ML15062A275

Pressure and Temperature Limits Report, Revision 2

ML15062A275
Person / Time
Site:	Wolf Creek
Issue date:	02/24/2015
From:	Koenig S Wolf Creek
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RA 15-0015
Download: ML15062A275 (27)

Text

WOLF CREEK NUCLEAR OPERATING CORPORATION Steven R. Koenig Manager Regulatory Affairs February 24, 2015 RA 15-0015 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

Subject:

Docket No. 50-482: Wolf Creek Generating Station Pressure and Temperature Limits Report, Revision 2 Gentlemen:

Enclosed is Revision 2 of the Wolf Creek Generating Station (WCGS) Pressure and Temperature Limits Report (PTLR). Revision 2 of the PTLR is being submitted pursuant to Section 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)," of the WCGS Technical Specifications.

This letter contains no commitments. If you have any questions concerning this matter, please contact me at (620) 364-4041.

Sincerely, Steven R. Koen' SRK/rlt Enclosure cc: M. L. Dapas (NRC), w/e C. F. Lyon (NRC), w/e N. F. O'Keefe (NRC), w/e Senior Resident Inspector (NRC), w/e P.O. Box 411 / Burlington, KS 66839 / Phone: (620) 364-8831 An Equal Opportunity Employer M/F/HCNET

Enclosure to RA 15-0015 WOLF CREEK GENERATING STATION - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT, Revision 2 (25 pages)

WOLF CREEK GENERATING STATION - UNIT I PRESSURE AND TEMPERATURE LIMITS REPORT Revision 2

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 Low Temperature Overpressure Protection System 1 3.0 Reactor Vessel Material Surveillance Program 9 4.0 Reactor Vessel Surveillance Data Credibility 9 5.0 Supplemental Data Tables 14 6.0 References 15 Wolf Creek - Unit 1 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT List of Figures Figure Page 2.1-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup 3 Rates of 60 and 1 00°F/hr). Applicable to 54 EFPY (Without Margins for Instrumentation Uncertainty) 2.1-2 Wolf Creek Reactor Coolant System Cooldown Limitations 5 (Cooldown Rates of 0, 20, 40, 60 and 1 OO 0 F/hr) Applicable to 54 EFPY (Without Margins for Instrumentation Uncertainty) 2.2-1 Maximum Allowed PORV Setpoint for the Low Temperature 7 Overpressure Protection System List of Tables 2.1-1 Wolf Creek Heatup Data at 54 EFPY Without Margins for 4 Instrumentation Uncertainty 2.1-2 Wolf Creek Cooldown Data at 54 EFPY Without Margins for 6 Instrumentation Uncertainty 2.2-1 Data Points for Maximum Allowed PORV Setpoint 8 Wolf Creek - Unit I ii Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT 1.0 Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

This PTLR for WCGS has been prepared in accordance with the requirements of Technical Specification (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 Low Temperature Overpressure Protection (LTOP) Systems 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 using a methodology that is in accordance with the NRC-approved methodology specified in Specification 5.6.6 (Ref. 1). In addition, the new Wolf Creek heatup and cooldown P/T limit curves were developed using ASME Code Case N-641, which allows the use of the static crack initiation fracture toughness curve (Kjc).

NRC approval of this methodology was received in Reference 2. NRC acceptance for referencing this methodology was received in Amendment No. 180 (Ref. 3).

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 120OF 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 (Ref. 2)

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

b. A maximum cooldown of 100 0 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 P/T limits for heatup, cooldown, inservice hydrostatic and leak testing, and criticality are specified by Figures 2.1-1 and 2.1-2 (Ref. 5).

2.2 Low Temperature Overpressure Protection (LTOP) System Setpoints (LCO 3.4.12)

The power-operated relief valves (PORVs) shall each have lift settings in accordance with Figure 2.2-1. The LTOP System (Cold Overpressure Mitigation System/PORVs) arming temperature is 368 0 F. These lift setpoints have been developed using the NRC approved methodologies specified in Technical Specification 5.6.6.

