ULNRC-06998, Pressure and Temperature Limits Report, Revision 8

Pressure and Temperature Limits Report, Revision 8

ML26042A369
Person / Time
Site:	Callaway
Issue date:	02/11/2026
From:	Ameren Missouri, Union Electric Co
To:	Office of Nuclear Reactor Regulation
Shared Package
ML26042A367	List: ULNRC-06998, Pressure and Temperature Limits Report, Revision 8
References
ULNRC-06998
Download: ML26042A369 (0)

Text

Figure 14.9 CALLAWAY PLANT PRESSURE AND TEMPERATURE LIMITS REPORT Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents 1

Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR) 2 Operating Limits 2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.2.

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

Reactor Vessel Material Surveillance Program Reactor Vessel Surveillance Data Credibility Supplemental Data Tables References Callaway Energy Center Page 1 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT LIST OF FIGURES FIGURE 2.1-1 Callaway Plant Reactor Coolant System Heatup Limitations (Heatup Rates of 60° and 100°F/hr) Applicable to 54 EFPY(With Margins for Instrumentation Errors) 2.1-2 Callaway Plant Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60 and 100°F/hr) Applicable to 54 EFPY (With Margins for Instrumentation Errors)

Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System Callaway Energy Center Page 2 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT LIST OF TABLES Callaway Plant Heatup Limits at 54 EFPY with Margins for Instrumentation Errors Callaway Plant Cooldown Limits at 54 EFPY with Margins for Instrumentation Errors Callaway Plant COMS Maximum Allowable PORV Setpoints at 54 EFPY Callaway Plant Surveillance Capsule Data Interim Chemistry Factors Callaway Plant Lower Shell Plate R2708-1 Credibility Evaluation Callaway Plant Surveillance Weld Metal Credibility Evaluation Comparison of Callaway Unit l Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Calculation of Chemistry Factors Using Surveillance Capsule Data Reactor Vessel Beltline Material Unirradiated Toughness Properties Reactor Vessel Extended Beltline Material Unirradiated Toughness Properties Fluence (n/cm*, E>1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for Callaway Plant Summary of Adjusted Reference Temperature (ART) Values at the 1/4T and 3/4T Locations for 54 EFPY Beltline Material Summary of Adjusted Reference Temperature (ART) Values at the 1/4T and 3/4T Locations for 54 EFPY Extended Beltline Material Calculation of Adjusted Reference Temperature Values at 54 EFPY for the Limiting Callaway Plant Reactor Vessel Material (Lower Shell Plate R2708-1)

RTpts Calculations for Callaway Plant Beltline Region Materials at 54 EFPY RTprs Calculations for Callaway Plant Extended Beltline Region Materials at 54 EFPY Callaway Energy Center Page 3 of 27 Revision 8

1 Figure 14.9 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 Specifications (TS) 5.6.6. The TS addressed in this report are listed below:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Cold Overpressure Mitigation System (COMS)

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 P/T 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 RTypr for these regions when the pressure exceeds 20% of the preservice hydrostatic test pressure (3106 psig).

2.1 RCS Pressure and Temperature (P/T) Limits (LCO 3.4.3) 2.1.1.

The RCS temperature rate-of-change limits are:

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

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

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

2.1.2 The RCS P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-] and 2.1-2.

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 AppendixG 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 TS LCO 3.4.6, LCO 3.4.7, and LCO 3.4.12.

Callaway Energy Center Page 4 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTIES BASIS LIMITING MATERIAL: Lower Shell Plate R2708-I LIMITING ART VALUES AT 54 EFPY:

YT, 143°F (Axial Flaw)

¥%4T, 129°F (Axial Flaw) 2500 i aan

-l OperlimAnalysis Version:5.4 Run:3220 Operlimxtsm Version: 5.4.4 2250 4

[Leak Test Limit l

2000

lUnacceptable 7

Operation

]

Heatup Rate 100°FHr Critical Limit 60°FHr Critical Limit Limiting 100°F/Hr Temperaure of the Vessel.

Flange an Acceptable

\(170°F}

4 Operation Eee i

©a

o

=

=

ri

ewa wo a° a

oO a!

Criticality Limit based on Boltup inservice hydrostatic test Temperature.

"temperature (215°F) for the (70°F) service period up to 54 EFPY the lower limit for RCS pressure is -14.7 psig

-280 ee ett dee ee eee Pee err errr 0

50 6=6©100

)<<=6150 06200)

>>=250 386300

<<= 350 03686400 0386450) 0365000 550 Moderator Tem perature (Deg. F)

Figure 2.1-1 Callaway Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60°F and 100°F/hr).

