is based upon f,a(E>l.0 Mev) = 9.86x10'" at 16 EFPY, WCAP-15178 (Reference 14).

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

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT f

Table 4.5 RTers Values for Byron Unit 2 for 32 EFPY Material Fluence'd FF" CF ARTns"'

Margin

. RTsnrin'"

RTns'"

(n/cm',

(*F)

(*F)

(*F)

(*F)

(*F)

E>l.0 MeV)

Intermediate Shell Forging 1.99 1.19 20 23 8 23.8

-20 28 Lower Shell Forging 1.99 1.19 37 44.0 34

-20 58 Lower Shell Forging 1.99 1.19 18.7 22.3 17

-20 19

-+ Using S/C Data'"

Nrzzle Shell Forging 0.503 0.808 31 25.0 25 10 60 Inter, to Lower Shell Cire. Weld 1.96 1.18 54 63.7 56 10 130 Inter. to Lower Shell Cire. Weld 1.96 1.18 65.9 77.8 28 10 116 *

-* Using S/C Data

Nrzzle Shell to Inter. ShcIl Cire.

0.503 0.808 41 33.1 33.1 40 106 Weld N:zzle Shell to Inter. Shell Cire.

0.503 0.808 16.7 13.5 13.5 40 67 Weld

{

-+ Using S/C Data'"

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

(b) FF (Fluence factor) = f' "' 5 0 (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 12).

(d) ARTns = CF

• FF Initial RTsw alues are measured values (See Table 4.2)

(c) v (f) RTns = RTswan + ARTns + Margin ('F)

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

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BYRON - UNIT 2 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.6 RTers Values for Byron Unit 2 for 48 EFPY

- Material Fluence

FF*

CF ARTns"'

Margin RTsnin-l" RTns*

(n/cm,

(*F)

(*F)

(*F)

(*F)

(*F) 2 E>1.0 MeV)

Intermediate Shell Forging 2.98 1.29 20 25.8 25.8

-20 32 Lower Shell Forging 2.98 1.29 37 47.7 34 20 62 r S ell Forging 2.98 1.29 18.7 24.1 17 20 21

-+ Using S/C Data

Nozzle Shell Forging 0.753 0.920 31 28.5 28.5 10 67 Inter. to Lower Shell Circ. Weld 2.93 1.29 54 69.7

$6 10 136 Inter, to Lower Shell Cire. Weld 2.93 1.29 65.9 85 28 10 123'8

-+ Using S/C Data Nozzle Shell to Inter. Shell Cire.

0.753 0.920 41 37.7 37.7 40 115 Wild Nizzle Shell to Inter. Shell Cire.

0.753 0.920 16.7 15.4 15.4 40 7I Weld

-+ Using S/C Data"'

(a) Fluence projections for 48 EFPY from Byron 2 PTS report, WCAP-157177 (Reference 9)

(b) FF (Fluence Factor) = f< 2mm ri (c) Calculated using a CF based on surveillance capsule data per RG 1.99, Position 2 (Reference 12).

(d) ARTns = CF

• FF (c) Initial RTwot values are measured values (See Table 4.2)

(f) RTns = RTwoT(m + ARTns + Margin ('F)

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

t l

17

p 1

i I

BYRON - UNIT 2 1

i PRESSURE AND TEMPERATURE LIMITS REPORT f

5,0 -

References 1.

WCAP-14040-NP A, Revision 2," Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Anarachek, J.D.' et. al., January 1996.

2.

WCAP-14824, Revision 2 " Byron Unit i Heatup and Cooldown Limit Curves for Normal Operation and Surveillance Weld Metal Integration for Byron & Braidwood,"

November,1997 plus Westinghouse errata letters CAE-97-220, dated November 26, 1997 and CAE-97-231 and CAE-97-233, dated January 6,1998.

3.

WCAP-14064," Analysis of Capsule W from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," Malone, M.J., et al, July 1994.

l 1

4.

WCAP-12431," Analysis of Capsule U from the Commonwealth Edison Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," Terek, E., et al., October 1989.

5.

Westinghouse Letter to Commonwealth Edison Company, CAE-96-106, " Byron Unit I and 2 LTOPS Setpoints Based on 10 and 12 EFPY P/T Limits," January 17,1996.

6.

WCAP-10398," Commonwealth Edison Company, Byron Station Unit 2 Reactor Vessel Radiation Surveillance Program," Singer, L.R., December 1983.