Wolf Creek - Unit 1 1 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT 2.2 (continued)

The revised Wolf Creek heatup and cooldown P/T limit curves (Ref. 5) were generated by Westinghouse as part of the testing and analysis of pressure vessel surveillance samples from Capsule X of the WCGS reactor pressure vessel radiation surveillance program. It should be noted that the static crack initiation fracture toughness curve (Ki,), as given in ASME Code Case N-641 and included in Topical Report WCAP-1 4040-A, Revision 4 (Ref. 4) as an option for the development of P/T limit curves, is used as a basis for developing P/T limit curves. The NRC staff accepts this Code Case as an option for the development of P/T limit curves, as the use of optional guidelines for the development of P/T limit curves also meets the regulatory requirements of Appendix G to 10 CFR Part 50 and the guidance provided in SRP Section 5.3.2.

However, the use of Code Case N-641 presently includes a restriction on the maximum allowed PORV setpoint for the LTOP system, which is derived based on the revised heatup and cooldown limit curves. The maximum pressure for the LTOP is 100% of the pressure allowed by the P/T limit curves. This is different from the previous analysis that used the Kia (dynamic crack initiation/crack arrest) fracture toughness curve, along with the use of ASME Code Case N-514, which allows a 10% relaxation of the Appendix G limits below the LTOP enabling temperature.

As a result, the revised PORV setpoint limits for the LTOP system are determined based on 100% of the pressure allowed by the revised P/T limit curves, and the analysis results of the limiting design basis mass and heat input transients. The thermal hydraulic analysis for the mass and heat input transients use the same specialized version of the LOFTRAN code, previously approved by the NRC staff for this type of application.

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) process and instrumentation uncertainties; (2) single failure. To ensure mass and heat input transients more severe than those assumed cannot occur, it is required to lockout both Safety Injection pumps and one centrifugal charging pump (one centrifugal charging pump and the normal charging pump are operational) while in MODES 4, 5, and 6 with the reactor vessel head installed, and limit the heat input due to starting a reactor coolant pump, if secondary temperature is more than 50°F above reactor coolant temperature.

Wolf Creek - Unit 1 2 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R2508-3 LIMITING ART VALUES AT 54 EFPY: 1/4T, 104 0 F 3/4T, 93 0F 2500 2250 Leak Test Limit 2000 Unacceptable Acceptable 1750 Operation A t 0 Operation C. 1500 Heatup Rate ._,_.

60 Deg. F/Hr

'- *

• Critical Limit

- 1250 Heatup Rate __60 Deg. F/Hr 0100 Deg. F/Hr O. * * ._.C ditica

". 1000 -H ----- 00ODeg.l Lim it F/Hr 750,*

500.*

-Boltup ,Temnp.

Criticality Limit based on inservice hyd rostatic test 250 - 60OF temperature (164°F) for the 0 - service period up to 54 EFPY

• " The lower limit for RCS pressure is -14.7 psig 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 2.1-1 Wolf Creek Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°OF/hr) Applicable for the First 54 EFPY (Without Margins for Instrumentation Uncertainty)

Wolf Creek - Unit 1 3 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 2.1-1 Wolf Creek Heatup Limits at 54 EFPY Without Margins for Instrumentation Uncertainty 60°Flhr 6 0 °Flhr Crit. 100°Flhr 100°Flhr Crit. Leak Test Limit Limit Limit Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 60 60

-14.7 621 164 164

-14.7 621 60 60

-14.7 621 164 164

-14.7 621 147 164 2000 2485 I

65 621 164 621 65 621 164 621 70 621 164 621 70 621 164 621 75 621 164 621 75 621 164 621 80 621 164 621 80 621 164 621 85 621 164 621 85 621 164 621 90 621 164 621 90 621 164 621 95 621 164 621 95 621 164 621 100 621 164 621 100 621 164 621 105 621 164 621 105 621 164 621 110 621 164 621 110 621 164 621 115 621 164 621 115 621 164 621 120 621 165 621 120 621 165 621 125 621 170 621 125 621 170 621 130 621 175 621 130 621 175 621 135 621 180 621 135 621 180 621 140 621 180 1074 140 621 180 881 140 1074 185 1129 140 881 185 915 145 1129 190 1190 145 915 190 953 150 1190 195 1257 150 953 195 995 155 1257 200 1333 155 995 200 1043 160 1333 205 1416 160 1043 205 1096 165 1416 210 1507 165 1096 210 1156 170 1507 215 1609 170 1156 215 1222 175 1609 220 1721 175 1222 220 1295 180 1721 225 1846 180 1295 225 1376 185 1846 230 1983 185 1376 230 1466 190 1983 235 2134 190 1466 235 1565 195 2134 240 2301 195 1565 240 1675 200 2301 200 1675 245 1796 205 1796 250 1930 210 1930 255 2078 215 2078 260 2241 220 2241 265 2421 225 2421 Wolf Creek - Unit 1 4 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT MATERIAL PROPERTY BASIS LIMITING MATERIAL: LOWER SHELL PLATE R2508-3 LIMITING ART VALUES AT 54 EFPY: 1/4T, 104 0 F 3/4T, 93 0F 2500 2250 Unacceptable 2000 Operation 1750 Acceptable V) 1500 Operation Cooldown Rates, °F/Hr 1250 U) steady-state I- -20 M. 1000 -40