Applicable for 54 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 5 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1-1 Callaway Plant Heatup Limits at 54 EFPY With Margins for Instrumentation Errors Criticality 100°F/hr a

Criticality Limit Heatup Limit T

p

+

p T

)

T (CF)

(psig)

(°F)

(psig)

CF) l (psig)

_CF)

(psig)

~14.7 215 14.7 70

-14.7

-14.7 a

215 887 70 558° 697 561 220 31

[5 a

25.

Ee 225 980

_80

558, 755 60°F/hr Heatup 561 230 1034 85 790 561 235 1094 828 (61 240 1161

240, 871 561 245 1234 918 561 p50 1315 558 250 971 561 255 1405 58 1029 561 260 1503 558 11094 561 26s 1612 1165 561 270 1733 1244 561 275 1865 0

er 1331 280 2012 1427 285 2173 11533 290 23520 5

1650 1778 1921 2078 2250 2441 L

Temperature (°F) l 198 215 Leak Test Limit Pressure (psig) 2000 2485 Callaway Energy Center Page 6 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTIES BASIS LIMITING MATERIAL: Lower Shell Plate R2708-1 LIMITING ART VALUES AT 54 EFPY:

YT, 143°F (Axial Flaw)

¥%T, 129°F (Axial Flaw) 2500 a

se een lOpettimanatysis Version:5.4 Run:3220 Opertim.xtm Version: 54.1}

Unacceptable

_ Operation oe lA Cooldown]

Rates Shown

°FHr Steaily-state 20 (PSIG Acceptable

, Operation 20°F:Hr and steady-state a

peea we we ow in oa mad

@-=

=

oe GC

.S)

KH the lower limit for RCS pressure is -14.7 psig ono 1. PemeeT f 0

50 06100

<<©6150 6200 6250 06300 36350 36400

§=686450)6800)6=5 50 Moderator Temperature (Deg. F)

Figure 2.1-2 Callaway Unit l Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 100°F/hr). Applicable for 54 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 7 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1-2 Callaway Plant Cooldown Limits at 54 EFPY With Margins for Instrumentation Errors Steady State Z

20°F /hr 40°F/hr 60°F/hr T

Pp T

p T

p T

p

(°F)

(psig).

(°F)

(psig) 7)

(psig)

(°F)

(psig)

"14,7 70 147° 70 44700 0 AT 0

[_

100°F/hr T

p CF)

(psig)

-14.7 00 1

4

[3 42 3

7 Callaway Energy Center Page 8 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT CALLAWAY COMS Maximum Allowable PORV Setpoints 8o oo aa.

oe£ B

av

[7]

© a

• PORV #2 we om PORV #1 RCS Temperature (°F)

Figure 2.2-1 Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System PORV #2 = High PORV Setpoint PORV #1 = Low PORV Setpoint Callaway Energy Center Page 9 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2-1 Callaway Plant COMS Maximum Allowable PORV Setpoints at 54 EFPY Maximum Allowable Function Generator Setpoints (Breakpoints) 7 Breakpoint Temperature High Setpoint Low Setpoint Number RCS (°F)

(psig) i (psig) 456 456 485 485 563 563 350 563 450 2335 500 2335 Note: Setpoints assume that all 4 or less RCP's are in operation.

Callaway Energy Center Page 10 of 27 Revision 8

3 Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Reactor Vessel Material Surveillance Program The reactor vessel material surveillance program is in compliance with Appendix H to 10 CFR 50, entitled "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 reports are as follows:

1.

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

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

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

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

WCAP-18001-NP, September 2015, "Analysis of Capsule W from the Ameren Missouri Callaway Unit 1 Reactor Vessel Radiation Surveillance Program."

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 five 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.

Callaway Energy Center Page 11 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Criterion 1:

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

The beltline region ofthe 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."

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 R2708-2, 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. AII 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. Lower Shell Plate R2708-1 has the same material heat number as the Lower Shell Plate R2708-3; therefore, surveillance credibility conclusions, Position 2.1 chemistry factor determinations and Position 2.2 percent USE reduction calculations also apply to this plate.

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 R2708-1 was chosen for the surveillance program since it had the highest RTnpr and the second 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 beltline 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.

Callaway Energy Center Page 12 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT 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 Appendix C of WCAP-18001-NP, September 2015, "Analysis of Capsule W from Ameren Missouri Callaway Unit l 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 criterion.