4

7.. WCAP-14063," Commonwealth Edison Company, Byron Unit 2 Heatup and Cooldown Limit Curves for Normal Operation," Peter, P. A., November 1994.

' 8.

WCAP-14044," Westinghouse Surveillance Capsule Neutron Fluence Reevaluation,"

l Lippencott, E.P., April 1994.

l 9.

WCAP-15177, " Evaluation of Pressurized Thermal Shock for Byron Unit 2," Revision 0, T. J. Laubham, et al., June 1999,

10. 10 CFR Part 50, Appendix G," Fracture Toughness Requirements," Federal Register, Volume 60, No. 243, dated December 19,1995.

I1. 10 CFR 50.61," Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events," (PTS Rule) May 15,1991.

12. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.99, " Radiation Embrittlement of Reactor Vessel Materials," Revision 2, May 1988.

I References (continued) l L

BYRON - UNIT 2

{

)

PRESSURE AND TEMPERATURE LIMITS REPORT

13. Comed Calculation BRW-96-9071/BYR 96-294," Channel Accuracy for Power Operated

_ Relief Valve (PORV) Setpoints and Wide Range RCS Temperature Indication (Unit 2

' Original Steam Generators)" Revision O.

14. ; WCAP-15178, Revision 0," Byron' Unit 2 Heatup and Cooldown Limit Curves for

' Normal Operation," T, J. Laubham, et al., June 1999.

15. Westinghouse Letter to Comed, CAE-97-202," Byron Unit 2 COMS Serpoints for 12

- EFPY," October 23,1997.

i

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

17. NRC Letter from R. A. Capra, NRR, to O. D. Kingsley, Commonwealth Edison Co.,

" Byron Station, Units 1 and 2, and Braidwood Station, Units 1 and 2, Acceptance for Referencing of Pressure Temperature Limits Report (TAC Numbers M98799, M98800, 3

M98801, and M98802)," January 21,1998.

18. WCAP-15176, Revision 0, " Analysis of Capsule X from Commonwealth Edison

. Company Byron Unit 2 Reactor Vessel Radiation Surveillance Program," T. J. Laubham, et al., March 1999.

19. WCAP-15183, Revision 0," Commonwealth Edison Company Byron Unit 1 Surveillance Program Credibility Evaluation," T. J. Laubham, et al., June 1999.

19 i ~

Attachment C1 ByTon Station CMTRs for Weld Wire Heats p.\\99byttrs\\990095. doc

i,.s o.n m n,.....

,/..

.* n.,

l Po.,e c,a.. A,.

u.

s.m.

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" V/2

^

itst no.

RECORD OF FILLER WIRE OUALIFICATION TEST 1

c

• } l101 r'

(

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i i

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TEST wo.

Y f u'. I r.

at :>..tts 1=rf wraf Taraturnt uv o s..w

• r H e 1100 - 11)T r 1or >v rirs. turnace coclec to6b~ Fat 10F/Hr.

l TENSILE PROPERTIES ULT ineATE VIELD PT.

T.E L ONG.

% RED. CW TEST ho.

HEAT TREATE NT 572. PSI PSI 8.2s0FF5ET AREA WF W

ANFr, oj,4 pv s,7vu

d. )

u.v J

l CHARPY V-NOTCH IMPACT TEST 240FT./LB. ENERGY LOAD Q

HEAT TREATE NT TEST NO.

FT./LD5.1 L AT. E R P.1 % SHEAR TEST No.

FT./Le5.l L AT.E M P. I % SMtan WW SEE AT' ACHED SHIET l

\\

\\

GUIDED BEND TESTS asATEnAL a p evinw at appmovtor nt>ECTEn-l FACE a007 SIDE NAW5MIPS. 250<1500 1 Y'*

A5esE. Cotes'L NUCLEAR STEAas M WERAftut5 ElC*o OR uArno F 4M5teNi Age ALy% gt.

• L L8' ' t '. i WE HEREBY CERTIFY THAT THE ABOVE MATERI AL HAS BEEN nac ioca s,wi c E s aus na t io.

TESTED IN ACCORDANCE WITH THE ABOVE LISTED SPECIFICA.

ACCEPTABLE TlON AND l$ IN CONF 0hMANCE wlTH ALL REQUIREMENTS.

• GROOVE WELD JANUARY 8, 1crT5

.r E Folio No. SC4-220 g r g oATE [

550-063 iGNEo FLUX FOLIO NO.