-60

-100 750 500 Boltup Temperature, 607F 250 0 The lower limit for RCS pressure is -14.7 psig

-. .- I . . . .-Ir. . . 1, 11 - 1 pl l f l l l 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F)

FIGURE 2.1-2 Wolf Creek Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 1 00°F/hr) Applicable for the First 54 EFPY (Without Margins for Instrumentation Uncertainty)

Wolf Creek - Unit 1 5 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 2.1-2 Wolf Creek Cooldown Limits at 54 EFPY Without Margins for Instrumentation Uncertainty Steady State 2 0 0F/hr 40 °Flhr 60 °F/hr 100°F/hr Temp. Press. Temp. Press. Temp. Press. Temp. Press. Temp. Press.

(OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) (OF) (psig) 60 60

-14.7 621 60 60

-14.7 621 60 60

-14.7 621 60 60

-14.7 621 60 60

-14.7 588 I

65 621 65 621 65 621 65 621 65 614 70 621 70 621 70 621 70 621 70 621 75 621 75 621 75 621 75 621 75 621 80 621 80 621 80 621 80 621 80 621 85 621 85 621 85 621 85 621 85 621 90 621 90 621 90 621 90 621 90 621 95 621 95 621 95 621 95 621 95 621 100 621 100 621 100 621 100 621 100 621 105 621 105 621 105 621 105 621 105 621 110 621 110 621 110 621 110 621 110 621 115 621 115 621 115 621 115 621 115 621 120 621 120 621 120 621 120 621 120 621 125 621 125 621 125 621 125 621 125 621 130 621 130 621 130 621 130 621 130 621 135 621 135 621 135 621 135 621 135 621 140 621 140 621 140 621 140 621 140 621 140 1387 140 1387 140 1387 140 1387 140 1387 145 1469 145 1469 145 1469 145 1469 145 1469 150 1560 150 1560 150 1560 150 1560 150 1560 155 1660 155 1660 155 1660 155 1660 155 1660 160 1771 160 1771 160 1771 160 1771 160 1771 165 1893 165 1893 165 1893 165 1893 165 1893 170 2028 170 2028 170 2028 170 2028 170 2028 175 2178 175 2178 175 2178 175 2178 175 2178 180 2343 180 2343 180 2343 180 2343 180 2343 Wolf Creek - Unit 1 6 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT 2500-2000  !

1500 -

-PORV Setpoint Limits Z

O -PORV #1 Breakpoints c-a -M PORV #2 Breakpoints o 1000 500 -_

0 50 100 150 200 250 300 350 400 450 MEASURED RTD TEMPERATURE (DEG F) 2 RCPs running below 100 OF 4 RCPs running above 100 OF FIGURE 2.2-1 Maximum Allowed PORV Setpoint for the Low Temperature Overpressure Protection System Wolf Creek - Unit I 7 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 2.2-1 Maximum Allowed PORV Setpoints/Breakpoints Temperature Pressure (psig)

(OF) Max. Allowed PORV #1 PORV #2 60 425 415 425 78 425 415 425 88 425 -- --

118 425 415 425 158 425 415 425 168 529 -- -

208 529 460 525 218 540 -- --

268 650 570 650 318 800 680 800 343 910 -- --

368 1127 680 800 418 2350 -- --

425 -- 2350 2350 Wolf Creek - Unit 1 8 Revision 2

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 WCGS Updated Safety Analysis Report. The withdrawal schedule is presented in USAR Table 5.3-11. The surveillance capsule reports are as follows:

1. WCAP-1 1553, August 1987, "Analysis of Capsule U from the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Programs."