Criterion 3:

When there are two or more sets of surveillance data from one reactor, the scatter of ARTnpr values about a best fit line drawn as described in Regulatory Position 2.1 normally should be less than 28°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 shift 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 E185-82.

The functional form of the least squares method as described in Regulatory Position 2.1 will be utilized to determinea best-fit line for this data and to determine if the scatter of the ARTnoprt 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 13 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-1 Callaway Plant Surveillance Capsule Data Interim Chemistry Factors FF x

[_ ARTyopt U

l 0.

599 3.970.453 Lower Shell Platel Y

25.1

26.08 1.080 R2708-1 V

16.3 19.88 l 1.487 (Longitudinal) i

= X 25.9 33.39 1.662 W

58.6 83.81 2.045 U

s 0303-0673 LI 17.56 0.453 Y

V x

W Capsule l Material Capsule FF)

ARTyor FF 46.7 l

48.52" 1,080 45.2 l 55.12) l1.487 30.8 39.70 1.662 l 95.2 l

136.15 2.045 Sum:

456.25 13.454 CF po708-1=,(FF* ARTyp1)/D (FF)

= (456.25°F)/(13.454)

= 33.9°F L_.

l 0303 l

0673 l

66.2%

[

4454 l 0.453 Surveillance 1.15 1.039, 35.0%

36.37_

1.080 Weld 2.25 1.220 46.2 56.34 1.487 Material = ["

2.98 1.289 49.7

~

64.07

~

1.662 i

5.79 l

1.430 65.89 94.10 2.045 l

i Sum:

295.43 l

6.727

[ CBsurv, wets=(FF*

ART yp1)/D(FF) = (295.43°F (6.727) = 43.9°F Lower Shell Plate R2708-1 (Transverse) f =calculated fluence from Ex-Vessel dosimetry analysis results, (x 10 l n/em, E> 1.0 MeV).

These values were evaluated as part of WCAP-18203-NP, Revision 0.

FF = fluence factor = f(078-91189, ARTwnpr values are the measured 30 ft-1b shift.

These measured ARTnopr 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 l reactor vessel; therefore, no temperature adjustment is required.

Even though a reduction should not occur, using the negative measured ARTnpr value produces the most conservative results for this credibility evaluation.

The functional form of the least squares method as described in Regulatory Guide 1.99 Position 2.1 is presented in Tables 4.0-2 and 4.0-3. This method will be utilized to determinea best-fit line for this data and to determine if the scatter of these ARTNpr values about this line is less than 28°F for the weld and less than 17°F for the plate. Scatter of the ARTnpr values calculated using the fluence values for each capsule, as presented in WCAP-18203-NP, Revision 0, will be determined and used to judge credibility.

Callaway Energy Center Page 14 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT l

l.

Table 4.0-2 Callaway Plant Lower Shell Plate R2708-1 ee Peete Measured b

Best Fit Scatter of Base Material Capsule FF ARTnopr ARTwor (CF)

ARTyot (°F)

(Base a

. (30 ft-lb) °F)

Metals) 0673

-59 22.81 28.71 NO 1.039 25.1 35.22 10.12 Yes 1.220 16.3 41.34 25.04 l

1.289 i

25.9 43.70 17.80 (1430

586 4848 =

10.12 0.673 26.1

[l 22.81 3.29 1.039 46.7 35.228 1220 45.2 l

4134 3.86 1.289 308 48.70Ss<<*d2DD

_l 1490.

952 4848 l

46.72

<17°F Cc t

l

_ Lower Shell Plate R2708-1 (Longitudinal) l l

Lower Shell Plate R2708-1 (Transverse)

=éeK<

K ce RK Xx Notes:

a)

Best Fit Line Per Reg. Guide 1.99, Rev. 2 Position 2.1: Interim Chemistry Factor (33.9°F) x FF.

b)

Scatter ARTxpr= Absolute Value [Best Fit ARTjpr Measured ARTypt].

Table 4.0-2 indicates that three measured plate ARTNpr values are below the lower bound Io of 17°F, meaning the best fit line is over-predicting these measured ARTnpr values. Table 4.0-2 also indicates that one measured plate ARTNpr value is above the upper bound Io of 17°F. Froma statistical point of view +

lo (17°F) would be expected to encompass 68% of the data. Since, four of the measured plate ARTNpr values are outside of the 1o bound of 17°F, the plate surveillance data set has a scatter greater than + lo.

In summary, the measured plate data is not within the acceptable range. Therefore, the plate data does not meet this criteria.

