ULNRC-06350, Pressure and Temperature Limits Report

Pressure and Temperature Limits Report

ML17068A381
Person / Time
Site:	Callaway
Issue date:	03/09/2017
From:	Wink R Ameren Missouri, Union Electric Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-06350
Download: ML17068A381 (28)

Text

Aif18t9fl Callaway Plant MISSOURI March 9, 2017 ULNRC-06350 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-000 1 10 CFR 50.36 Ladies and Gentlemen:

DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 UNION ELECTRIC CO.

RENEWED OPERATING LICENSE NPF-30 PRESSURE AND TEMPERATURE LIMITS REPORT Enclosed is the Callaway Plant Pressure and Temperature Limits Report (PTLR), Revision 7. This revision modifies the RC$ heatup and cooldown curves with values that are applicable up to 35 Effective Full Power Years (EFPY). Revision 7 is considered an interim revision as data is currently being analyzed that will support development of curves that are applicable up to 54 EFPY and include consideration of vessel extended belt line components. Revision 7 was developed in accordance with the NRC approved methodology in WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoint and RCS Heatup and Cooldown Limit Curves, February, 2004. This report is provided to the NRC Staff for information in accordance with the requirements of Technical Specification 5.6.6.c.

This letter does not contain new commitments.

If there are any questions, please contact Jim Nurrenbem at 314-225-1908.

Sincere,

/ Roger C. Wink DRB/tlw

/ Manager, Regulatory Affairs Enclosure STARS Milance

ULNRC-063 50 March 9,2017 Page 2 cc: Mr. Kriss M. Kennedy Regional Administrator U. S. Nuclear Regulatory Commission Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road

$teedman, MO 65077 Mr. L. John Kios Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O8H4 Washington, DC 20555-0001

ULNRC-063 50 March 9, 2017 Page 3 Index and send hardcopy to QA File A160.0761 Hardcopy:

Certrec Corporation 6100 Western Place, Suite 1050 fort Worth, TX 76107 (Certrec receives ALL attachments as long as they are non-safeguards and may be publicly disclosed.)

Electronic distribution for the following can be made via Responses and Reports ULNRC Distribution:

F. M. Diya I. E. Herrmann B. L. Cox R. C. Wink T. B. Elwood Corporate Communications NSRB Secretary Mr. Greg Voss, REP Manager (SEMA)*

STARS Regulatory Affairs Mr. Jay Silberg (Pillsbury Winthrop Shaw Pittman LLP)

Mr. Steve feeler (DNR)*

Mr. Robert Stout (DNR)*

Enclosure to ULNRC06350 CALLAWAY PLANT PRESSURE AND TEMPERATURE LIMITS REPORT Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents 1 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) 3 2 Operating Limits 3 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 3 2.2 Cold Overpressure Mitigation System (COMS) Setpoints (LCO 3.4.12) 3 3 Reactor Vessel Material Surveillance Program 10 4 Reactor Vessel Surveillance Data Credibility 10 5 Supplemental Data Tables 16 6 References 17 Callaway Energy Center Page 1 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT LIST OF FIGURES FIGURE PAGE 2.1-1 Callaway Plant Reactor Coolant System Heatup Limitations (Heatup 4 Rates of 60 and 100 °f/hr) Applicable to 35 EFPY(with Margins for Instrumentation Errors) 2.1-2 Callaway Plant Reactor Coolant System Cooldown Limitations 6 (Cooldown Rates of 0, 20, 40, 60 and 100 °F/hr) Applicable to 35 EFPY (with Margins for Instrumentation Errors) 2.2-1 Maximum Allowed PORV Setpoint for the Cold Overpressure $

Mitigation System LIST OF TABLES TABLE PAGE 2.1-1 Callaway Plant Heatup Limits at 35 EfPY with Margins for 5 Instrumentation Errors 2.1-2 Callaway Plant Cooldown Limits at 35 EFPY with Margins for 7 Instrumentation Errors 2.2-1 Callaway Plant COMS Maximum Allowable PORV Setpoints at 35 9 EFPY 4.0-1 Callaway Plant Surveillance Capsule Data 13 4.0-2 Callaway Plant Lower Shell Plate R270$-1 14 4.0-3 Callaway Plant Surveillance Weld Metal 15 5.0-1 Comparison of Callaway Unit 1 Surveillance Material 30 ft-lb 18 Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 5.0-2 Calculation of Chemistry Factors Using Surveillance Capsule Data 19 5.0-3 Reactor Vessel Beltline Material Unirradiated Toughness Properties 20 5.0-4 Fluence (10 n/cm2, E>l.0 MeV) on the Pressure Vessel Clad/Base 21 Metal Interface for Callaway Plant 5.0-5 Summary of Adjusted Reference Temperature (ART) Values at the 1/4T 22 and 3/4T Locations for 35 EFPY 5.0-6 Calculation of Adjusted Reference Temperature Values at 35 EFPY for 23 the Limiting Callaway Plant Reactor Vessel Material (Lower Shell Plate R270$-1) 5.0-7 RTpTS Calculations for Callaway Plant Beltline Region Materials at 35 24 EFPY Callaway Energy Center Page 2 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