2. WCAP-1 3365, Revision 1, April 1993, "Analysis of Capsule Y from the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Programs."

3. WCAP-1 5078, Revision 1, August 1998, "Analysis of Capsule Vfrom the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Programs."

4. WCAP-1 6028, Revision 0, March 2003, "Analysis of Capsule X from the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Programs."

Note: The last W and Z capsules were withdrawn during Refueling Outage 14 (Spring 2005) and are now stored in the spent fuel pool and will be retained for the extended licensed operating period. During Refueling Outage 14, Ex-Vessel Dosimetry was installed at Wolf Creek to provide continuous monitoring of the beltline region of the reactor vessel.

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 there has been four surveillance capsules removed from the Wolf Creek reactor vessel and tested. 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 Regulatory Guide 1.99, Revision 2, to the Wolf Creek reactor vessel surveillance data and determine if the Wolf Creek surveillance data is credible.

Wolf Creek - Unit 1 9 Revision 2

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 Wolf Creek reactor vessel consists of the following beltline region materials:

• Intermediate shell plate R2005-1, 2, 3

• Lower shell plate R2508-1, 2, 3

" Intermediate & Lower Shell Longitudinal Weld Seams (Heat # 90146),

• Intermediate & Lower Shell Circumferential Weld Seams (Heat # 90146),

Per WCAP-10015, the Wolf Creek surveillance program was based on ASTM E185-79.

When the surveillance program material was selected it was believed that copper and phosphorus were the elements most important to embrittlement of reactor vessel steels.

Lower shell plate R2508-3 had the highest initial RTNDT and the lowest initial USE of all plate materials in the beltline region. In addition, lower shell plate R2508-3 had approximately the same copper and phosphorous content as the other beltline plate materials. Therefore, based on the highest initial RTNDT and lowest initial upper shelf energy, lower shell plate was chosen for the surveillance program.

The weld material in the Wolf Creek surveillance program was made of the same wire as all the reactor vessel beltline welds, thus it was chosen as the surveillance weld material.

Hence, this criterion is met for the Wolf Creek reactor vessel.

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-1 6028, Revision 0, March 2003, "Analysis of Capsule X from the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Program," (Ref. 7).

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 Wolf Creek surveillance materials unambiguously. Hence, the Wolf Creek surveillance program meets this criterion.

Wolf Creek - Unit 1 10 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT 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 28°F for welds and 170 F for base metal. 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 28 0 F for welds and less than 17 0 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. In addition, the recommended NRC methods for determining credibilitywill be followed. The NRC methods were presented to the industry at a meeting held by the NRC on February 12 and 13, 1998. At this meeting the NRC presented five cases. Of the five cases, Case 1 ("Surveillance data available from plant but no other source") most closely represents the situation listed above for the Wolf Creek surveillance weld metal. Note, for the plate materials, the straight forward method of Regulatory Guide 1.99, Revision 2 will be followed.

Wolf Creek - Unit 1 11 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-1 Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule Data Material Capsule Capsule FF(b) ARTNDT(c) FF*ARTNDT FF 2 f(a)

Lower Shell U 0.316 0.684 36.46 24.94 0.468 Plate R2508-3 Y 1.19 1.05 16.03 16.83 1.10 (Longitudinal) V 2.22 1.22 52.03 63.48 1.49 X 3.49 1.33 61.06 81.21 1.77 Lower Shell U 0.316 0.684 23.79 16.27 0.468 Plate R2508-3 Y 1.19 1.05 35.39 37.16 1.10 (Transverse) V 2.22 1.22 54.53 66.53 1.49 X 3.49 1.33 53.96 71.77 1.77 SUM: 378.19 9.656 CFR25 08-3 = X(FF

• ARTNDT) + X( FF 2) = (378.19) + (9.656) = 39.1-F Surveillance U 0.316 0.684 27.21 18.612 0.468 Weld Material(d) Y 1.19 1.05 45.09 47.34 1.10 V 2.22 1.22 46.33 56.49 1.49 X 3.49 1.33 68.36 90.92 1.77 SUM: 213.362 4.828 CF Surv. Weld = YI(FF

• ARTNDT) + E( FF 2) = (213.362) - (4.828) = 44.1°F Notes:

(a) f = Calculated Fluence (1019 n/cm 2 , E > 1.0 MeV). These values were re-evaluated as part of the capsule X analysis. (See Appendix D of WCAP-1 6028, Revision 0)

(b) FF = fluence factor = f (c) ARTNDT values are the measured 30 ft-lb shift values taken from WCAP-1 6028, [OF].