Callaway Energy Center Page 15 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT l

Table 4.0-3 Callaway Plant Surveillance Weld Metal Credibility Evaluation Measured

- (a)

Best Fit Scatter of

< 28°F Capsule FF ARTwnprt (30

°

° l

l fi-lb) °F)

. ARTnoprt( F) l AR Tnor ( 2)

_ (Weld Metal) l__u

06730~C(

62s 29.54 36.66 NO y l 1039 l 350

4561 1061 l

Yes l

12200 2

46.2.

53.54 i.

734.

l Yes

_ 1.289 49.7 56.59 6.89 Yes 1.430 65.8 62.78 l

3.02 "Yes Notes:

a)

Best Fit Line per Reg. Guide 1.99, Rev. 2 Position 2.1: Interim Chemistry Factor (43.9°F) x FF.

b)

Scatter ARTNpr= Absolute Value [Best Fit ARTnpt Measured ARTnpt].

One measured weld ARTnpr value is above the upper lo at 28°F. The fact that one of the measured weld ARTnor values is out of lo 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.

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

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 °F.

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 responses to the five criteria of Regulatory Guide 1.99, Revision 2, Section B, the Callaway Plant surveillance weld data is credible and the surveillance plate data is non-credible.

Callaway Energy Center Page 16 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Supplemental Data Tables Table 5.0-1 Table 5.0-2 Table 5.0-3 Table 5.0-4 Table 5.0-5 Table 5.0-6 Table 5.0-7 Table 5.0-8 Table 5.0-9 Comparison of Callaway Plant Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions.

Calculation of Chemistry Factors Using Surveillance Capsule Data.

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

Provides the unirradiated reactor vessel extended beltline and nozzle material toughness data.

Provides a summary of the pressure vessel neutron fluence values at 54 EFPY used for the calculation of theART values.

Provides a summary of the adjusted reference temperature (ART) for reactor vessel beltline materials at the 1/4T and 3/4T locations for 54 EFPY.

Provides a summary of the adjusted reference temperature (ART) for reactor vessel extended beltline materials at the 1/4T and 3/4T locations for 54 EFPY.

Shows the calculation of the ART at 54 EFPY for the limiting reactor vessel material (lower shell plate R2708-1).

Provides RTprs values for 54 EFPY for the beltline materials.

Table 5.0-10 Provides RTprs values for 54 EFPY for the extended beltline materials.

Callaway Energy Center Page 17 of 27 Revision 8

6 1.

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT References Technical Specification 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)."

NRC letter dated [March 24, 2000], [CALLAWAY PLANT, UNIT 1

- ISSUANCE OF AMENDMENT RE: PRESSURE TEMPERATURE LIMITS REPORT]

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

WCAP-18216-NP, Revision l, "Callaway Unit 1 Heatup and Cooldown Limit Curves for Normal Operation through End of License Extension, July 2017.

Westinghouse Letter, LTR-SCS-17-38, Revision 0 "Callaway Unit 1 Cold Overpressure Mitigation System (COMS) Setpoint Analysis for 54 EFPY.", July 31, 2017.

WCAP-18203-NP, Revision 0, "Ex-Vessel Neutron Dosimetry Program for Callaway Cycle 21,"

(DORT-based fluence calculation), December 2016.

Callaway Energy Center Page 18 of 27 Revision 8

Figure 14.9 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, Materials

- Lower Shell 1 Plate R2708-1 (Longitudinal) l Lower Shell

' Plate R2708-1 (Transverse)

' Weld Metal L

l HAZ Metal L

___ Revision 2, Predictions 30 ft-lb Transition Upper Shelf Energy Fluence Temperature Shift Decrease L(x 10! n/em)

Predicted

" Measured Predicted Measured

(°F)

(a)

(°F)

(b) l

_(%)

(a)

(%)

(c) 0.303 l

34.3 00° 14.5 2

1.15 530 l

25.1 20.0 2.25 62.2 16.3 23.0

2.98 65.7 25.9 25.0 5.79 72.9 58.6 29.0 0.303 34300~C<<T 61 14.5 11S 53.0 46.7 20.0 2.25 i

62.2 45.2 23.0 2.98 65.7 30.8 25.0

_5.79 729 95.2 29.0 0.303 21.4

, 66.2 14.5, 11S 33.0

=

35.0 20.0

2.25 38.8

_ 46.2 23.0 2.98 41.0_

49.7 25.0 579 9;

