CP-200800612, Commanche Peak, Units 1 & 2 - Transmittal of Revised Pressure and Temperature Limits Report

Commanche Peak, Units 1 & 2 - Transmittal of Revised Pressure and Temperature Limits Report

ML081230101
Person / Time
Site:	Comanche Peak
Issue date:	04/24/2008
From:	Blevins M, Madden F Luminant Generation Co, Luminant Power
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-200800612, ERX-07-003, Rev 1, TXX-08072
Download: ML081230101 (20)

Text

Mike Blevins Luminant Power Executive Vice President P 0 Box 1002

& Chief Nuclear Officer 6322 North FM 56 Mike.Blevins@Luminant.com Glen Rose, TX 76043 Luminant T 254 897 5209 C 817 559 9085 F 254 897 6652 CP-200800612 Ref. # Tech. Spec. 5.6.6 Log # TXX-08072 April 24, 2008 /

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION DOCKET NOS. 50-445 AND 50-446 PRESSURE AND TEMPERATURE LIMITS REPORT

Dear Sir or Madam:

Pursuant to Technical Specification 5.6.6, Luminant Generation Company LLC (Luminant Power) hereby submits the revised Pressure and Temperature Limits Report for Comanche Peak. This revision was adopted on February 8, 2008.

This communication contains no new licensing basis commitments regarding Comanche Peak Units 1 and 2. Should you have any questions, please contact Mr. Robert Kidwell at (254) 897-5310.

Sincerely, Luminant Generation Company LLC Mike Blevins By: .

I*re~d--. Madden Director, Oversight & Regulatory Affairs Enclosure Comanche Peak Pressure and Temperature Limits Report (PTLR) c- E. E. Collins, Region IV B. K. Singal, NRR Resident Inspectors, Comanche Peak A member of the STARS (Strategic Teaming and Resource Sharing) Alliance Callaway - Comanche Peak

• Diablo Canyon

• Palo Verde

• South Texas Project

• Wolf Creek

ERX-07-003, Revision 1 COMANCHE PEAK NUCLEAR POWER PLANT (CPNPP)

PRESSURE AND TEMPERATURE LIMITS REPORT (APPLICABLE UP TO 36 EFPY)

December 2007 Prepared: Date:

Kevin Roland Principal Engineer, Westinghouse Electric Co.

Reviewed: /nZ6r C S'Asz Date: 12142zo7-Hugo C. da Silva Principal Engineer, Westinghouse Electric Co.

Approved: _ _ _ _ Date:

W.M.Boatwrigh o Manager, Westinghouse Engineering Services - Texas ERX-07-003, Rev. 1

DISCLAIMER The information contained in this report was prepared for the specific requirement of Luminant, and may not be appropriate for use in situations other than those for which it was specifically prepared. LUMINANT PROVIDES NO WARRANTY HEREUNDER, EXPRESSED OR IMPLIED, OR STATUTORY, OF ANY KIND OR NATURE WHATSOEVER, REGARDING THIS REPORT OR ITS USE, INCLUDING BUT NOT LIMITED TO ANY WARRANTIES ON MERCHANTABILITY OR FITNESS FOR A PARTICULAR PURPOSE.

By making this report available, Luminant does not authorize its use by others, and any such use is forbidden except with the prior written approval of Luminant. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall Luminant have any liability for any incidental or consequential damages of any type in connection with the use, authorized or unauthorized, of this report or of the information in it.

ERX-07-003, Rev. 1

TABLE OF CONTENTS DISCLAIMER........................................................... ii TA BLE O F C O NT E NT S ............................................................................................................. iii LIST O F TA BLE S .................................................................................................................... iv LIST O F FIG UR ES ............................................................................................ V SECTION PAGE 1.0 INT R O D UC T IO N .......................................................................................................... 1 2 .0 O P E RA T ING LIM ITS .................................................................................................... 2 2.1 RCS Temperature Rate-of-Change Limits (LCO 3.4.3).................................... 4 2.2 P/T Limits for Heatup, Cooldown, Inservice Leak & Hydrostatic Testing, and C ritica lity (LC O 3 .4 .3 ) ........................................................................................ 4 2.3 LTOP System Setpoints (LCO 3.4.12) .............................................................. 6 2.4 Reactor Vessel Material Surveillance Program ................................................. 6 3 .0 R E F E R E NC E S ............................................................................................................. 7 iii ,, 7 ERX-07-003,F Rev. 1