• ~ -

W. VEW ON INSPECTION AGENCY woRns AEM INSPECTOR CONTRACT No.

ll 'I

3_'

4.o

.?i bY CFAtLFY-V TRANSITION CtCVES 1-7-75 MTE -

COW RACT

.7 h72 CCMPONZNT CUS'!UKER HEAT NO.

' SERIAL NO.

TEST IDCATION 14 TEST SPEC.

k

\\

uT=

m,.

EXP.W310N 7

SHEAR (OF)

FT.

LBS.

f100 100

+200 75 75 77

.06h

.072

.073 100

+150 70 67 65

.057

.057

.056 M

os 95

+110' 59 58 So

.053

.05h

.053 85 ~

oO oo

+ 70 50 53

'57

.Okh

.050

.0ho 75 85 80 g

+ 40 L7 h3 hk

.0ho

.026

.0to 60 55 60

+ 10 28

?7 27

.051 020

.071 75 10 25 10 12 20-

. 012

.015l.01o' 10 15 20

-60

'8 6-12

.006

.006

.012 5

5 15

-100 2-4 h

.005

.007

.007 0

0 0

• • **8 n t,

REV1Sc.O a/o/,.

l j/

e......s.....

g n.

.....a WF 336

'O.':0 0F F ' L L ER W l RE Oll A L I F I Ct.1100 TEST ttst no.

• • 7..l"".

j 12 W-WQIT32 LUfDE' MnMoNi~(E0W Cu6P)

QND.,E3.

8.0. i.Q*I_JS7 3

+

4 a 7 0 0 *>

/32" AUTO SU Q RC

.s V:tia e m, e............

i,.....

l 33 12 450-500-600 642040AK

..n vi.

A..

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l. i.

l t.e i.

l

-l

=,

i, s.

e s.

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Lau nu.-

.an 19769

.000'

.12

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.010

.011

.52

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, y l Sn Ti Pb S

y 1 As i I;-.

i l

1 l

i i

l n.p.l.002

. 01cl.002

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I 1

l CHOP v.6 I C.HT S l [IQ 44#* A I T pf a f att N r TLST =0.

T t an'. F F, )

l Et51.Lt$

1100-1150 F.

FOR 50 HRS.

WF 336

-30 F F

! l - 10 t1 0

FURNACE COOLED TO 600 F CENTER

-200F NF.NF i (

/

AT 10 F/HR.

l l

1 TENSILE PROPERTIES i

ULT l heaTE YltLD PT.

• *

• oneG.

'n RtD. OF TEST sec.

HE AT TatAThENT STR.. PS I PS I a#tA WE )_36 (ABCVE) 80.50p 63.000 77.5 61.0 CHARPY V. NOTCH IMPACT TEST 2J0FT./LB. ENERGY LOAD if wcAT Tu u sase.t Tts? me.

rv./Les,jtat.txP. e.sutAnlTestno.

r T. / L'es. LAT.Eur.{t.swcan (ABOVE)

UF'496 56 l _049 79 M74 A 3 67

.049 05 SURFACE 49 8.047 1 70

/t CENTER) 71

.053 90 7

e

+10 F i 40 l_069 i An L

.s.7n0F/ 51

_044 l An 0

( RT N

\\

/'

\\ NDT is -300F T

WF336 72

.064 _100 Dr3'36 59 I.053 80

\\

/

cmm i 79 I 076 61 nn ir-vn v u l go

.0 51_

RO e

A i+2000Fi 81 i.075'100 1 +10cv I T)0

.054 85 m

i GUIDED SEND TESTS waTrasat arenovat l appmovco e t.st cT to Fact ROOT stet mAYSMik$. 250/1500 Asut. conea L =

Aurzftn

-a e,tto.

a.4, F $ hnitNi aapat v.. g.,

et: fil 6tl.HY frRTif Y THAT THE ADOVF MATERI AL HAS 8.

g ramv es ti, i t t. :

TESTED IN AC(0@ ant *t. EITH TbE APOVE LtSTED SPECl w.

" d i' W" *

• H L f**M'a* Tina TION ANo IE lh LONt0N EhCE nlTH ALL REQUIREME T

21. j'Qjg N.:s

" gt

..., e.,,.,,

r.:

mni 3 i

etRE rotso No.