45.5 65.8 29.0 0.303 66.3

1.15 56.5 2.25 56.8 2.98 l

42.3 5.79 O16 l

Capsule I

InN Dan I

N

WWON SM W

¢(

U

_Y V

X W

U Y

V X

W U

Y Vv X

- Ww U

-Y V

X W

Notes:

a)

Based on Regulatory Guide 1.99, Revision 2, methodology using the updated weight percent values of copper and nickel of the surveillance material per WCAP-18216-NP.

b)

Calculated using measured Charpy data plotted using CVGRAPH, Version 6.0.

c)

Values are based on the definition of upper shelf energy given in ASTM EI85-82.

d)

Actual measured value for ARTNopr is -5.9°F. This physically should not occur; therefore for conservatism a value of zero will be used.

e)

USE values were calculated to have increased. Physically, this should not occur; therefore, conservative values of zero percent will be used.

Callaway Energy Center Page 19 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-2 Calculation of Chemistry Factors Using Surveillance Capsule Data Capsule [

l (C) 7 FF x l

a)

Material Capsule FF ARTypr ARTNDT FF ul 0303 [l 0.673 00

=f 00

=f 0.453 l

Lower Shell Plate R2708-1 (Longitudinal) 1.15 1.039 l

9 25.1

26.08 1.080 Y

Vv 2.25 1.220 16.3 19.88 1.487 X

2.98 1.289 259 l

3339 l 1.662 Ww 5.79 1.430 58.6 83.81 l

2.045 U

0.303 [l 0.673 26.1 17.56 0.453 YY 1.15 1.039 46.7 "48.52

<<1.080 V

x Ww Lower Shell Plate R2708-1 (Transverse) 2.25 1.220 45.2 55.12 1.487 2.98 1.289 l 30.8 39.70, 1.662 5.79 1.430 95.2

~~

136.15 2.045 Sum:

460.22 13.454 CF R2708-1=

> (EF* ARTyp1)/> (FF) = (460.22°F)/(13.454) = 34.2°F

[

0.303 l 0.673 66.2 l

44sa 0.453 l

115) l 1.039 35.0 36.37 l

1.080.

Surveillance 2.25 1.220 [46.2 56.34 1.487 Manet

)

2.98 1.289 49.7 64.07 l

1.662 5.79 1.430 6589 l

94.10 2.045 Sum; 295.43.

6.727 CFsun. Weld=),(FF* ART pt)

(FF)

= (295.43°F)/(6.727) = 43.9°F Materia Notes:

a) f=calculated fluence from Ex-Vessel dosimetry analysis results, (x 10'? n/em, E> 1.0 MeV). These values were evaluated as part of WCAP-18203-NP, Revision 1.

b)

FF = fluence factor = f 078-0 "les 9.

c)

ARTyopr values are the measured 30 ft-lb shift.

d)

The surveillance weld metal ARTnpr values have not been adjusted bya ratio factor. This method is conservative, because the ratio factor is less than 1.

e)

The actual value is -5.9°F, but for conservatism a value of zero is considered.

Callaway Energy Center Page 20 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-3

Reactor Vessel Beltline Material Unirradiated Toughness Properties Material

___ Description Replacement Closure Head

-40°F Flange

Vessel Flange R2701-L l

4 Intermediate Shell

° Plate R2707-1

40"F

_ Intermediate Shell

=

10°F Plate R2707-2 Intermediate Shell

-10°F Plate R2707-3 Lower Shell Plate R2708-1 Lower Shell Plate R2708-2 Lower Shell Plate R2708-3 Intermediate and Lower Shell Longitudinal Weld Seams © Intermediate to Lower Shell Circumferentia 1 Weld Seam Surveillance l

Weld © Notes:

a)

Material details for replacement head provided in CMTR, 58-9181839-000.

b)

The initial RTNDT values for the plates and welds are based on measured data (WCAP-18216-NP, Revision 1).

c)

All vessel beltline 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 beltline weld seams. Chemical composition of surveillance weld is based on data from WCAP-111374 RI.

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 (including a 10°F instrument uncertainty).