LIST OF TABLES TABLE PAGE 2-1 Limiting Materials and Reference Temperatures for CPNPP Unit 1 and Unit 2 R eacto r V essels ..................................................................................................... .. 8 2-2 Calculation of Chemistry Factor Values for Unit 1 Surveillance Capsule Test R e s u lts ........................................................................................................................ 9 2-3. Calculation of Chemistry Factor Values for Unit 2 Surveillance Capsule Test R e s u lts ...................................................................................................................... 10 2-4 PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 1 with Delta-76 Steam Generators - Applicable Up To 36 EFPY ...................................... 11 2-5 PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 2 with D5 Steam Generators - Applicable Up To 36 EFPY ................................................. 11 2-6 Unit 2 Reactor Vessel Material Surveillance Program - Withdrawal Schedule ........ 12 iv ERX-07-003, Rev. 1

LIST OF FIGURES FIGURE' PAGE 2-1 Reactor Coolant System Heatup Limitations - Applicable Up To 36 EFPY .............. 13 2-2 Reactor Coolant System Cooldown Limitations - Applicable Up To 36 EFPY .......... 14 v

ERX-07-003, Rev. 1

1.0 INTRODUCTION

This report presents the Reactor Coolant System (RCS) Pressure and Temperature (P/T) limits for Comanche Peak Nuclear Power Plant (CPNPP) Unit I and Unit 2 in accordance with the requirements of Technical Specification 5.6.6. A description of the Low Temperature Overpressure Protection (LTOP) System power-operated relief valve (PORV) setpoints is also provided in this report. In addition, the requirements of the reactor vessel material surveillance program are discussed.

The following two Technical Specification Limiting Conditions of Operation (LCO) are addressed in this report:

LCO 3.4.3 RCS Pressure and Temperature (P/T) Limits LCO 3.4.12 Low Temperature Overpressure Protection (LTOP) System The analytical methods used to determine the RCS pressure and temperature limits are described'in Reference 1. The methods used to develop the LTOP System PORV setpoints are also described in Reference 1.

This report covers CPNPP Unit 1 and Unit 2 operation for 36 Effective Full Power Years (EFPY).

Note that Revision 0 of this PTLR was submitted to the NRC in support of Operating License Amendment 132. The NRC reviewed the submittal and has determined that the PTLR meets the requirements set forth in GL 96-03 for plant-specific PTLRs; therefore, it is acceptable for use at CPNPP. In Revision 1, the LTOP System PORV setpoints for CPNPP Unit 2 with the D5 steam generators are changed to those of Table 14 of Reference 5.

ERX-07-003, Rev. 1

2.0 OPERATING LIMITS RCS P/T Limits The RCS PIT limits presented in this report consist of the RCS (except the pressurizer) temperature rate-of-change limits and P/T limits during heatup, cooldown, inservice leak and hydrostatic testing, and criticality. The P/T limits for both CPNPP units are based on the approved methodology presented in Reference 1.

The RCS P/T limits are based on the results of the evaluations of the most recently analyzed reactor vessel specimen capsules as presented in References 2 and 3 for Units 1 and 2, respectively. The more limiting material is used to develop RCS P/T limits that bound both CPNPP units.

The RCS P/T limits calculated for selected heatup and cooldown rates for CPNPP Unit 1 and Unit 2 are extracted from Reference 4.

LTOP System The LTOP System acts as a backup to the reactor operators to mitigate RCS pressurization transients at low temperatures so the integrity of reactor coolant pressure boundary (RCPB) is not compromised by violating the pressure and temperature limits of Appendix G of 10 CFR 50..

The reactor vessel is the limiting RCPB component for demonstrating such protection.

The LTOP System provides reduced setpoints for the pressurizer Power-Operated Relief Valves (PORVs) as a function of the RCS temperature. The methodology used to select the setpoint pressures is described in Reference 1. Allowances for instrument uncertainties have been included in the development of these setpoints.

The LTOP System PORV setpoints for CPNPP Unit 1 (with the A76 steam generators) are extracted from Reference 6 and those for CPNPP Unit 2 (with the D5 steam generators) are extracted from Reference 5.

2 ERX-07-003, Rev. 1

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reduction in ductility that results from neutron radiation manifests itself as an increase in the Nil Ductility Reference Temperature (RTNDT) and a reduction of the upper-shelf energy of reactor vessel beltline materials, including welds. At CPNPP, these quantities were predicted at 36 EFPY using the methods of WCAP-14040-NP-A, Revision 4 [1]. The predictions showed that the materials in the Unit 1 and Unit 2 reactor vessels responded similarly to neutron irradiation but at 36 EFPY, the plate material in the Unit 1 beltline was most limiting. Forecast properties of the limiting material were used to establish P/T limits for heatup and cooldown curves and LTOP setpoints.

The reactor vessel specimen capsules are withdrawn when the projected neutron fluence would exceed one-times the projected end-of-life vessel fluence and less than two-times the projected end-of-life vessel fluence, in accordance with Reference 7.

For Unit 1, the required specimen capsules U and Y have been withdrawn and evaluated [2].