This PTLR for Callaway Plant has been prepared in accordance with the requirements of Technical Specification (IS) 5.6.6. The TSs addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (PIT) Limits LCO 3.4.12 Cold Overpressure Mitigation System (COMS) 2 Operating Limits The parameter limits for the Specifications listed in Section 1.0 are presented in the following subsections. The limits were developed in accordance with the NRC-approved methodology specified in Specification 5.6.6 (Ref. 1). NRC approval of this methodology was received in a Safety Evaluation Report dated February 27, 2004 from NRC to Westinghouse (TAC No. MB5754). The three provisions listed for acceptability of the methodology are met by this report and WCAP- 14040-A, Revision 4, which describes the employed methodology.

This report meets the requirements of GL 96-03 Attachment 1, provision 2.

The revised PIT Limit curves account for a requirement of 10 CFR 50, Appendix G that the temperature of the closure head flange and vessel flange regions must be at least 120 °F higher than the limiting RINDI for these regions when the pressure exceeds 20% of the preservice hydrostatic test pressure (3106 psig).

2.1 RCS Pressure and Temperature (PIT) Limits (LCO 3.4.3) 2.1.1 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 100 °F in any 1 -hour period.

b. A maximum cooldowti of 100 °F in any 1-hour period.

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

2.1.2 The RCS PIT limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 and 2.1-2.

2.2 Cold Overpressure Mitigation System (COMS) Setpoints (LCO 3.4.12)

The pressurizer power-operated relief valves (PORVs) shall each have lift settings in accordance with Figure 2.2-1. The (COMS) arming temperature is 275 °F. These lift setpoints have been developed using the NRC approved methodologies specified in Technical Specification 5.6.6.

The maximum allowed PORV setpoint for COMS is derived by analysis which models the performance of the COMS assuming limiting mass and heat input transients. Operation with a PORV setpoint less than or equal to the maximum setpoint ensures that Appendix 0 criteria will not be violated with consideration for: (1) pressure and temperature instrumentation uncertainties, (2) single failure of one PORV, and (3) effects of reactor coolant pump operation.

To ensure mass and heat input transients more severe than those assumed cannot occur, Technical Specifications place limitations on the number of safety injection pumps and centrifugal charging pumps that are capable of injecting, unisolating accumulators, and starting reactor coolant pumps during the appropriate COMS MODES. These limitations are outlined in IS LCO 3.4.6, LCO 3.4.7, and LCO 3.4.12.

Callaway Energy Center Page 3 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTIES BASIS LIMITING MATERIAL: Lower Shell Plate R270$-l LIMITING ART VALUES AT 35 EFPY: 1/4T, 130°f

%T, 117°f 2500 Leak Test Limit 2250 2000 Unacceptable Acceptable Operation Operation 1750 Critical Limit 60 °F/Hr 1500 Heatup Rate 60 bE/Hr 0

U 1250 In a)

Critical Limit 100 °F/Hr I

Heatup Rate 100°F/Hr 1000 U

750 Criticality Limit based on inservice 500 hydrostatic test temperature (200 0F) for the service period up to 35 EFPY.

250 0

The lowerlimit for RCS pressure is -14.7 psig

-250 0 100 200 300 400 500 Moderator Temperature f°F)

Figure 2.1-1 Callaway Plant Reactor Coolant System Heatup Limitations (Heatup Rates of 60 °f and 100 °f/hr). Applicable for 35 EFPY (With Margins for Instrumentation Errors). Includes vessel flange requirements of 170 °f and 561 psig per 10 CfR 50, Appendix G.

Boltup temperature includes 10 °F instrument uncertainty Callaway Energy Center Page 4 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1-1 Callaway Plant Heatup Limits at 35 EfPY with Margins for Instrumentation Errors Criticality 100 °f/hr Criticality 60 °f/hr Heatup Heatup Limit T p T p T p 1 p