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

Wolf Creek - Unit 1 12 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT Table 4.0-2 Wolf Creek Surveillance Capsule Data Scatter about the Best-Fit Line for Surveillance Forgin Materials Material Capsule CF FF Measured Predicted Scatter <17 0 F (Base ARTNDT(a) ARTNDT(b) ARTNDT Metals)

(°F) <17 0F (Weld)

Lower Shell U 39.1 0.684 36.46 26.74 9.72 Yes Plate Y 39.1 1.05 16.03 41.06 -25.03 No R2508-3 V 39.1 1.22 52.03 47.70 4.33 Yes (Longitudinal) X 39.1 1.33 61.06 52.00 9.06 Yes Lower Shell U 39.1 0.684 23.79 26.74 -2.95 Yes Plate Y 39.1 1.05 35.39 41.06 -5.67 Yes R2508-3 V 39.1 1.22 54.53 47.70 6.83 Yes (Transverse) X 39.1 1.33 53.96 52.00 1.96 Yes Surveillance U 44.1 0.684 27.21 30.16 -2.95 Yes Program Y 44.1 1.05 45.09 46.31 -1.22 Yes Weld Metal V 44.1 1.22 46.33 53.80 -7.47 Yes X 44.1 1.33 68.36 58.65 9.71 Yes Notes:

(a) Based on measured Charpy data plotted with CVGRAPH 4.1.

(b) Best estimate ARTNDT = CF

• FF, where the CF is based on the measured surveillance data.

Table 4.0-2 indicates that only one data point falls outside the +/- 1a of 1 70F scatter band for the lower shell plate R2508-3 surveillance data. One out of 8 data points is still considered credible. No weld data point fall outside the +/- 1 a of 28 scatter band for the surveillance weld data, therefore the weld data is also credible per the third criterion.

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 0 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 0 F. Hence this criteria is met.

Wolf Creek - Unit 1 13 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT 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 Wolf Creek surveillance program does not contain correlation monitor material.

Therefore, this criterion is not applicable to the Wolf Creek surveillance program.

Conclusion:

Based on the proceding responses to all five criteria of Regulatory Guide 1.99, Revision 2, Section B and 10 CFR 50.61, the Wolf Creek surveillance plate and weld data is credible.

5.0 Supplemental Data Tables Table 5.0-1 Comparison of Wolf Creek Surveillance Material 30-ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Table 5.0-2 Calculation of Chemical Factors Using Surveillance Capsule Data Table 5.0-3 Wolf Creek Reactor Vessel Beltline Unirradiated Material Properties Table 5.0-4 Summary of the Peak Pressure Vessel Neutron Fluence Values at 54 EFPY used for the Calculation of Adjusted Reference Temperature (ART) Values Table 5.0-5 Summary of Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the 1/4-T and %-T Locations for 54 EFPY Table 5.0-6 Calculation of the Adjusted Reference Temperatures (ARTs) at 54 EFPY for the Limiting Reactor Vessel Material (Lower Shell Plate R-2508-3)

Table 5.0-7 RTPTS Calculation for Wolf Creek Beltline Region Material at Life Extension (54 EFPY)

Wolf Creek - Unit 1 14 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT 6.0 References

1. Technical Specification 5.6.6, "Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).-

2. NRC letter dated February 27, 2004, Final Safety Evaluation for Topical Report WCAP-14040, Revision 3, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," May 2004.

3. License Amendment No. 180, dated January 27, 2009, from Balwant K. Singal, USNRC, to Rick A. Muench, WCNOC.

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

5. WCAP-1 6029, Revision 0, "Wolf Creek Heatup and Cooldown Limit Curves for Normal Operation," May 2003.

6. WCAP-1 6030, Revision 0, "Evaluation of Pressurized Thermal Shock for Wolf Creek,"

May 2003.

7. WCAP-16028, Revision 0, "Analysis of Capsule X from the Wolf Creek Nuclear Operating Corporation, Wolf Creek Reactor Vessel Radiation Surveillance Program," March 2003.