464-221 pATE JUI.Y 31. 1973 g

Flux roLlo No.

559-063

' M.

sioNro e,

't0014 #4[h works MT. VERNON 1NSPECTIoN AGENCY T

2 v

Ng CONTRACT NO.

INSPECTOR

.ss au -

II -9

f " l,._... JOB IDDITITICATION-' ' coDr E3o5 WES7DIGHOUSE SURVEILLANCE Mho.00%.51

/

~

DATE' November 5. 1975' TEST IDCATION HAZ 1/k T LATERAL

%5 TRAC"UR:.

EP.

FT. LB.

F.

' V/J,UES-AVG.

' IX'PANSION

• 200

- 155 lo5 160 170

.080

.087

.08h 100 100 100

[

'+100.

172.

150

- 260 161

.086

.085

.070 100 100 100 e

- _ _ _ '70-151 135-17o 152

.076

.078

.08h

~85 85 100

+ 40 131 155 lh3 1h3

.073

.081

.070 75 85 75

+ 10 153 112 127 131

.076

.065

.06h 75 65 70

- 20 115 115 106 112

.063

.065

.062 65 60 55

- 50 90 94 yr 9k

.065

.051

.057 70 50 50

-100 ho' 25 30 32

.026

.015

.019

-35 30 30

-DROI WEIGIT: NB NE 9 -30F and NB,B@hF yD1's - tl 0 p9r : ri.,W t

t.g1 l'S-db'>*46 e.4i.

_.... h..=.:.:

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4 e.a s e matr=s <

g,e or m...m::e Joe No.

4 0- 6 -51 custm Wenth-hmin,

-"--rimnet Niv re for the purypillance Weld Assembly -

1/kT Meat Afreeted Zone ev J. B.-Toon

• December 12, lo75 II-10

r-1

/

LJ.

Y CHARPY-V TRANSITION CURVES _

CONTRACT SPL 136 DATE MAY 6.

1974 COMPONENT CUSIVMER HEAT NO.

442002/8873 SERIAL NO.

3 ~ER TEST I4 CATION TEST SPEC.

E ATTACHMENT TO WF336 IATERAL TEMP.

(OF) 7T. LBS.

EXPANSION 7.

SHEAR

+200 72 vo 81

.06h

.076

.074 100 100 1ec l

+150

.;'100 75 79 78

.062

.065

.064 100 100 100

+70 66 78 71

.057

.06R

.060 00 os 00

+40 71 55 56

.058

.037

.041 90 65 65 41n 43 48 43

.029

.01s

.034 45 ss 50

-20 43 47 41

.028

.032

.029 45 45 45

-50 37 28 40

_022

.017

.025 25 20 25

-100 18 23 17

.012

.013

.009 5

5 5

i E

1 I.

1.

Tes-

& des MAY 61974 l

I il -Il

1

..s T.E*.*IS ED:

1-6-76

)

,, a eost a ct =t nai n ou cu ng l,

ut. vtamou, iso Awa g

tis 1wo.W Sol CORC QE PI L LER WI RE QU All F IC AT 10tJ TEST I

~ - m i.. c 64 -

mt.,,,,,,,,,,,,,,,

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vott>-

shavtL >. i t U ieru, ivet os u.ha hi a m ata u..a u

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L50-800 67=. ACE cp 3h-37 1h Ac t til Mit AI.

.V, A l Y ' i o ' i l

h

• t Lok NU.

s* s i n C

v4 ng un

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5 MJ.

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I.00

1. ~. 2 j.021 l 20hlo

$1F

. 0 "i 0

.00

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.012

.01c l.ad l.026i.00: 1 i

l l

l l

I i

l l

DROP WEIGHT 5 i4( A1 TREAfoENt TEST No.

T I MP. t r )

l at u 5 l

Th07 1100 - 1150 F for 60 FDS FUR %Cr c00Erp m i'r Sol M

L-re.. i 0

600 F @ 15 F/1?R.

10 7 up I

ho*F

!!3.T l

TENSILE PROPERTIES ULTIMATE viELD PT.

• ELONG.

T. RfD. OF TEST No.

HEAT T RE ATME NT

• STA. PSI PSI

8. 2 ?.oF F 5 g 7 AREA

_ WF 501 ABOVE 93,750 o8,000 25.0 62.0

,w.

l CHARPY V. NOTCH IMPACT TEST 240FT./LB. ENERGY LOAD HEAT TRE ATh(NT TES T No.