The third required specimen capsule, Capsule X, was withdrawn during 1RF11 in the fall of 2005, with a fluence within the range of one-times to two-times the 52 EFPY Peak Fluence [2],

but has not yet been evaluated. Two of the standby capsules (Capsules V and W) were withdrawn' in 1RF09 and stored for later evaluation, if necessary. The third standby capsule was withdrawn during 1RF1 1 in the fall of 2005 and stored for later evaluation, if necessary.

Because all reactor vessel surveillance capsules have been withdrawn and stored, a capsule removal schedule is not required for Unit 1.

For Unit 2, the required specimen capsules U and X have been withdrawn and evaluated [3].

The third required specimen capsule, Capsule W, is scheduled to be withdrawn during 2RF1 1 in the spring of 2010, with a fluence within the range of one-times to two-times the 54 EFPY Peak Fluence [3]. The schedule for the third capsule withdrawal differs from the specific recommendations contained in Reference 3, but satisfies the requirements of Reference 7 based on an expected end-of-life fluence corresponding to the 54 EFPY Peak Fluence. Two of the standby capsules (Capsules V and Y) were withdrawn in 2RF07 and stored for later evaluation, if necessary. The third standby capsule is scheduled to be withdrawn during 2RF1 1 in the spring of 2010 and stored for later evaluation, if necessary.

3 ERX-07-003, Rev. 1

2.1 RCS Temperature Rate-of-Change Limits (ILCO 3.4.3) 2.1.1 Maximum Heatup Rate The RCS heatup rate limit is 100°F in any 1-hour period.

2.1.2 Maximum Cooldown Rate The RCS cooldown rate limit is 100OF in any 1-hour period.

2.1.3 Maximum Temperature Change During Inservice Leak and Hydrostatic Testing During inservice leak and hydrostatic testing operations above the heatup and cooldown limit curves, the RCS temperature change limit is 10OF in any 1-hour period.

2.2 P/T Limits for Heatup, Cooldown, Inservice Leak & Hydrostatic Testing, and Criticality (LCO 3.4.3)

The limiting materials and adjusted reference temperatures at the 1/4t and 3/4t locations for each unit's reactor vessel are extracted from Reference 4 and are presented in Table 2-1. These values are based on the evaluation of two surveillance capsule specimens for each unit which include evaluations of the credibility of data per Regulatory Guide 1.99, Revision 2. All surveillance data for Unit 1 is. credible. For Unit 2, the surveillance p!ate data (for the intermediate shell plate R3807-1) is not credible, while the surveillance weld data is credible.

The limiting reference temperatures for pressurized thermal shock (RTPTS) values for each unit's reactor vessel were previously docketed in accordance with 10CFR50.61 and are extracted from References 8 and 9 for presentation in Table 2-1. Analyses of the withdrawn surveillance capsules from the Unit 1 and Unit 2 reactor vessels have confirmed the similarity between the two vessels in irradiated and non-irradiated material properties. The results of these surveillance capsule evaluations have confirmed that the early projections for CPNPP vessel materials were conservative. In addition, the majority of the irradiation-induced shift in vessel material properties occurs early in life.

Therefore, with substantial-margin to the RTPTS screening criteria, the conservative fluence projections for the CPNPP vessel materials, and the 4

ERX-07-003, Rev. 1

absence of a significant change in the projected values of RTPTS, the Pressurized Thermal Shock reports have not been revised.

2.2.1 Calculation of Chemistry Factors using Surveillance Capsule Test Results Best-estimate, plant-specific, copper and nickel weight percent values were used to calculate the chemistry factors in accordance with Regulatory Guide 1.99, Revision 2. Additionally, surveillance capsule data is available for two capsules already removed from both Comanche Peak reactor vessels; this data was used to calculate chemistry factor values per Position 2.1 of the Regulatory Guide. The calculations of the Chemistry Factors for the Unit 1 and Unit 2 reactor vessels are summarized in Table 2-2 and Table 2-3, respectively.

2.2.2 P/T Limits for Heatup, Inservice Leak & Hydrostatic Testing, and Criticality The P/T limits for heatup, inservice leak & hydrostatic testing, and criticality, based on the limiting material from the Unit 1 and Unit 2 reactor vessels, are specified in Figure 2-1.

2.2.3 P/T Limits for Cooldown The P/T limits for cooldown, based on the limiting material from the Unit 1 and unit 2 reactor vessels, are specified in Figure 2-2.

5 ERX-07-003, Rev. 1

2.3 LTOP System Setpoints (LCO 3.4.12)

The nominal PORV setpoints for use with the Low Temperature Overpressure (LTOP) System are shown in Table 2-4 and Table 2-5. The PORV setpoints in Table 2-4 are applicable to Unit 1 with A76 steam generators and were extracted from Section 4.3.3 of Reference 6. The PORV setpoints in:Table 2-5 are applicable to Unit 2 with D5 steam generators and were extracted from Table 14 of Reference 5. The replacement A76 steam generator design for CPNPP Unit 1 has a larger RCS volume and a larger primary-to-secondary heat transfer area than the original D4 steam generator designs.