Cu(%)

Ni (%)

Initial RTypr 40°F 50°F 10°F 20°F

-60°F 0.045 l

0.065 Callaway Energy Center Page 21 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-4 Reactor Vessel Extended Beltline and Nozzle Material Unirradiated Toughness l

Material Description Nozzle Shell Plate Nozzle Shell Plate Nozzle Shell Plate Nozzle Shell to Intermediate Shell Weld, Seam Nozzle Shell Longitudinal Weld Seams Inlet Nozzle Inlet Nozzle Inlet Nozzle Inlet Nozzle Outlet Nozzle

__ Outlet Nozzle Outlet Nozzle Outlet Nozzle Inlet Nozzle to Shell Weld Seams Outlet Nozzle to Shell Weld Seams Material Identification R2706-1 R2706-2 R2706-3 103-121 101-122A 101-122B 101-122C R2702-1 R2702-2

~

R2702-3 R2702-4 R2703-1 R2703-2

'R2703-3 R2703-4 105-121 A 105-121 B 105-121 C 105-121 D 107-121 A 107-121 B 107-121 C 107-121 D Callaway Energy Center Properties Material Heat Number l

c4202-1 4242-1 l

8307-1 90211 IAQIE l

AACE.

[

87000

~ GABID FAOED 87000 87000 l

EACAE 12-5468 12-5468 11-5450 11-5450 Cu(%)

0.05 0.06 0.08 0.028 0.04 0.025 0.045 0.02 0.03 0.045 0.045 0.035 0.16 0.16 0.16 0.16

__EV-3132 j 0.16 EV-3133 l

_EV-3134 EV-3135 3P7246 HABIE lEAAHF 4P7656 BABEF

~ EAAHF FABAF 3P7246 4P7656 BABEF__

_4P7656 l BABEF 3P7317 FABAF

__ EAAHF

~~

4P7656 l

BABEF 3P7246 4P6524 3P7317 HABIE

~ FAAHF EAAHF Page 22 of 27 0.16 0.16 0.59 0.65 0.57 0.0825 l

1.005 l

0.98 0.13 1.00 0.98 0.13 0.13 1.005 0.90 0.88 0.78 0.79 0.64 0.67 0.68 0.66 0.035 1.01 ik ee 0.025 0.97 0.015 0.01 0.035 0.025 0.97 0.025 0.97

~~ 0,043 l

0.01

~

0.015 0.025 0.97 l

0.035 l

0.06 0.043 Revision 8 Ni (%)

l Initial RTyxpr 20°F 30°F 30°F

-60°F

~80°F

-70°F

-40°F

-50°F

-60°F

-40°F

-40°F

-80°F 10°F 10°F

-10°F

-10°F

-10°F 10°F 10°F 0°F

-60°F

-80°F

-40°F

-70°F

-70°F

-40°F

=

-60°F l

60°F

-70°F

-70°F

~70°F

-70°F

-80°F

-60°F

-40°F

-70°F

-70°F

-60°F

-80°F

-80°F

-80°F

-50°F

-40°F

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT

~ Table 5.0-5 Fluence (n/em, E>1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface

_ for Callaway Plant roy:

Lower Shell to Inlet Nozzle to RPV Base Metal Lower V l Head Nozzle Shell to Int.

Nozzle Shell EFPY Beltline Maximum ower Vesse Shell Cire. Weld Welds Circ. Weld (Lowest Extent) 105 TITEHT 8.14E+14 1.48E+16 1.72E+15 27]

l.S2E+18 l

L70EHS 3.49E+16 4.13E+15 3.43 l

l 2.45E+18.

l

~-2.83E+15 5.18E+16 6.05E+15 4.70 3.23E+18 3.74E+15 6.85E+16 7.95E+15 5.97 3.98E+18 4.59E+15 8.28E+16 9.62E+15 7:9 4.71E+18

-5.40E+15

{.00E+17

)

1ATE+16 8.59

-5.40E+18 l

6.26E+15 LAQ2BHI7 1.30E+16 9.94 6.16E+18 7.1 9B +15 1.29E+17

)

L.SIE+16 1B 6.80E+18 7.89E+15 1.43E+17 L.68E+16-12.49 TSTEH8 8.83E+15 1.63E+17 LOIE+6 13.87

© 832EH8 9.78E+15

~179E+17 l

-2.09EH16 1560 9.08E+18 1.08E+16 l.97E+17 2.29E+16 16.48 9.90E+18 1.18E+16 2.1SE+17 248 H16.