8. WCAP-1 5080, Revision 1, "Evaluation of Pressurized Thermal Shock for Wolf Creek," September 1998.

Wolf Creek - Unit I 15 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-1 Comparison of Wolf Creek Surveillance Material 30 ft-lb Transition Temperature Shifts and Upper Shelf Energy Decreases with Regulatory Guide 1.99, Revision 2, Predictions Fluence(d) 30 ft-lb Transition Upper Shelf Energy Materials Capsule (x 1019 Temperature Shift Decrease n/cm2, Predicted Measured Predicted Measured E>1.0 MeV) (OF) (a) (OF) (b) (%) (a) (%) (c)

Lower Shell Plate U 0.316 34.88 36.46 14.5 2 R2508-3 Y 1.19 53.55 16.03 20 11 V 2.22 62.22 52.03 23 13 (Longitudinal) X 3.49 67.83 61.06 25 4 Lower Shell Plate U 0.316 34.88 23.79 14.5 0 R2508-3 Y 1.19 53.55 35.39 20 0 V 2.22 62.22 54.53 23 6 (Transverse) X 3.49 67.83 53.96 25 0 Surveillance U 0.316 33.24 27.21 16 8 Program Y 1.19 51.03 45.09 22 6 V 2.22 59.29 46.33 25 11 Weld Metal X 3.49 64.64 68.36 28 7 Heat Affected U 0.316 --- 58.41 --- 13 Zone Y 1.19 --- 12.98 --- 0 V 2.22 --- 55.91 --- 0 Material X 3.49 69.66 --- 16 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 E185-82.

(d) The fluence values presented here are the calculated values, not the best estimate values.

Wolf Creek - Unit 1 16 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-2 Calculation of Chemistry Factors using Wolf Creek Surveillance Capsule Data Material Capsule Capsule FF(b) ARTNdT(c) FF*ARTNDT FF 2 f(a)

Lower Shell U 0.316 0.684 36.46 24.94 0.468 Plate R2508-3 Y 1.19 1.05 16.03 16.83 1.10 (Longitudinal) V 2.22 1.22 52.03 63.48 1.49 X 3.49 1.33 61.06 81.21 1.77 Lower Shell U 0.316 0.684 23.79 16.27 0.468 Plate R2508-3 Y 1.19 1.05 35.39 37.16 1.10 (Transverse) V 2.22 1.22 54.53 66.53 1.49 X 3.49 1.33 53.96 71.77 1.77 SUM: 378.19 9.656 CFR2508-3 = Z(FF */ARTNDT) + Y( FF 2) = (378.19) + (9.656) = 39.1*F Surveillance U 0.316 0.684 18.26 (27.21) 12.49 0.468 Weld Material(d) Y 1.19 1.05 30.26 (45.09) 31.77 1.10 V 2.22 1.22 31.09 (46.33) 37.93 1.49 X 3.49 1.33 45.90 (68.36) 61.05 1.77 SUM: 143.24 4.828 CF Smr. Weld = Z(FF

• ARTNDT) + E( FF 2) = (143.24) ÷ (4.828) = 29.7*F Notes:

(a) f = Calculated Fluence (1019 n/cm 2, E > 1.0 MeV). These values were re-evaluated as part of the capsule X analysis. (See Appendix D of WCAP-16028, Revision 0)

(b) FF = fluence factor = fO.28-0.1Iog f)

(c) ARTNDT values are the measured 30 ft-lb shift values given in the Capsule X analysis report, WCAP-16028, [OF].

(d) The Surveillance Weld ARTNDT values have been adjusted by a ratio of 0.671 (CFvw +

CFsw = 32.6 ÷ 48.6). The pre-adjusted values are in parenthesis.

