ULNRC-05356, Figure 14.9, Pressure and Temperature Limits Report

Figure 14.9, Pressure and Temperature Limits Report

ML070160266
Person / Time
Site:	Callaway
Issue date:	01/05/2007
From:	Fitzgerald D AmerenUE
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
ULNRC-05356, WCAP-14040-A, Rev 4
Download: ML070160266 (28)

Text

Union Electric One Ameren Plaza 1901 Chouteau Avenue PO Box 66149 St. Louis, MO 63166-6149 314.621.3222 January 5, 2007 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Stop P1-137 Washington, DC 20555-0001 Ladies and Gentlemen:

ULNRC-05356 WAmerenUE DOCKET NUMBER 50-483 CALLAWAY PLANT UNIT 1 FIGURE 14.9 CALLAWAY PLANT PRESSURE AND TEMPERATURE LIMITS REPORT Enclosed is the Callaway Plant Pressure and Temperature Limits Report (PTLR), Revision 5. Revision 5 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," May, 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 any new commitments. If you have any questions concerning this report, please contact us.

Sincerely,

,4&C 4j 41-%

~

David T. Fitzgerald Manager, Regulatory Affairs PMB/sjd

Enclosure:

Callaway Pressure and Temperature Limits Report, Revision 5 40c) /

a subsidiary ofAmaren Corporation

ULNRC-05356 January 5, 2007 Page 2 Mr. Bruce S. Mallett Regional Administrator U.S. Nuclear Regulatory Commission Region IV 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 Senior Resident Inspector Callaway Resident Office U.S. Nuclear Regulatory Commission 8201 NRC Road Steedman, MO 65077 Mr. Jack N. Donohew (2 copies)

Licensing Project Manager, Callaway Plant Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop O-7D1 Washington, DC 20555-2738 Missouri Public Service Commission Governor Office Building 200 Madison Street PO Box 360 Jefferson City, MO 65102-0360 Mr. Ron Reynolds Director Missouri State Emergency Management Agency P.O. Box 116 Jefferson City, MO 65102

Figure 14.9 CALLAWAY PLANT PRESSURE AND TEMPERATURE LIMITS REPORT Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table of Contents Section Page 1.0 Reactor Coolant System (RCS) PRESSURE AND 1

TEMPERATURE LIMITS REPORT (PTLR) 2.0 Operating Limits 1

2.1 RCS Pressure and Temperature Limits 1

2.2 Cold Overpressure Mitigation System 1

3.0 Reactor Vessel Material Surveillance Program 9

4.0 Reactor Vessel Surveillance Data Credibility 9

5.0 Supplemental Data Tables 16 6.0 References 16 Callaway Plant i

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1-1 Callaway Plant Reactor Coolant System Heatup Limitations 3

(Heatup Rates of 600 and 1000F/hr) Applicable to 28 EFPY (With Margins for Instrumentation Errors) 2.1-2 Callaway Plant Reactor Coolant System Cooldown Limitations 5

(Cooldown Rates of 0, 20, 40, 60 and 100'F/hr) Applicable to 28 EFPY (With Margins for Instrumentation Errors) 2.2-1 Maximum Allowed PORV Setpoint for the Cold Overpressure 7

Mitigation System List of Tables Table Page 2.1-1 Callaway Plant Heatup Data at 28 EFPY With Margins for 4

Instrumentation Errors 2.1-2 Callaway Plant Cooldown Data at 28 EFPY With Margins for 6

Instrumentation Errors 2.2-1 Callaway Plant COMS Maximum Allowable PORV Setpoints 8

at 28 and 35 EFPY Callaway Plant ii Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 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) 2.0 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 RTNDT 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.

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

b.

A maximum cooldown of 100TF 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 P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 and 2.1-2.

I Callaway Plant I

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT 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 G 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.

I Callaway Plant 2

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTIES BASIS LIMITING MATERIAL: Lower Shell Plate R2708-1 LIMITING ART VALUES AT 28 EFPY:

V4 T, 3 T, 1280F 1 120F 2500 2250 2000 1750 e--.A500 T1250 (D

-*1000 750 500 250 0~

0 100 200 300 400 500 Moderator Temperature (Deg. F)

Figure 2.1-1 Callaway Unit 1 Reactor Coolant System Heatup Limitations (Heatup Rates of 60OF and 1000F/hr). Applicable for 28 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 I Callaway Plant 3

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1-1 Callaway Plant Heatup Limits at 28 EFPY With Margins for Instrumentation Errors 60oF/hr Heatup Criticality Limit 100o/hr Heatup Criticality Limit T