FT./L85.; L AT.E XP.

% bMEAR TEST NO.

FT.eL85.

LAT.EXP.

% SHEa4 k'F501 51

.037 80 WF 501 71

.057 100 ABOVE 1/4T 50

.0 37 75 1/hT c;9

.057 100

+70 6h 051 o5

+150 6o

.057 100 wr 501 70

.0s6 100 W 501 78

.062 100 l

1/hT 63

.048 o0 1/4T 79

.065 100

+100 73

.058 100

+200 74

.001 100 GUIDED BEND TESTS l

MATE #iAL A p pR ov at a r P movtti R E.n c i t FACE ROOT SIDE Nav5 HIPS. 250/1500 8 X*

A5ME. Comed *L NUCLEAR STEAM GENCAATORS clCCD OR MACR 0 APR181$}$

F#s5n#t ANALv5:5.

Groove mLLD TEST riE HEREBY CERTIFY THAT THE ABOVE MATERIAL HA B e.

b RAotoGRAPwsc taauiNAftou TESTED IN ACCORDANCE wlTH THE ABOVE LISTED S ACC EPT/312. Ww Cu - P UP TO E-TION AND 15 IN CONFOhhiANCE WITH ALL REQUIREMENTS.

• GICOVE WELD LOW CU IDWP CODE 1974 WIUILR /WDEHLA 1

TEBRUARY 13, 1975 dl;E FOLIO NO.

DATE klGNED O

FLUX FOLIO NO.

JAN

' 10 FEB 131975 8 -U AGENCY CORKS M.

H G

b1 Ub CONTRACT NO.

IRSPECTOR

/- r

. /, /

/Yo M //X / We Il-t3

Attachment C2 Byron Station CMTRs for Weld Wire Heats p 39byttrs\\990095 doc

E

.E

./

1,., r.r...

,i

. e, t. 5 a

.5 e

t "r~~---

' m  ;

RECORD OF FILLER WIRT Oll t.L 1 F I f./.1 16.

i f .

r,aun t a i

.,...,:. av. ;r I

l i

12-2-WOI-32 I~ 1.1141)L Mi ' k 5' I l r.Q.P )

Aulu duBTr C

~~TTiNiTG.Vif T6T i:uBL

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I i

5/32" i

. m' rn.....

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4

.c.

3 1 a.i,

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l I

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=

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0 8

l FURNACE COOLED TO 6000F 1

I i

AT 10 F/HR.

g I

i i

g I

I I

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r or e SttStNo.

l WF447 ABOVE h)a250 64.500l 27.5 6 '). O I

fli A R PY V. N(h t li imp f t 5 TEST 240ft

<t e (fiRGv LOAD

• 5 46 L at. t a v. l

• 1,e.t s a ft$1 %

rT 'tBE Laf.Eup lti e 6..

p.7 at 1 st a %s s.1 1a$1 pa i ABOVE

__,_, _,.,, [Sl.fi ATTACHEIL.SIIEF"I I

i g

.. _.._]._. l 1

l I

l I

i l

a.a 'e i. s a t appoovat a**povtol atstetti GtilDE D 8[N(i T[ ",11-raC[

woof

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X a

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usrao ou warno 1

I i 4 *%uus a aty..:.

' " " " * * *' L o O

WI HrRLOY CLR1114 Ti4A1 l e t, Alinvr MA t t i t al siA. til l e,

= au n d',* ae u n taawewatins TESTED IN ACCORDANf( ellH litt Af toVL i ikit ti %I.s e* t i f( A j

T10N AND I! IN ( ON6 Ol*AN( 4

• I let Al i HI Otilwi un t.t t I

i.

8 197.4 Al'

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AurmsT s. 1974 V

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l.. bbi IliIQ t c,,,, n __ _,

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f,

s,, - d l

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..9 a e iiw a g,t,p(( "..,.,.

a.

g Q

l ll IS'

tJ g'

tt CH.'."JY-V T!1.';EITIO!; (UPNES C0hTRACT DATE CUSTOMER CO.;PO!;El T SERIAL f;0.

WF447

'riE.'.T 1:3.

I TEST SPEC.

TOST IDCATION TEMP.

LATEP).L

[

g (OF)

FT.

LDS.