Wolf Creek - Unit 1 17 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT Table 5.0-3 Wolf Creek Reactor Vessel Beltline Unirradiated Material Properties Material Description Cu (%)(a) Ni(%)(a) Initial RTNDT(a)

Closure Head Flange R2504-1 --- 0.66 20°F(d)

Vessel Flange R2501-1 --- 0.70 20°F(d)

Intermediate Shell Plate R2005-1 0.04 0.66 -20°F Intermediate Shell Plate R2005-2 0.04 0.64 -20°F Intermediate Shell Plate R2005-3 0.05 0.63 -20°F Lower Shell Plate R2508-1 0.09 0.67 0°F Lower Shell Plate R2508-2 0.06 0.64 10°F Lower Shell Plate R2508-3 0.08(c) 0.58(c) 40°F Intermediate & Lower Plate Longitudinal 0.04(c) 0.09(c) -50OF Weld Seams (Heat # 9 0 1 4 6 )(b)

Intermediate to Lower Shell Plate 0.04(c) 0.09(c) -50OF Circumferential Weld (Heat # 9 0 1 4 6 )(b)

Surveillance Weld (Heat # 9 0 1 4 6 )(b) 0.06(c) 0.17(c) - - -

Notes:

(a) Based on measured data.

(b) All vessel beltline welds seams were fabricated with weld wire heat number 90146. The intermediate to lower shell girth weld seam, 101-171, was fabricated with Flux Type 124 Lot # 1061. The intermediate and lower shell longitudinal weld seams (101-124A,B, C &

101-142A,B,C) were fabricated with Flux Type 0091 Lot # 0842. The surveillance weld was fabricated with weld wire 90146, Flux Type 124 Lot # 1061. The surveillance weld metal was made with the same weld wire heat as all the vessel beltline weld seams and is therefore representative of all the beltline weld seams.

(c) Updated from previous PTS Report (WCAP-1 5080 (Ref. 8)) based on new chemical analysis presented in WCAP-16028.

(d) These values are used for considering requirements for the heatup/cooldown curves. Per the methodology given in WCAP-14040-A (Ref. 4), the minimum boltup temperature is 600 F.

Wolf Creek - Unit 1 18 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 5.0-4 Summary of the Peak Pressure Vessel Neutron Fluence Values at 54 EFPY used for the Calculation of Adjusted Reference Temperature (ART) Values (n/cm 2, E > 1.0 MeV)

Material Surface(a) 1/4T (b) 3/4T(b)

Intermediate Shell Plate rmediat 3.51 x 1019 2.09 x 1019 7.42x 1018 R2005-1 Intermediate Shell Plate rmediat 3.51 x 1019 2.09 x 1019 7.42x 1018 R2005-2 Intermediate Shell Plate rmediat 3.51 x 1019 2.09 x 1019 7.42x 1018 R2005-3 Lower Shell Plate R2508-1 3.51 x 1019 2.09 x 1019 7.42x 1018 Lower Shell Plate R2508-2 3.51 x 1019 2.09 x 1019 7.42x 1018 Lower Shell Plate R2508-3 3.51 x 1019 2.09 x 1019 7.42x 1018 Intermediate & Lower Shell 3.08 x 1019 1.84x 1019 6.52x 101 "

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

Intermediate & Lower Shell 3.08 x 10'9 1.84x 1019 6.52x 1018 Longitudinal Weld Seam 101-124B.C &101-142B,C (2100 & 3300 Azimuth)

Intermediate to Lower Shell 3.51 x 1019 2.09 x 1019 7.42x 1018 Plate Circumferential Weld Seam 101 -171 Notes:

(a) The fluence was taken from the peak azimuthal location. (see Table 2 of WCAP-1 6030)

(b) Attenuation of the fluence at the specific depth is calculated by the formula: f (depth x)= fsurace

• e(°24x), where x inches (vessel beltline thickness is 8.63 inches) is the depth into the vessel wall measured from the vessel clad/base metal interface.

Wolf Creek - Unit 1 19 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 5.0-5 Summary of Adjusted Reference Temperatures (ARTs) for the Reactor Vessel Beltline Materials at the '/4-T and %/4-TLocations for 54 EFPY Material__ _ _ _ _ _ _ _ _

54 EFPY ART(a)

RG 1.99 Rev. 2  %-T ( 0F)  %-T (fF)

Method Intermediate Shell Plate Position 11 42 28 R2005-1 Intermediate Shell Plate R2005-2Position 1-1 42 28 R2005-2 Intermediate Shell Plate Position 1-1 51 37 R2005-3 Lower Shell Plate R2508-1 Position 11 104 87 Lower Shell Plate R2508-2 Position 11 88 78 Lower Shell Plate R2508-3 Position 11 135 121 104(b) 93(b)

Position 2-1 Intermediate & Lower Plate Position 11 28 10 Long. Weld Seams (Heat # 90146) Position 2-1 14 5 Intermediate & Lower Plate Position 11 28 10 Long. Weld Seams (Heat # 90146) Position 2-1 14 5 Notes:

(a) ART = Initial RTNDT + ARTNDT + Margin (OF)

(b) These ART values are used to generate the heatup and cooldown curves.