P T

P T

P T

P (OF)

(psig)

(7F)

(psig)

(TF)

(psig)

(OF)

(psig) 70 0

200 0

70 0

200 0

70 561 200 561 70 561 200 561 75 561 200 561 75 561 200 561 80 561 200 561 80 561 200 561 85 561 200 561 85 561 200 561 90 561 200 561 90 561 200 561 95 561 200 561 95 561 200 561 100 561 200 561 100 561 200 561 105 561 200 561 105 561 200 561 110 561 200 561 110 561 200 561 115 561 200 561 115 561 200 561 120 561 200 561 120 561 200 561 125 561 200 561 125 561 200 561 130 561 200 561 130 561 200 561 135 561 200 561 135 561 200 561 140 561 200 561 140 561 200 561 145 561 200 561 145 561 200 561 150 561 200 561 150 561 200 561 155 561 200 561 155 561 200 561 160 561 205 561 160 561 205 561 165 561 210 561 165 561 210 561 170 561 210 1010 170 561 210 792 170 1010 215 1067 170 792 215 827 175 1067 220 1129 175 827 220 867 180 1129 225 1199 180 867 225 911 185 1199 230 1275 185 911 230 961 190 1275 235 1360 190 961 235 1016 195 1360 240 1453 195 1016 240 1076 200 1453 245 1557 200 1076 245 1144 205 1557 250 1671 205 1144 250 1218 210 1671 255 1797 210 1218 255 1301 215 1797 260 1936 215 1301 260 1392 220 1936 265 2089 220 1392 265 1493 225 2089 270 2258 225 1493 270 1604 230 2258 275 2445 230 1604 275 1726 235 2445 235 1726 280 1862 240 1862 285 2011 245 2011 290 2176 250 2176 295 2358 L 255 2358 Leak Test Limit Temperature (0F) 185 200 Pressure (psig) 2000 2485 I Callaway Plant 4

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTIES BASIS LIMITING MATERIAL: Lower Shell Plate R2708-1 LIMITING ART VALUES AT 28 EFPY:

1 T, 128-F 3/4 T, 112 0F 2500 2000 c,

1500 U_

"o M

S 1000 500 0

0 100 200 300 400 500 Moderator Temperatore (Deg. F)

Figure 2.1-2 I

Callaway Unit 1 Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20, 40, 60, and 1000F/hr). Applicable for 28 EFPY (With Margins for Instrumentation Errors). Includes vessel flange requirements of 170OF and 561 psig per 10 CFR 50, Appendix G.

Boltup temperature includes 10 'F instrument uncertainty I

I Callaway Plant 5

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.1-2 Callaway Plant Cooldown Limits at 28 EFPY With Margins for Instrumentation Errors Stead-State 20 'F/hr 40°F/hr 60*F/hr 100°F/hr T

P T

P T

P T

P T

P

(°F)

(psig)

(0F)

(psig)

(°F)

(psig) 0F)

(psig)

(0F)

(psig) 70 0

70 0

70 0

70 0

70 0

70 561 70 561 70 559 70 516 70 430 75 561 75 561 75 561 75 529 75 446 80 561 80 561 80 561 80 544 80 464 85 561 85 561 85 561 85 560 85 483 90 561 90 561 90 561 90 561 90 505 95 561 95 561 95 561 95 561 95 529 100 561 100 561 100 561 100 561 100 556 105 561 105 561 105 561 105 561 105 561 110 561 110 561 110 561 110 561 110 561 115 561 115 561 115 561 115 561 115 561 120 561 120 561 120 561 120 561 120 561 125 561 125 561 125 561 125 561 125 561 130 561 130 561 130 561 130 561 130 561 135 561 135 561 135 561 135 561 135 561 140 561 140 561 140 561 140 561 140 561 145 561 145 561 145 561 145 561 145 561 150 561 150 561 150 561 150 561 150 561 155 561 155 561 155 561 155 561 155 561 160 561 160 561 160 561 160 561 160 561 165 561 165 561 165 561 165 561 165 561 170 561 170 561 170 561 170 561 170 561 170 1267 170 1267 170 1267 170 1267 170 1267 175 1343 175 1343 175 1343 175 1343 175 1343 180 1427 180 1427 180 1427 180 1427 180 1427 185 1519 185 1519 185 1519 185 1519 185 1519 190 1621 190 1621 190 1621 190 1621 190 1621 195 1734 195 1734 195 1734 195 1734 195 1734 200 1859 200 1859 200 1859 200 1859 200 1859 205 1997 205 1997 205 1997 205 1997 205 1997 210 2149 210 2149 210 2149 210 2149 210 2149 215 2318 215 2318 215 2318 215 2318 215 2318 219.5 2485 219.5 2485 219.5 2485 219.5 2485 219.5 2485 I Callaway Plant 6