(°F) (psig) (°F) (psig) (°f) (psig) (of) (psig) 70 -14.7 202 -14.7 70 -14.7 202 14.7 70 561 202 561 70 561 202 561 75 561 202 561 75 561 02 561 80 561 202 561 80 561 202 561 85 561 202 561 85 561 202 561 90 561 202 561 90 561 02 561 95 561 202 561 95 561 02 561 100 561 202 561 100 561 02 561 105 561 202 561 105 561 02 561 110 561 202 561 110 561 02 561 115 561 02 561 115 561 02 561 120 561 02 561 120 561 02 561 125 561 02 561 125 561 02 561 130 561 02 561 130 561 02 561 135 561 02 561 135 561 02 561 140 561 02 561 140 561 02 561 145 561 02 561 145 561 02 561 150 561 02 561 150 561 02 561 155 561 02 561 155 561 02 561 160 561 05 561 160 561 205 561 165 561 10 561 165 561 210 561 170 561 10 956 170 561 210 752 170 956 15 1007 170 752 215 784 175 1007 20 1064 175 784 220 820 80 064 25 1127 180 820 225 860 85 1127 30 1196 185 860 230 904 90 196 35 1272 190 904 235 954 95 272 40 1357 195 954 240 009 200 357 245 1450 200 1009 245 069 205 450 250 1553 205 1069 250 137 210 553 255 1667 210 1137 255 211 215 1667 260 1793 215 1211 260 293 120 1793 165 1932 20 1293 265 384 225 1932 170 085 25 1384 170 485 230 2085 175 254 30 1485 175 596 235 254 180 440 35 1596 280 1718 140 440 240 1718 285 1853 45 1853 190 002 50 002 195 166 255 166 300 348 260 348 Temperature (°f) 185 202 Leak Test Limit Pressure (psig) 2000 2485 Callaway Energy Center Page 5 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE ANI) TEMPERATURE LIMITS REPORT MATERIAL PROPERTIES BASIS LIMITING MATERIAL: Lower Shell Plate R2708-l LIMITING ART VALUES AT 35 EfPY: 1/4T, 130°F

%T, 117°F 2500 2250 2000 1750 Unacceptable Operation 1500 bO 1250 C)

I 0 Cooldown Rates f°F/Hr):

1000 Steady-state

-20 U -40 (C -60 U 750

-100 500 Boltup Temperature 70 F 250 0

The lower limit forRCSpressureis-14.7 pg

-250 0 100 200 300 400 500 Moderator Temperature (F)

Figure 2.1-2 Callaway Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100 °F/br). Applicable for 35 EFPY (With Margins for Instrumentation Errors). Includes vessel flange requirements of 170 °F and 561 psig per 10 CFR 50, Appendix G.

Boltup temperature includes 10 °F instrument uncertainty.

Callaway Energy Center Page 6 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1-2 Callaway Plant Cooldown Limits at 35 EFPY with Margins for Instrumentation Errors Steady State 20 °F/hr 40 °F/hr 60 °F/hr 100 °FThr T p T p 1 p T p T p

(°f) (psig) (°F) (psig) (°f) (psig) (°F) (psig) (°F) (psig) 70 -14.7 70 -14.7 70 -14.7 70 -14.7 70 -14.7 70 561 70 561 70 555 70 511 70 424 75 561 75 561 75 561 75 524 75 439 80 561 80 561 80 561 80 538 80 456 85 561 85 561 85 561 85 553 85 475 90 561 90 561 90 561 90 561 90 496 95 561 95 561 95 561 95 561 95 519 100 561 100 561 100 561 100 561 100 545 105 561 105 561 105 561 105 561 105 561 110 561 110 561 110 561 110 561 110 561 115 561 115 561 115 561 115 561 115 561 120 561 120 561 120 561 120 561 120 561 125 561 125 561 125 561 125 561 125 561 130 561 130 561 130 561 130 561 130 561 135 561 135 561 135 - 561 135 561 135 561 140 561 140 561 140 561 140 561 140 561 145 561 145 561 145 561 145 561 145 561 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 170 561 170 561 170 561 170 561 170 561 170 1239 170 1239 170 1239 170 1239 170 1239 175 1312 175 1312 175 1312 175 1312 175 1312 180 1392 180 1392 180 1392 180 1392 180 1392 185 1481 185 1481 185 1481 185 1481 185 1481 190 1579 190 1579 190 1579 190 1579 190 1579 195 1688 195 1688 195 1688 195 1688 195 1688 200 1808 200 1808 200 1808 200 1808 200 1808 205 1940 205 1940 205 1940 205 1940 205 1940 210 2087 210 2087 210 2087 210 2087 210 2087 215 2248 215 2246 215 2248 215 2248 215 2248 220 2427 220 2427 220 2427 220 2427 220 2427 221.5 2485 221.5 2485 221.5 2485 221.5 2485 221.5 2485 Callaway Energy Center Page 7 of 24 Revision 7

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Enclosure to ULNRC06350 PRESSURE AD TEMPERATURE LIMITS REPORT Table 2.2-1 Callaway Plant COMS Maximum Allowable PORV Setpoints at 35 EFPY Maximum Allowable Function Generator Setpoints (Breakpoints)

Breakpoint Temperature High Setpoint Low Setpoint Number RCS (°F) (psig) (psig) 1 70 477 442 2 80 477 442 3 90 477 442 4 100 477 442 5 180 477 442 6 230 498 449 7 238 752 695 8 280 752 695 9 350 2335 2185 Callaway Energy Center Page 9 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT 3 Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with Appendix H to 10 CFR 50, Reactor Vessel Material Surveillance Program Requirements, and Section 5.3 of the Callaway Final Safety Analysis Report.

The surveillance capsule withdrawal schedule is presented in FSAR Table 5.3-10. The surveillance capsule reports are as follows:

1. WCAP-11374, Revision 1, June 1987, Analysis of Capsule U from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program.