EXPl.I:3IO!!

l SHEAR l100 l

+150 81 83 77

.071

.074

.068 100 100 l 9 5._

+100 58 75 68 052

,n65

.056 95 95

+70 51 63 52

.046

.055 043 80 85 80 t

r

+40 s1 4R 41 041

.018

.014 65 65 60 (3d

+10 La 9n A1 n1A n1s 01; ss 7s sc

-20 36 34

.027

.022 25 25 l

-50 14 16

.009

.012 5

5

-100 75 17

.016

.n09 0

0 I

I

,0 h w

i m

TESTED BY

?

Il-J L,

1 L

/

p

~

O sabcock a wiicox

...e, ce e,.e.m c,..

H.gnway 69 West. Mt. Veere. ino t.7620 Telephone (812) 839-60?!

h

'{])f.'i) M n q m]~

-o~~-

/bc y

.r 8

(

[Y JUL 2 21977 3,17 13, 1977

{,l gh C

a., pas,.<.f.J.f,.b.Ic PURCHASING /

WESTINGHOUSE NUCLEAR ENERCT SYSTEMS Monroeville Nuclear Center Northern Pike Road Monroeville Pennsylvania 15146 Letter No.

012-569 Attna Mr. G.A. Brenneman REF:

WNES 546-CVJ-195410 BP B&W 640-0012 (2.1.2/2.7.2) i SUBJ: Wald Cont.usables

,4

'a &,4y,. -

c.,gg,,,,,

k Attached is the revised weld consumable certification dated 7/11/77 for WF-562.

?

Yours very truly, BABCOCK & WILC0K COMPANT n_

?

K. B. Stuckay i

Project Manager Special Vessel Contracts IBS/ asst W <, :

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THE BABC0CK & W I LC 0x CondP A'o - -' */ ' '

'e s..

posta ct=t sat s o% caove MT. vta40N. thDlama n st ac. '" : M h'tCORD OF -Ti LLER WI RE OU All'F IC ATION ' TEST s

~t

• e.66* *;;p,68'?Els

-."....'a t s"I*L6 i

(

• M :

• g ;l_.g 6 4 & n.p ; d. *

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• LLY t,

.tpev46 Sa't f L aIPui I.1k--O

'12-10 f,c

$go.ce;

- = g 7 3,..

CHEMICAL ANALYSIS L

i.

t p u, i

ic..

E e.,

.'eMer I Br.re '.. i rL 10 l.11

.6L.l2.00

.09

.010

.012 I

.L'

^?

?!24 22 %L '

l Groove, l.06f-

.10

.65~11.LL

.h7

. 0'.2

.010 I. LEI

.0-M ? '. *

"-E t - h u. E l.M7-.000 l A [1.he ho

. 0'. k 2. 012 '

L'2 l

• c

.'Ok.-

• • L

~

'I~

Cb. 00h Gn..One B.

.001

• b.

.M1 As..006 Zr..002 W.

. 0 "'

-'a ttst no.

f t ur t'r ats.. s s

wt a t f at atuc ht 1100-1150F FOR 50 WRS n'*JWE COOLED TO WP 562 40T

??.*:

i 600? 6 15F/MF 1/k T l

l l

l

\\ /

I i

/

l l

l TENSILE PROPERTIES ftST h0.

Mtat totategNT -

p$g a. 2 oF F 5 f t

' a(t a W 562

-ABOVE 81.000 6h.000 oA.6 Ao.-

CHARPY V. NOTCH IMPACT TEST 240FT./LB. ENERGY LOAD Mt at tatatutNT ttSt NO.

ft./L95.

L at. t llP. % SNtan itSt No.

ft,/Le5.

Lat.(IP,

9, t eel &&

APOYr FT'r A**A*% PS t't.fT'r'*

1

[

GUIDED SEND TESTS uattniat ape =ovat ! aereovtoi nt st et 8aCI

1. 007 510t Nav5MlPS 250/t500 1 asut comes L nucLtan 5ttau GENtnatoR5 i

utta0 on uaCAO Felsunt amatvses, g

Geoovt atto vtSt

• E HEREBY CERTIFY THAT THE AB0VE MATERIAL HAS BEEN

)

amo80casawic Esausmat:04 TESTED IN ACCORDANCE *lTH THE ABOVE LISTED SPECIFICA-TION AND 15 t e. CONFOWANCE WITH ALL REQUIREMENTS.