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT CALLAWAY COMS Maximum Allowable PORV Setpoints S

n-0.ft 0

50 100 150 200 250 300 350 RCS Temperature (deg. F) 400 I-High PORV setpoint -

Low PORV setpoint I Figure 2.2-1 Maximum Allowed PORV Setpoint for the Cold Overpressure Mitigation System I Callaway Plant 7

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 2.2-1 Callaway Plant COMS Maximum Allowable PORV Setpoints at 28 EFPY I

Maximum Allowable Function Generator Setpoints (Breakpoints)

Breakpoint Temperature High Setpoint Low Setpoint Number RCS (OF)

(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 Note: Setpoints assume that all 4 or less RCP's are in operation.

I Callaway Plant 8

Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT 3.0 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 capsule reports are as follows:

1.

WCAP-1 1374, 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."

4.0 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.

I Callaway Plant 9

Revision 5

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 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."

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. 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 R2708-1 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 I Callaway Plant 10 Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT 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.

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 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 280F 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 determine a best-fit line for this data and to determine if the scatter of the ARTNDT values about this line is less than 28TF for welds and less than 17TF 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.

I Callaway Plant 11 Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-1 Callaway Plant Surveillance Capsule Data Material Capsule Capsule FF(b)

ARTNI(c)

FF x

&fa)

ARTNDT Lower Shell U

0.331 0.696 0.0(e) 0.0 0.48 R2708-1 Y

1.27 1.07 25.15 26.91 1.14 (Longitudinal)

V 2.52 1.25 16.45 20.56 1.56 X

3.33 1.32 25.71 33.94 1.74 Lower Shell U

0.331 0.696 25.86 18.00 0.48 Plate Y

1.27 1.07 46.39 49.64 1.14 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 Surveillance CFo5 = (1FF

• ARTNDT)/-(FF 2) = (245.70-F)/(9.84) = 25.0-F Weld U

0.331 0.696 68.53(d) 47.70 0.48 Material(d)

Y 1.27 1.07 36.92(d) 39.50 1.14 V

2.52 1.25 48.21(d) 60.26 1.56 X

3.33 1.32 51.81(d) 68.39 1.74 Sum:

4.32 205.47 215.85 4.92 CFsurv. Weld = Y-(FF* ARTNDT)/Z(FF2) = (215.85°F)/(4.92) = 43.90F Notes:

(a) f= calculated fluence from capsule X dosimetry analysis results, (x 1019 n/cm 2, 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(O.2 - D. 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 I 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 271h Edition of the CRC Standard Mathematical Tables (page 497), for a straight line fit by the method of least squares, the values of bo and b, are obtained by solving the normal equations nbo

+

bY-xi =

Eyi and bo0 xi

+

b, Xi2 =

XXyii I Callaway Plant 12 Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT These equations can be re-written as follows (bo = a and b, = b):

n 7-yi i=1 n

an

+ b~xi i=1 and n

F~xiYi i=1 n

n

= a~xi

+ b~xi i=1 i=l 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=8.6411

+ 9.84b 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 Callaway Plant Lower Shell Plate R2708-1 Base Material FF Measured Best Fit(a)

Scatter of

< 170F ARTNDT (30 ARTNDT (0F)

ARTNDT (TF)

(Base ft-lb)(0F)

Metals)

Lower Shell 0.696 0.00 17.4

-17.4 NO Plate R2708-1 1.07 25.15 26.75

-1.6 Yes (Longitudinal) 1.25 16.45 31.25

-14.8 Yes 1.32 25.71 33.0

-7.29 Yes Lower Shell 0.696 25.86 17.4 8.46 Yes Plate R2708-1 1.07 46.39 26.75 19.64 NO (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 I Callaway Plant 13 Revision 5

I I l I Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-2 indicates that one measured plate ARTNDT value is below the lower bound la of 17OF by less than 1F. 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 a of 17TF by less than 3'F. From a statistical point of view + Io (17OF) 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 + la bounds. The fact that two of the measured plate ARTNDT values are outside of the 1a bound of 17TF 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 FF Measured Best Fit(a)

Scatter of

< 280F (Weld ARTNDT (30 ft-ARTNDrr (1F)

ARTNDT (TF)

Metal) 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.

I Callaway Plant 14 Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT One measured weld ARTNDT value is below the lower la at 280F. The fact that one of the measured weld ARTNDT values is out of la bound of 280F 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 claddingfbase metal interface within + 250F.