2. WCAP- 12946, June 1991, Analysis of Capsule Y from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program.

3. WCAP-14895, July 1997, Analysis of Capsule V from the Union Electric Company Callaway Unit 1 Reactor Vessel Radiation Surveillance Program.

4. WCAP-15400, June 2000, Analysis of Capsule X from the AmerenUE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program.

4 Reactor Vessel Surveillance Data Credibility Regulatory Guide 1.99, Revision 2, describes general procedures acceptable to the NRC staff for calculating the effects of neutron radiation embrittlement of the low-alloy steels currently used for light-water-cooled reactor vessels. Position C.2 of Regulatory Guide 1.99, Revision 2, describes the method for calculating the adjusted reference temperature and Charpy upper-shelf energy of reactor vessel beltline materials using surveillance capsule data. The methods of Position C.2 can only be applied when two or more credible surveillance data sets become available from the reactor in question.

To date, four surveillance capsules have been removed and analyzed from the Callaway Plant reactor vessel. To use these surveillance data sets, they must be shown to be credible. In accordance with the discussion of Regulatory Guide 1.99, Revision 2, there are five requirements that must be met for the surveillance data to be judged credible.

The purpose of this evaluation is to apply the credibility requirements of the Regulatory Guide 1.99, Revision 2, to the Callaway Plant reactor vessel surveillance data and determine if the Callaway Plant surveillance data is credible.

Criterion 1:

Materials in the capsules should be those judged most likely to be controlling with regard to radiation embrifflement.

The beitline region of the reactor vessel is defined in Appendix G to 10 CFR Part 50, Fracture Toughness Requirements, as follows:

the reactor vessel (shell material including welds, heat affected zones, and plates or forgings) that directly surrounds the effective height of the active core and adjacent regions of the reactor vessel that are predicted to experience sufficient neutron radiation damage to be considered in the selection of the most limiting material with regard to radiation damage.

Callaway Energy Center Page 10 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT The Callaway Plant reactor vessel consists of the following beltline region materials:

- Intermediate shell plate R2707-1,

- Intermediate shell plate R2707-2,

- Intermediate shell plate R2707-3,

- Lower shell plate R2708-1,

- Lower shell plate R270$-2,

- Lower shell plate R2708-3, and

- Intermediate shell longitudinal weld seams, lower shell longitudinal weld seams, and an intermediate to lower shell circumferential weld seam. All vessel beltline weld seams were fabricated with weld wire heat number 90077. The intermediate to lower shell circumferential welds seam 101-171 was fabricated with flux Type 124 Lot Number 1061. The intermediate and lower shell longitudinal weld seams were fabricated with Flux Type 0091 Lot Number 0842.

The Callaway Plant surveillance program utilizes longitudinal and transverse test specimens from lower shell plate R2708- 1. The surveillance weld metal was fabricated with weld wire heat number 90077, flux Type 124, Lot Number 1061.

At the time when the surveillance program was selected it was believed that copper and phosphorus were the elements most important to embrittlement of reactor vessel steels. Since all plate material had approximately the same content of copper and phosphorus, lower shell plate R270$-1 was chosen for the surveillance program since it had the highest RTNDT and the lowest initial upper shelf energy of the plate material. In addition, the current pressurized thermal shock (PTS) evaluation shows that if surveillance data is not used, lower shell plate R2708-1 is the plate that is predicted to have the highest embrittlement rate.

Per Regulatory Guide 1.99, Revision 2, weight-percent copper and weight- percent nickel are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld. Since the surveillance weld metal was made with the same weld wire heat as all of the vessel beltline weld seams, it is representative of the limiting beitline weld metal.

Based on the above discussion, the Callaway Plant surveillance materials are those judged most likely to be controlling with regard to radiation embrittlement and the Callaway Plant surveillance program meets this criteria.

Criterion 2:

Scatter in the plots of Charpy energy versus temperature for the irradiated and unirradiated conditions should be small enough to permit the determination of the 30 ft-lb temperature and upper shelf energy unambiguously.

Plots of Charpy energy versus temperature for the unirradiated and irradiated condition are presented in WCAP- 15400, June 2000, Analysis of Capsule X from AmerenUE Callaway Unit 1 Reactor Vessel Radiation Surveillance Program.

Based on engineering judgment, the scatter in the data presented in these plots is small enough to permit the determination of the 30 ft-lb temperature and the upper shelf energy of the Callaway Plant surveillance materials unambiguously. Hence, the Callaway Plant surveillance program meets this Callaway Energy Center Page 11 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT criterion.

Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of ARTNDT values about a best-fit line drawn as described in Regulatory Position 2.1 normally should be less than 2$ °f for welds and 17 °F for base material. Even if the fluence range is large (two or more orders of magnitude),

the scatter should not exceed twice those values. Even if the data fail this criterion for use in shifi calculations, they may be credible for determining decrease in upper shelf energy if the upper shelf can be clearly determined, following the definition given in ASTM El 85-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determine a best-fit line for this data and to determine if the scatter of the ARTNDT values about this line is less than 28 °F for welds and less than 17 °f for the plate.