17.70 1.06E+19 127EH6 2.32E+17 l

2.67E+16 18.99 1.13E+19 1.36E+16 7

250EH7 2.87E+16 50.41 121E+19 1

ASE+16 2.69E+17 3.10E+16 1.76 1.29E+19 1 S4E+16 288E+17 3.32E+H16 23.09 l

1.36E+19 l

1,62E+16 3.07EH17 3.53E+16 94.45 1.43E+19 LI0EH6 325EH7 3.74E+16 25.75 L49EH19

l 1.76E+16 3.43EH17 3.95E+16 2707

[

1S6EHO

1. 85E+16 3.60E+17 AIA

+16 30.00 L.72E+19 2.02E+16 3.97E+I7

/

4STETIG 36.00 2.04E+19 2.39E+16 4.72E+17 "5

45E+16 42.00 2.36E+19 2.76E+16 5.47E+17 6.33E+16 48.00 2.68E+19 3.12E+16 6.23E+17 7.20E+16 54.00 3.00E+19 3.49E+16 6.98E+17 8.08E+16 Note:

a)

Since the inlet nozzle-to-nozzle shell welds are at a lower elevation than the outlet nozzle-to-nozzle shell

weld, the neutron fluence values reported herein for the inlet nozzle-to-nozzle shell welds are conservatively applicable to the outlet nozzle-to-nozzle shell welds.

Callaway Energy Center Page 23 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-6 Summary of Adjusted Reference Temperature (ART) Values at the Material Description AT ARTF)

Ly TART(CF).

Intermediate Shell 105 Plate R2707-1 Intermediate Shell 8?

__ Plate R2707-2 Intermediate Shell Plate R2707-3

[

_ Lower Shell Plate R2708-1 Using Surveillance Capsule a

Data)

Lower Shell Plate R2708-2 Lower Shell Plate R2708-3 Using Surveillance Capsule Data?

Intermediate & Lower Shell Longitudinal Weld Seams 101-124A & 101-142A (90°Azimuth)

Using Surveillance Capsule Data Intermediate & Lower Shell Longitudinal Weld Seams 101-124B&C & 101-142B&C

_ (210° & 330° Azimuth)

Using Surveillance Capsule Data?

Intermediate to Lower Shell Circumferential Weld Seam 101-171 Using Surveillance Capsule 19 Data l

Ce Notes:

a)

ART= Initial RT yor

+ ARTyor + Margin (°F). These ART values are used to generate the heatup and cooldown curves.

b)

When two or more 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, provided that the data is deemed credible. If the surveillance capsule data gives lower values, either may be used. The plate material surveillance data is non-credible. The weld material surveillance data is credible.

Callaway Energy Center Page 24 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-7 Summary of Adjusted Reference Temperature (ART) Values at the 1/4T and 3/4T Locations for 54 EEPY Extended Beltline Material l

Material Heat 54 EFPY ART° Number

__ '&T ART (CF)

% T ART (°F)

Nozzle Shell Plate R2706-1

[4202-1 20 20

_ Nozzle Shell Plate R2706-2 4242-1 Z

30 Nozzle Shell Plate R2706-3 B8307-1 30 90211

-60 IAOJE

-80

[--<<

SAACE

-70 87000 40 Nozzle Shell Longitudinal Weld Seams 101-GABID

-50 122 A,B, &C FAOED

-60 oon l

EACE L.

2 80 Notes:

(a)

Per TLR-RES/DE/CIB-2013-01, only materials with fluence values equal to or greater than 1x10" n/em? (E > 1.0 MeV) at end-of license (EOL = 35 EFPY) are considered extended beltline and must be evaluated for embrittlement effects. Materials that are listed in Table 5.0-4, but are not included in this Table have fluence values less than 1x10" n/em? (E > 1.0 MeV) at end-of-license extension (EOLE = 54 EFPY) and may be neglected from embrittlement evaluation.

ART= Initial RTyp

+ ARTnpr + Margin (°F).

Calculated extended beltline ARTypr values were less than 25°F and were reduced to zero per TLR-RES/DE/CIB-2013-01. This resulted in no calculated embrittlement (Initial RT yp; = ART).

Material Description ozzle Shell to Intermediate Shell Weld Seam 103-121 Table 5.0-8 Calculation of Adjusted Reference Temperature Values at 54 EFPY for the

_ Limiting Callaway Plant Reactor Vessel Material (Lower Shell Plate R2708-1)

Parameter ART Value Location T

L

%4T Chemistry Factor, CF @F) 51.0 51.0 Fluence, f (10"? n/em?)

i) 0.635

__Fluence Factor, FF 1.159 l

0,873 I

ARTnoprt= CF x FF, (°F) 59.1 l

44.5 Initial RTor, I (CF) 50 50 l

Margin, M (°F) l 34 l

34 F

ART =1+(CF x FF)+M

(°F)

Per Regulatory Guide 1.99, 143 Rev 2 i

L l

Notes:

a)

Fluence, f, is based upon f,y,.(10"? n/em, E > 1.0 MeV) = 3.00 at 54 EFPY. The Callaway Plant reactor vessel wall thickness is 8.63 inches at the beltline region.

b)

Fluence Factor, FF, per Regulatory Guide 1.99, Revision 2, is defined as FF c)

Margin is calculated as M = 2(6; + o,)"* The standard deviation for the initial RT yp; margin term 6, is 0°F since the initial RTyp is a measured value. The standard deviation for ARTyp; term oy, is 17°F for the plate, except that c, need not exceed 0.5 times the mean value of ART npr.