Wolf Creek - Unit 1 20 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 5.0-6 Calculation of Adjusted Reference Temperatures (ARTs) at 54 EFPY for the Limiting Wolf Creek Reactor Vessel Material (Lower Shell Plate R 2508-3)

Parameter ART Value Location 1/4-T 1/  %-T Chemistry Factor, CF (OF) 39.1 39.1 Fluence - 1019 n/cm 2 (E > 1.0 MeV), f(a) 2.09 0.742 Fluence Factor, FF(b) 1.201 0.916 ARTNDT = CF x FF, (OF) 46.959 35.816 Initial RTNDT, I (°F) 40 40 Margin, M (OF)(C) 17 17 ART= I +(CFxFF)+ M (OF) 104 93 per Regulatory Guide 1.99, Rev. 2 Notes:

(a) Fluence, f, is based upon fs,, (1019 n/cm 2, E > 1.0 MeV) = 2.09 at 54 EFPY. The Wolf Creek reactor vessel wall thickness is 8.625 inches at the beltline region.

(b) Fluence Factor (FF) per Regulatory Guide 1.99, Revision 2, is defined as FF = fO.28 -0.1olog.f)

(c) Margin is calculated as M = 2(ca2+ GA2) 0 5 . The standard deviation for the initial RTNDT margin term ag, is 0°F since the initial RTNOT is a measured value. The standard deviation for ARTNDT term GA, is 170F for the plate, except that GA need not exceed the 0.5 times the mean value of ARTNDT.

Wolf Creek - Unit 1 21 Revision 2

PRESSURE AND TEMPERATURE LIMITS REPORT TABLE 5.0-7 RTpTs Calculation for Wolf Creek Beltline Region Materials at Life Extension (54 EFPY)

Material Fluence, f FF(b) CF ARTPTS Margin RTNDT(U) RTPTS (n/cm 2, E>1.0 (OF) (c) (OF) (d) (OF) (e) (OF) (f) (OF) (g)

MeV)(a)

Intermediate Shell Plate R2005-1 3.51 x 1019 1.33 26.0 34.58 34.00 -20 49 Intermediate Shell Plate R2005-2 3.51 x 1019 1.33 26.0 34.58 34.00 -20 49 Intermediate Shell Plate R2005-3 3.51 x 1019 1.33 31.0 41.23 34.00 -20 55 Lower Shell Plate R2508-1 3.51 x 1019 1.33 58.0 77.14 34.00 0 111 Lower Shell Plate R2508-2 3.51 x 10'9 1.33 37.0 49.21 34.00 10 93 Lower Shell Plate R2508-3 3.51 x 1019 1.33 51.0 67.83 34.00 40 142

= Using Surv. Capsule Data 3.51 x 10"' 1.33 39.1 52.00 17.00 40 109 Intermediate & Lower Plate Long. 3.51 x 1019 1.33 32.6 43.36 43.36 -50 37 Weld Seams (Heat # 90146)

=> Using Surv. Capsule Data 3.51 x 1019 1.33 29.7 39.50 28.00 -50 18 Intermediate & Lower Plate Long. 3.51 x 1019 1.33 32.6 43.36 43.36 -50 37 Weld Seams (Heat # 90146)

= Using Surv. Capsule Data 3.51 x 1019 1.33 29.7 39.50 28.00 -50 18 Notes:

(a) The fluence was taken from the peak azimuthal location (See Table 2 of Ref. 6).

(b) FF = f.28-0.1*109f); where f is the clad/base metal interface fluence.

(c) Chemistry Factor is taken from Table 5 of Ref. 6.

(d) ARTPTS = CF

• FF (e) Margin = 2*(au 2 2

+CFA )

1 2

/ .

(f) Initial RTNDT values are measured values.

(g) RTPTS = RTNDT(U) + ARTPTS + Margin (OF) (This value was rounded per ASTM E29, using the "Rounding Method")

Wolf Creek - Unit 1 22 Revision 2