Following is the calculation of the best-fit line as described in Regulatory Position 2.1 of Regulatory Guide 1.99, Revision 2.

Callaway Energy Center Page 12 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-1 Callaway Plant Surveillance Capsule Data Material Capsule FFtb) FE x Capsule tRTNoT FE2 RTNDT U 0.331 0.696 0.0(e) 0.0 0.48 LowerShell y 1.27 1.07 25.15 26.91 1.14 R2708-1 V 2.52 1.25 16.45 20.56 1.56 (Longitudinal)

X 3.33 1.32 25.71 33.94 1.74 U 0.331 0.696 25.86 18.00 0.48 Lower Shell Y 1.27 1.07 46.39 49.64 1.14 Plate R2708-1 V 2.52 1.25 44.82 56.03 1.56 (Transverse) x 3.33 1.32 30.77 40.62 1.74 Sum: 8.6411 215.15 245.7 9.84 CF05=(FF* ARTNDT)/(FF2) = (245.70 °F)/(9.84) = 25.0 °F 6853(d)

U 0.331 0.696 47.70 0.48 Surveillance Y 1.27 1.07 36.92 39.50 1.14 Weld 48*21(d)

V 2.52 1.25 60.26 1.56 MateriaIt 5181(d)

X 3.33 1.32 68.39 1.74 Sum: 4.32 205.47 215.85 4.92 CFSurvWeId=(FF* ARTNDT)/(FE2) = (215.85 °F)/(4.92) = 43.9 °F Notes:

a) f=calculated fluence from capsule X dosimetry analysis results, (x 1019 nlcm2, E> 1.0 MeV).

These values were reevaluated as part of capsule X analysis. (See Section 6 of WCAP- 15400.)

b) FF = fluence factor = f(28

- O.1*log f) c) ARTNDT values are the measured 30 ft-lb shift.

d) These measured ARTNDT values do not include the adjustment ratio procedure of Reg. Guide 1.99, Revision 2, Position 2.1, since this calculation is based on the actual surveillance weld metal measured shift values and based on the copper and nickel content the ratio would be 1.

In addition, the only surveillance data available is from the Callaway Unit 1 reactor vessel; therefore, no temperature adjustment is required.

e) The actual value is -7.33, but for conservatism a value of zero is considered.

The scatter of ARTNDT values about the functional form of a best-fit line drawn as described in Regulatory Position 2.1 is presented in Table 4.0-2.

Per the 27th Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of least squares, the values of b0 and b1 are obtained by solving the normal equations nb0 + b1x = yj and box + b1x12 Xjj Callaway Energy Center Page 13 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AN]) TEMPERATURE LIMITS REPORT These equations can be re-written as follows (b0 a and b1 = b):

n n

= na + bZx1 and 1=1 i=1 n n n ZXiYI = aZxí + bx2 i=1 1=1 i=1 Lower shell plate R2708- 1:

Based on the data provided in Table 4.0-1 these equations become:

215.15 =3a +8.6411b and 245.70 = 8.641 la + 9.$4b Thus, b = 24.8405 and a = 0.1669, and the equation of the straight line which provides the best-fit line in the sense of least squares is:

Y = 24.8405 (X) + 0. 1669 The scatter in predicting a value Y corresponding to a given X value is:

e=Y-Y Table 4.0-2 Ca]laway Plant Lower Shell Plate R2708-1 Measured Best-fit Scatter of < 17 °F Base Material fF ARTNDT (30 ft-lb)(°F) ARTNDT (°F) ARTNDT (°f) (Base Metals) 0.696 0.00 17.4 -17.4 NO Lower Shell 1.07 25.15 26.75 -1.6 Yes Plate R2708-1 (Longitudinal) 1.25 16.45 31.25 -14.8 Yes 1.32 25.71 33.0 -7.29 Yes 0.696 25.86 17.4 8.46 Yes Lower Shell 1.07 46.39 26.75 19.64 NO Plate R2708-1 (Transverse) 1.25 44.82 31.25 13.57 Yes 1.32 30.77 33.0 -2.23 Yes Notes:

a) Best-fit Line Per Equation 2 of Reg. Guide 1.99, Rev. 2 Position 1.1 Callaway Energy Center Page 14 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-2 indicates that one measured plate ARTNDT value is below the lower bound hr of 17 °F by less than 1 °F, meaning the best-fit line is slightly over predicting this measured ARTNDT value. Table 4.0-2 also indicates that one measured plate ARTNDT value is above the upper bound 1 c of 17 °F by less than 3 °F. From a statistical point of view +/- 1c (17 °f) would be expected to encompass 68% of the data.

Therefore, it is still statistically acceptable to have two of the plate data points fall outside the +/- 1 y bounds. The fact that two of the measured plate ARTNDT values are outside of the 1 bound of 17 °f can be attributed to several factors, such as 1) the inherent uncertainty in Charpy test data, 2) the use of a symmetric hyperbolic tangent Charpy curve fitting program versus asymmetric tangent Charpy curve fitting program or hand drawn curves using engineering judgment, and/or 3) rounding errors.