= £(028-0.10"log f)

Callaway Energy Center Page 25 of 27 Revision 8

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-9 RT prs Calculations for Callaway rent Beltline Region Materials at 54 EFPY Material Fluence (I 0!

CF ARTpts fc)

Margin

• RTxorw

- n/em, E> 1.0

(°F)

(°F)

(°F)

(°F)

MeV)

Intermediate Shell 3.00 370 478 Plate R2707-1 Intermediate Shell 56.8 l

Plate R2707-2 et BI 0-Intermediate Shell l

Plate R2707-3

on 47.8 Lower Shell Plate R2708-1 65.8 iLL.

Using S/C Data Lower Shell Plate R2708-2 Lower Shell Plate R2708-3 Using S/C Data Inter. & Lower Shell Long. Weld Seams 101-124A & 101-142A (90°Azimuth)

Using S/C Data Inter. & Lower Shell Long. Weld Seams 101-124B&C and 101-142B&C (210°

& 330° Azimuth)

Using S/C Data Inter. to Lower Shell Circumferential Weld Seam 101-wal Using S/C Data Notes:

a)

Initial RTypr values are measured values.

b)

RToprs =

RT ypu)

+ Margin + ARTprs, c)

ARTots = CF

• FF.

d)

Projected no. of EFPY at the EOLE.

Callaway Energy Center Page 26 of 27 Revision 8 RTers CF)

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-10 RTprs Calculationsfor Callaway Plant Extended Beltline Materials at 54 EFPY Material l Fluence (10 FF l

CF ARTprs° Margin l RTxpruw = RTprs'-

n/cm, E > 1.0 Ga)

(F)

@F)

(°F)

(°F)

MeV)

Nozzle Shell Plate R2706-1 0.0698 l 0349l 31.0 01108) 00 a)

Nozzle Shell Plate R2706-2 0.0698 0.349 9(12.9) 0.0 Nozzle Shell Plate R2706-3.

0.0698 0.349 0 (17.8) 0.0 Nozzle Shell to Intermediate Shell Weld Seam 103-121, Heat 0.0698 0.349 0 (9.2) 0.0

1. 90211 Nozzle Shell to Intermediate

\

Shell Weld Seam 103-121, Heat 0.0698 0.349 0(18.8) l 0.0

1. IAOJE Nozzle Shell to Intermediate Shell Weld Seam 103-121, Heat

0(11.9)

1. JAACE Nozzle Shell Longitudinal Weld Seams 101-122 A, B,

& C, Heat 013.5) li

1. 87000 j

i l

Nozzle Shell Longitudinal Weld Seams 101-122 A, B, & C, Heat 0 (9.4)

1. GABID l

Nozzle Shell Longitudinal Weld Seams 101-122 A, B, & C, Heat 0 (14.3)

1. FAOED Nozzle Shell Longitudinal Weld l

Seams 101-122 A, B,

& C, Heat

© 0 (16.6) l HEACAE l

Initial RTnpr values are measured values.

RTprs = RTwnpt(u) + Margin + ARTprts ARToprs = CF

• FF. Calculated extended beltline ARTnpr values were less than 25°F and were reduced to zero per TLR-RES/DE/CIB-2013-01. This resulted in no calculated embrittlement (Initial RTypt =

RTprs).Original calculated ARTpys values are provided in parentheses.

Projected no. of EFPY at the EOLE. Per TLR-RES/DE/CIB-2013-01, only materials with fluence values equal to or greater than 1x10'7 n/cm (E > 1.0 MeV) at EOL are considered extended beltline and must be evaluated for embrittlement effects. Materials that are listed in Table 5.0-4, but are not included in this Table have fluence less than 1x10" n/cm* (E > 1.0 MeV) at EOLE and may be neglected from embrittlement evaluation.

Callaway Energy Center Page 27 of 27 Revision 8