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 25TF. 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.

I Callaway Plant 15 Revision 5

z I

s Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT 5.0 Supplemental Data Tables Table 5.0-1 Table 5.0-2 Table 5.0-3 Table 5.0-4 Table 5.0-5 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 toughness data. The boltup temperature is also included in this Table.

Provides a summary of the pressure vessel neutron fluence values at 28 EFPY used for the calculation of the ART values.

Provides a summary of the adjusted reference temperature (ART) for reactor vessel beltline materials at the 1/4T and 3/4T locations for 28 EFPY.

Shows the calculation of the ART at 28 EFPY for the limiting reactor vessel material (lower shell plate R2708-1).

I I

Table 5.0-6 Table 5.0-7 Provides RTprs values for 35 EFPY.

6.0 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, SCP-06-66, "Final LTOPS Setpoint Analysis for Increased PORV stroke time", November 20, 2006.

I Callaway Plant 16 Revision 5

lib 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, Revision 2, Predictions Materials Capsule Fluence 30 ft-lb Transition Upper Shelf Energy (x 1019 Temperature Shift Decrease n/cm2 Predicted Measured Predicted Measured (OF) (a)

(oF) (b)

(%) (a)

(%) (c)

Lower Shell U

0.331 30.62 0.0(d) 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 58.08 25.71 25 5

Lower Shell U

0.331 30.62 25.86 14.5 11 Plate R2708-1 Y

1.27 47.08 46.39 20 13 (Transverse)

V 2.52 55.0 44.82 23.5 3

X 3.33 58.08 30.77 25 5

Weld Metal U

0.331 22.13 68.53 14.5 11 Y

1.27 34.02 36.92 20 14 V

2.52 39.75 48.21 23.5 8

X 3.33 41.98 51.81 25 8

HAZ Metal U

0.331 65.93 0

Y 1.27 56.38 14 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.

I I Callaway Plant 17 Revision 5

I I

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-2 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Capsule FF~b)

ARTNDec FF x

&fa)

ARTNDT Lower Shell U

0.331 0.696 0.0(e) 0.0 0.48 R2708-1 Y

1.27 1.07 25.15 26.91 1.14 (Longitudinal)

V 2.52 1.25 16.45 20.56 1.56 X

3.33 1.32 25.71 33.94 1.74 Lower Shell U

0.331 0.696 25.86 18.00 0.48 Plate Y

1.27 1.07 46.39 49.64 1.14 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:

245.70 9.84 Surveillance CF05 = Z(FF

• ARTNDT)/Z(FF2 ) = (245.700F)/(9.84) = 25.0'F Weld U

0.331 0.696 62.36(J) 43.40 0.48 Material(d)

Y 1.27 1.07 33.601d) 35.95 1.14 V

2.52 1.25 43.87(d) 54.84 1.56 X

3.33 1.32 47.15(d) 62.24 1.74 Sum:

196.43 4.92 CFsurv. Weld = X(FF* ARTNDrr)/X(FFW) = (196.43&F)/(4.92) = 39.90F Notes:

(a) f= calculated fluence from capsule X dosimetry analysis results, (x 10'9 nrcm 2, 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(0.2 8- 0.1*-og 0.

(c)

ARTNDt values are the measured 30 ft-lb shift.

(d)

The surveillance weld metal ARTNDT 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.

I I Callaway Plant 18 Revision 5

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-3 Reactor Vessel Beltline Material Unirradiated Toughness Pronerties Material Description Cu (%)

Ni (%)

Initial RTND-a)

Closure Head 0.73 300F(c)

Flange R2704-1 Vessel Flange 0.74 400F(C)

R2701-1 Intermediate Shell 0.05 0.58 40OF Plate R2707-1 Intermediate Shell 0.06 0.61 100F Plate R2707-2 Intermediate Shell 0.06 0.62

-10OF Plate R2707-3 Lower Shell Plate 0.07 0.58 50OF R2708-1 Lower Shell Plate 0.06 0.57 100F R2708-2 Lower Shell Plate 0.08 0.62 20°F R2708-3 Intermediate and 0.04 0.05

-60F Lower Shell Longitudinal Weld Seams(b)

Intermediate to 0.04 0.05

-60$F Lower Shell 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 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 weldwire 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.

(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.