In summary, all measured plate is within acceptable range. Therefore, the plate data meets this criteria.

Weld Metal:

Based on the data provided in Table 4.0-1 the equations become:

205.47 = 3a + 4.321b and 215.13 = 4.321a + 4.897b Thus, b = 60.9 and a = -19.3, and the equation of the straight line which provides the best-fit in the sense of the least squares is:

Y = 60.9 (X) -19.3 The scatter in predicting a value of Y corresponding to a given X value is:

E=Y-Y Table 4.0-3 Callaway Plant Surveillance Weld Metal Measured Best-Fit Scatter of < 28 °F FF ARTNDT(30 ARTNDT (°F) ARTNDT (°F) (Weld Metal) ft-lb)(°F) 0.696 68.53 30.55 37.98 NO 1.07 36.92 46.97 -10.05 Yes 1.25 48.21 54.88 -6.67 Yes 1.32 51.81 57.95 -6.14 Yes Notes:

a) Best-Fit-Line Per Equation 2 of Reg. Guide 1.99, Rev. 2 Position 1.1 Callaway Energy Center Page 15 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT One measured weld ARTNDT value is below the lower icr at 2$ °f. The fact that one of the measured weld ARTNDT values is out of icr bound of 28 °F can be attributed to several factors, such as 1) the inherent uncertainty in Charpy test data, 2) the use of a symmetric hyperbolic tangent Charpy curve fitting program versus asymmetric tangent Charpy curve fitting program or hand drawn curves using engineering judgment, and/or 3) rounding errors.

Criterion 4:

The irradiation temperature of the Charpy specimens in the capsule should match the vessel wall temperature at the cladding/base metal interface within +/- 25 of.

The capsule specimens are located in the reactor between the neutron pads and the vessel wall and are positioned opposite the center of the core. The test capsules are in baskets attached to the neutron pads.

The location of the specimens with respect to the reactor vessel beltline provides assurance that the reactor vessel wall and the specimens experience equivalent operating conditions such that the temperatures will not differ by more than 25 °f. Hence this criterion is met.

Criterion 5:

The surveillance data for the correlation monitor material in the capsule should fall within the scatter band of the data base for that material.

The Callaway Plant surveillance program does not contain correlation monitor material. Therefore, the criterion is not applicable to the Callaway Plant surveillance program.

Based on the preceding positive responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B, the Callaway Plant surveillance data is credible.

5 Supplemental Data Tables Table 5.0-1 Comparison of Callaway Plant Surveillance Material 30 fl-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions.

Table 5.0-2 Calculation of Chemistry factors Using Surveillance Capsule Data.

Table 5.0-3 Provides the unirradiated reactor vessel toughness data. The boltup temperature is also included in this Table.

Table 5.0-4 Provides a summary of the pressure vessel neutron fluence values at 35 EfPY used for the calculation of the ART values.

Table 5.0-5 Provides a summary of the adjusted reference temperature (ART) for reactor vessel beltline materials at the l/4T and 3/4T locations for 35 EfPY.

Table 5.0-6 Shows the calculation of the ART at 35 EfPY for the limiting reactor vessel material (lower shell plate R2708-1).

Table 5.0-7 Provides RTpTS values for 35 EfPY.

Callaway Energy Center Page 16 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT 6 References

1. Technical Specification 5.6.6, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR).

2. NRC letter dated March 24, 2000, Callaway Plant, Unit 1 Issuance of Amendment re: Pressure Temperature Limits Report.

3. WCAP-14040-A, Revision 4, Methodology Used to Develop Cold Overpressure Mitigating System Setpoint and RCS Heatup and Cooldown Limit Curves, May, 2004.

4. WCAP-16654-NP, Revision 0, Callaway Unit 1 Heatup and Cooldown Limit Curves for Normal Operation, November 2006.

5. Westinghouse Letter, $CP-06-66, Final LTOPS Setpoint Analysis for Increased PORV Stroke Time, November 20, 2006.

Callaway Energy Center Page 17 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-1 Comparison of Callaway Unit 1 Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Fluence Temperature Shift Decrease Materials Capsule (x 1019 nlcm2) Predicted Measured Predicted Measured (a)

(°f) (°F) (b) (%) (a) (%) (c)

Lower Shell U 0.331 30.62 o.o 14.5 0 Plate R2708-1 Y 1.27 47.08 25.15 20 6 (Longitudinal) V 2.52 55.0 16.45 23.5 0 X 3.33 52.08 25.71 25 5 U 0.331 30.62 25.86 14.5 11 Lower Shell Y 1.27 47.08 46.39 20 13 Plate R2708-1 V 2.52 55.0 44.82 23.5 3 (Transverse)