AC0rin GLv

• W to ASME Code lo7h Winter Addenda (IAv Cu, ~ow P) 5& -201

~ M611p3 DATE Juiv 18,1crt5 3

AIRE FOLIO NO.

EN M

• C M1 1

A/.{ SIGNED Flux FOLlo NO MUM WORKS Mt. Vemor.

N )V N EtS N acENCv J

Folio Items 9

GNSPECTOR COATRACT NO.

f

})-11

THE BAB 00k & *iLC0x CC*/F L 6

S

• c2 cue A

l 3

soara ctatna'.lo?.

WT vfE40h i hD i an.4 1ts, no. ' 1-RECORD OF F t LLER WIRE OU ALI F 1 CATION TEST

.e..

t auric

• % upgpm.nc-

<..u.. g,,a, %...,

i

2 2 ',: : :2 s y A c.1-t-,, r,,,,_, 7. e n s. w e * * * '-::

l Auto Nt Apr ru1L

_Jnde s pc x r m.:e, -. : 7 i

gg,.

inavt6 >.cic teru, t,*t or c v.

ni a.,, u. o

. c..

l.a '

I 32-35 12-16 Ar 500 -e-sur ::.

CHEMIC/.L ANALYSIS I

l ua.. l Si fa l5 l W:

l l

r ce si eae 6

• it r.L

-'.-i

.60 l 1. col.09 l.007 l.016 I.LE I T'

• !e ndor bare '41r=

.1*

1a1 i

i P=o:=

hreev-

--I

.to

.=t i 1.Lil.uo

. nu

- - =

I l

l l

l I

I DROP WEIGHTS TEST No.

Ttus W' l

sts l

e tat tataturwt 1100-1150~p for gn h L kTM L

• • • '?

I

!.'t J,':

1 Purnace Cooled to 600'y e Ic'ythe 4

i l

I l

I TENSlLE PROPERTIES ULT IM4TE vitLD PT t.t L ONG.

s att ter ftST NO.

wt af Tatain(NT Sta. PSI P51 6 2 %0F F 5 t f aHa l

Pt.nnn k7 cm l o r:.

/.i. n f

+

l WAlb g erpm l

A CHARPY V. NOTCH IMPACT TEST 240FT./LB ENERGT LOAD w

Mt af Tat A1h(N1 TES T NO.

rf./L85. LAT.EXP.' s SMEAa TEST MC.

' F T. /L 85.

La,1 rmP '... a n.

ABOVE SEE ATTACKED stu. r i

l i

I uattaint apenovat la**nowtolatste tt j

GUiOED BEND TESTS FACE acof Scot hav5MIPS 250/1900 1 Ye A5ut. Couu'L MUCLtaa I

Strau Cthta A70a5 utta0 on uaCa0 IIS$Ust shalv$tl.

caoovt atLC' TEST

• E HEREBY CERTIFY THAT THE ABOVE MATERIAL HAS BEEN "aDioGaa*uit taaui=Atioh A0 "I'*

TESTED IN ACCOROANCE S4TH THE AB0VE LISTED SPECIFICA.

CONF 0kMANCE WITH ALL REQUIREMENTS.

.t? *Ty1-~O AFiG Ceh 4pp, o TION AND If IN l

, lo7" Adde.dt omi

.m

'/'L

?

DATE M'e.' 4 Ic7A

)

j wtRE FOLIO NO.

[

[

M at C P-061 SIGNED Flux FOLIO NO, woRx5 m. vernon INSPECTION AGENCY R. NOBLO*

INSPECTOR CONTRACT NO.

11~ 10

e b

IMPACT TEST REULTS JOB IDENTIFICATION iTF61h DATE Januarr 6.1076 TEST LOCATION 1/LT I

'Q(F.I 1P.

F*. LB.

M

• 3 NUM VALUES AVG.

Td5WcW2W EL's i

250 77 80 78 71 76 72 100 100 100

-loo 77 68 71 70 61 65 loo 100 100

+70 72 71 6o 62 66 60 cm o5 os

• Lo 58 58 So sh 53 56 80 75 80

_ 10 ho 52 51 90 h2 ho 50 55 55 JAN 7 1976 m

7 56

1. ev -

O R. NOBLES E- " A*""

in U

1 11 51

Attachment D J

B Ton Station j

3 1

i i

"The B&W Owners Group Reactor Vessel Working Group Byron and Braidwood RV Weld Chemistry Initial RT a 51-5002206-00," Framatome

-Technologies

{

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