I Callaway Plant 19 Revision 5

a' G

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-4 Fluence (1019 n/cmZ, E > 1.0 MeV) on the Pressure Vessel Clad/Base Metal Interface for Callaway Plant EFPY 00 150 300 450 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 I Callaway Plant 20 Revision 5

1y s

Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-5 Summary of Adjusted Reference Temperature (ART) Values at the 1/4T and 3/4T Locations for 28 EFPY Material 28 EFPY ART(a) 1/ T ART (0F) 3/ T ART (OF)

Intermediate Shell Plate 101 84 R2707-1 Intermediate Shell Plate 81 62 R2707-2 Intermediate Shell Plate 61 42 R2707-3 Lower Shell Plate R2708-1 128(b) 112(b)

Using Surveillance Capsule 92 85 Data Lower Shell Plate R2708-2 81 62 Lower Shell Plate R2708-3 105 90 Intermediate & Lower Shell

-3

-19 Longitudinal Weld Seams 101-124A & 101-142A (900 Azimuth)

Using Surveillance Capsule 8

-4 Data Intermediate & Lower Shell

-3

-19 Longitudinal Weld Seams 101-124B&C and 101-142B&C (210 0 & 3300 Azimuth)

Using Surveillance Capsule 8

-4 Data Intermediate to Lower Shell

-3

-19 Circumferential Weld Seams 101-171 Using Surveillance Capsule 8

-4 Data Notes:

(a)

(b)

ART = Initial RTNDT + ARTNDT + Margin (0F)

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, Revision2, 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.

I Callaway Plant 21 Revision 5

W Be Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT I

Table 5.0-6 Calculation of Adjusted Reference Temperature Values at 28 EFPY for the Limiting Callawav Plant Reactor Vessel Material (Lower Shell Plate R2708-1)

Parameter ART Value Location 1/4 T 3/4 T Chemistry Factor, CF (TF) 44.0 44.0 Fluence, f (1019 n/cmZ)ka) 0.9682 0.3437 Fluence Factor, FF"b' 0.991 0.706 ARTNDT = CF x FF, (TF) 43.60 31.06 Initial RTNDT, I (>F) 50 50 Margin, M (0F)(C) 34 31.06 ART=I+(CFxFF)+M 128 112 (OF)

Per Regulatory Guide 1.99, Rev. 2 Notes:

(a)

(b)

(c)

Fluence, f, is based upon ff(1019 rncm 2, E > 1.0 MeV) = 1.625 at 28 EFPY. The Callaway Plant reactor vessel wall thickness is 8.63 inches at the beltline region.

Fluence Factor, FF, per Regulatory Guide 1.99, Revision 2, is defined as FF = f028-.i0*og Margin is calculated as M = 2(ac2 + OA2)0'5. The standard deviation for the initial RTNDT margin term ai is 0°F since the initial RTNDT is a measured value. The standard deviation for ARTNDT term aA, is 17'F for the plate, except that aA need not exceed 0.5 times the mean value of ARTNDT.

I Callaway Plant 22 Revision 5

z -12

• 1 Figure 14.9 PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-7 RTprs Calculations for Callaway Plant Beltline Region Materials at 35 EFPY(d)

Material Fluence FF CF ART-rs(c)

Margin RTNDT(U)(a)

RTp-rsb)

(1019 (OF)

("F)

(OF7)

(OF7)

(OF7) 2 n/cm2, E

> 1.0 MeV)

I 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 Lower Shell Plate 2.074 1.20 44.0 52.8 34.0 50 137 R2708-1 Psing S/C Data 2.074 1.20 25.0 30 17.0 50 97 Lower Shell Plate 2.074 1.20 37.0 44.4 34.0 10 88 R2708-2 Lower Shell Plate 2.074 1.20 51.0 61.2 34.0 20 115 R2708-3 Inter. & Lower Shell 1.167 1.04 29.7 30.9 30.9

-60 2

Long. Weld Seams 101-124A & 101-142A (900 Azimuth) s-ingS/CData 1.167 1.04 39.9 41.49 28.0

-60 9.5 Inter. & Lower Shell 2.042 1.19 29.7 35.3 35.3

-60 11 Long. Weld Seams 101-124B&C and 101-142B&C (2100

& 3300 Azimuth)

Psing S/C Data 2.042 1.19 39.9 47.48 28.0

-60 15.48 Inter. to Lower Shell 2.074 1.20 29.7 35.6 35.6

-60 11 Circumferential Weld Seams 101-171

• Jsing S/C Data 2.074 1.20 39.9 47.88 28.0

-60 15.88 Notes:

(a)

(b)

(c)

(d)

Initial RTNDT values are measured values RTyrs = RTNDT(U) + Margin + ARTpTs ARTpTs = CF

• FF Projected no. of EFPY at the EOL.

I Callaway Plant 23 Revision 5