X 3.33 58.02 30.77 25 5 U 0.331 22.13 68.53 14.5 11 Y 1.27 34.02 36.92 20 14 Weld Metal V 2.52 39.75 48.21 23.5 2 X 3.33 41.98 51.81 25 8 U 0.331 -- 65.93 -- 0 Y 1.27 -- 56.38 -- 14 HAZ Metal V 2.52 56.1

-- -- 0 X 3.33 -- 42.11 -- 0 Notes:

a) Based on Regulatory Guide 1.99, Revision 2, methodology using the mean weight percent values of copper and nickel of the surveillance material.

b) Calculated using measured Charpy data plotted using CVGRAPH, Version 4.1.

c) Values are based on the definition of upper shelf energy given in ASTM E185-85.

d) Actual measured value for ARTNDT is -7.33. This physically should not occur; therefore, for conservatism a value of zero will be used.

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Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-2 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Casjle FFtb)

FF x Capsule ARTNDT(C) FE2 RTNDT U 0.331 0.696 0.0(e) 0.0 0.48 Lower Shell Y 1.27 1.07 25.15 26.91 1.14 08-V 2.52 1.25 16.45 20.56 1.56 (Longitudinal)

X 3.33 1.32 25.71 33.94 1.74 U 0.331 0.696 25.86 18.00 0.48 Lower Shell Y 1.27 1.07 46.39 49.64 1.14 Plate R2708-1 V 2.52 1.25 44.82 56.03 1.56 (Transverse) X 3.33 1.32 30.77 40.62 1.74 Sum; 245.7 9.84 CF05=(EE* RTNoT)/(EF2) = (245.70 °F)/(9.84) = 25.0 °F U 0.331 0.696 62.3& 43.40 0.48 Surveillance Y 1.27 1.07 33.60 35.95 1.14 Weld V 2.52 1.25 43.87 54.84 1.56 Material X 3.33 1.32 47.l5 62.24 1.74 Sum; 196.43 4.92 CFSurvWeId=(EE* ARTNDT)/(EF2) = (196.43 °E)/(4.92) 39.9 °E Notes; a) f= calculated fluence from capsule X dosimetry analysis results, (x 1019 nlcm2, E> 1.0 MeV). These values were reevaluated as part of capsule X analysis (See Section 6 of WCAP- 15400)

= f(O2SOl*b0 b) FF = fluence factor c) ARTNDT values are the measured 30 fl-lb shift d) The surveillance weld metal zRTNDT values have been adjusted by a ratio factor or 0.91.

e) The actual value is -7.33, but for conservatism a value of zero is considered.

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Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-3 Reactor Vessel Beitline Material Unirradiated Toughness Properties Material Cu (%) Ni (%) Initial RTNDT(a)

Description Closure Head flange 0.03 0.85 200F Vessel flange

-- 0.74 40 OF(C)

R2701-1 Intermediate Shell 0.05 0.58 40 °f Plate_R2707-1 Intermediate Shell 0.06 0.61 10 °F Plate_R2707-2 Intermediate Shell 0.06 0.62 -10 °F Plate_R2707-3 Lower Shell Plate 0.07 0.58 50 °F R2708- 1 Lower Shell Plate 0.06 0.57 10 °F R270$-2 Lower Shell Plate 0.08 0.62 20 °f R2708-3 Intermediate and Lower Shell 0.04 0.05 -60 °f Longitudinal Weld Seams (b)

Intermediate to Lower Shell 0.04 0.05 -60 °F Circumferential Weld_Seam (b)

Surveillance 0.045 0.065 Weld (b)(c) --

Notes:

a) The initial RTNDT values for the plates and welds are based on measured data (WCAP 12948).

b) All vessel beitline weld seams were fabricated with weld wire heat number 90077. The intermediate to lower shell circumferential weld seam 101-171 was fabricated with flux Type 124 Lot Number 1061. The intermediate and lower shell longitudinal weld seams were fabricated with Flux Type 0091 Lot 0842. the surveillance weld metal was fabricated with weld wire heat number 90077, Flux Type 124 Lot Number 1061. Per Regulatory Guide 1.99, Revision 2, weight percent copper and weight percent nickel are the best-estimate values for the material, which will normally be the mean of the measured values for a plate or forging or for weld samples made with the weld wire heat number that matches the critical vessel weld. The surveillance weld metal was made with the same weld wire heat as all of the vessel beltline weld seams and is therefore, representative of all of the beitline weld seams.

c) These values are used for considering flange requirements for the heatup/cooldown curves. Per the methodology given in WCAP-14040-A, the minimum boltup temperature is 70 °f.

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Enclosure to ULNRC06350 PRESSURE AN]) TEMPERATURE LIMITS REPORT Table 5.0-4 Fluence (1019 n/cm2, E>1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for Callaway Plant EFPY 0° 15° 30° 45° 12.40 0.445 0.649 0.756 0.768 16 0.565 0.822 0.956 0.964 24 0.832 1.21 1.40 1.40 32 1.10 1.59 1.85 1.83 54 1.83 2.64 3.07 3.02 Callaway Energy Center Page 21 of 24 Revision 7

Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-5 Summary of Adjusted Reference Temperature (ART) Values at the 114T and 3/4T Locations for 35 EFPY

. 35 EFPY ART(a)

Material 1/4 1 ART (°F) 3/4 1 ART (°F)

Intermediate Shell 105 87 Plate_R2707-1 Intermediate Shell 83 66 Plate_R2707-2 Intermediate Shell 63 46 Plate_R2707-3 Lower Shell Plate R2708-l 130w 1 17(b)

Using Surveillance Capsule 93 86 Data Lower Shell Plate R2708-2 83 66 Lower Shell Plate R2708-3 108 93 Intermediate & Lower Shell Longitudinal Weld Seams 1 -16 101-124A& l01-142A (90°Azimuth)

Using Surveillance Capsule 10 -2 Data Intermediate & Lower Shell Longitudinal Weld Seams 101-124B&C & 101-142B&C 1 -16 (2 10°_&_330° Azimuth)

Using Surveillance Capsule 10 -2 Data Intermediate to Lower Shell Circumferential Weld 1 -16 Seams_101-171 Using Surveillance Capsule 10 -2 Data Notes:

a) ART = Initial RTNDT + ARTNDT + Margin (°f) b) These ART values are used to generate the heatup and cooldown curves.

When two or more credible surveillance data sets become available, the data sets may be used to determine ART values as described in Regulatory Guide 1.99, Revision 2, Position 2.1. If the ART values based on surveillance capsule data are larger than those calculated per Regulatory Guide 1.99, Revision 2, Position 2.1, the surveillance data must be used. If the surveillance capsule data gives lower values, either may be used.

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Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-6 Calculation of Adjusted Reference Temperature Values at 35 EFPY for the Limiting Callaway Plant Reactor Vessel Material (Lower Shell Plate R2708-1)

Parameter ART Value Location 3/4 T 3/4 T Chemistry Factor, CF (°F) 44.0 44.0 Fluence, f(109 n!cm2) 1.201 4.264 Fluence Factor, FF(b) 1.05 1 0.763 ARTNDT = CF x FF, (°F) 46.24 33.57 Initial RTNDT, I (°F) 50 50 Margin,_M_(Of)(C)

ART=I+(CFxFF)+M 130 117

(°F)

Per Regulatory Guide 1.99, Rev 2 Notes:

a) Fluence, f, is based upon fsuf (10 nlcm2, E> 1.0 MeV) = 2.016 at 35 EFPY. The Callaway Plant reactor vessel wall thickness is 8.63 inches at the beltline region.

= f(O280bo*bog f) b) Fluence Factor, FF, per Regulatory Guide 1.99, Revision 2, is defined as FF c) Margin is calculated as M = 2(o2 + A2)° The standard deviation for the initial RTNDT margin term is 0 °F since the initial RTNDT is a measured value. The standard deviation for ARTNDT term a is 17 °F for the plate, except that c need not exceed 0.5 times the mean value of ARTNDT.

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Enclosure to ULNRC06350 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-7 RTpTS Calculations for Callaway Plant Beitline Region Materials at 35 EFPY Material Fluence FF CF ARTi5 Margin RTNDT(U)(a) RTPTst1 (1019 (°F) (°F) (°F) (°F) (°F) nlcm2, E

> 1.0 MeV)

Intermediate Shell 2.074 1.20 31.0 37.2 34.0 40 111 Plate_R2707-1 Intermediate Shell 2.074 1.20 37.0 44.4 34.0 10 88 Plate_R2707-2 Intermediate Shell 2.074 1.20 37.0 44.4 34.0 -10 68 Plate_R2707-3 LowerShellPlateR27O8-1 2.074 1.20 44.0 52.8 34.0 50 137 Using S/C Data 2.074 1.20 25.0 30 17.0 50 97 LowerShellPlateR27O8-2 2.074 1.20 37.0 44.4 34.0 10 88 LowerShellPlateR27O8-3 2.074 1.20 51.0 61.2 34.0 20 115 Inter. & Lower Shell Long. Weld Seams 1.167 1.04 29.7 30.9 30.9 -60 2 101-124A&101-142A (90°Azimuth)

Using S/C Data 1.167 1.04 39.9 41.49 28.0 -60 9.5 Inter. & Lower Shell Long. Weld Seams 101-124B&Cand 2.042 1.19 29.7 35.3 35.3 -60 11 101-142B&C (210°

&_330°_Azimuth)

Using S/C Data 2.042 1.19 39.9 47.48 28.0 -60 15.48 Inter, to Lower Shell Circumferential 2.074 1.20 29.7 35.6 35.6 -60 11 Weld Seams 101-171 Using S/C Data 2.074 1.20 39.9 47.82 28.0 -60 15.28 Notes:

a) Initial RTNDT values are measured values b) RTpTS = RTNDT(U) + Margin + ART15 c) CF

• FF ARTpTS =

d) Projected no. of EFPY at the EOL.

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