2.4 Reactor Vessel Material Surveillance Program A withdrawal schedule for Unit 1 is not necessary, because all Unit 1 surveillance capsules have been withdrawn from the reactor vessel. The reactor vessel material surveillance capsule withdrawal schedule for Unit 2 is provided in Table 2-6.

6

-ERX-07-003, Rev. 1

3.0 REFERENCES

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

2. "Analysis of Capsule Y from the TU Electric Company Comanche Peak Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-1 5144-NP, Revision 0, January, 1999.

3. "Analysis of Capsule X from the TU Energy Comanche Peak Unit 2 Reactor Vessel Radiation Surveillance Program," WCAP-16277-NP, Revision 0, September, 2004.

4. "Comanche Peak Units I and 2 Heatup and Cooldown Limit Curves for Normal Operation," WCAP-16346-NP, Revision 0, October 2004.

5. TXU POWER - COMANCHE PEAK STEAM ELECTRIC STATION UNITS 1 AND 2 Revised LTOP System Setpoints - Final Report, WPT-1 6994, June 28, 2007, VL-07-001465.

6. "Comanche Peak Unit 1 Replacement Steam Generator Project NSSS Engineering Report," WCAP-16469-P, Revision 1, June, 2006.

7. 'ASTM E 185-82, "Standard Practice for Conducting Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels, E706 (IF)."

8. "Evaluation of Pressurized Thermal Shock for Comanche Peak Unit 1," WCAP-13437, docketed via TXU Electric letter logged TXX-92516, December 28, 1992.

9. "Evaluation of Pressurized Thermal Shock for Comanche Peak Unit 2,"'WCAP-14345, docketed via TXU Electric letter logged TXX-95243, dated September 19, 1995.

7 ERX-07-003,'Rev. 1

Table 2-1: Limiting Materials and Reference Temperatures for CPNPP Unit I and Unit 2 Reactor Vessels Adjusted Reference Reference Temperature -

Unit Limiting Material Pressurized Thermal Temperature (ART) Shock (RT-PTS) 1/4t 314t R-1107-1, 1 Intermediate 92 0 F 80°F 100°F Shell Plate R-3807-2, 2 Intermediate 84 0 F 69 0 F 94 0 F Shell Plate 8

ERX-07-003, Rev. 1

Table 2-2: Calculation of Chemistry Factor Values using Unit I Surveillance Capsule Test Results Material Capsule F(a) FF(b) ARTNDT(c) FF x ARTNDT FF 2 Lower Shell U 0.318 0.685 6.6 4.521 0.469 R1 108-2 (Longitudinal) Y 1.49 1.11 6.9 7.66 1.23 Lower Shell U 0.318 0.685 21.3 14.591 0.469 R1 108-2 (Transverse) Y 1.49 1.11 25.3 28.08 1.23 SUM 54.852 3.398 CFR108-2 = Z( FF x ARTNDT) +i( FF 2 ) = 54.852 - 3.398 = 16.1°F Weld Metal U 0.318 0.685 010(d'e) 0.0 0.469 (Heat# 88112) Y 1.49 1.11 17.6j(d) 19.54 1.23 SUM 19.54 1.699 CFWELD = X( FF x ARTNDT) - *,( FF 2) = 19.54 - 1.699 = 11.5 0 F Notes:

(a) F = Calculated Fluence (1019 n/cm 2, E > 1.0 MeV). See Table 2-2 of Reference 4.

(b) FF = Fluence Factor F(0. 28 - 0.1 *log F)

(c) All available data is from Comanche Peak Unit 1[2]. Therefore, no temperature adjustment is required.

(d) The measured ARTNDT values for the weld metal have been adjusted by a ratio of 1.04.

(e) The CVGRAPH calculated value is -14.14 0 F. 0.0°F was used in the calculation for conservatism.

NOTE: The Chemistry Factor from the previous analysis in Reference 2 was 15.7 0 F for the surveillance lower shell plate and 10.7 0 F for the surveillance weld. As can be seen above, there is only a minor change (i.e., <1 OF) to the Chemistry Factor values. Thus, the credibility evaluation from the previous analysis remains valid. All Unit 1 surveillance data is credible.

NOTE: The value of FF for CPNPP Unit 1 has been corrected to 0.685. The value reported in WCAP-16346-NP was incorrectly stated as 0.683.

9 ERX-07-003, Rev. 1

Table 2-3: Calculation of Chemistry Factor Values using Unit 2 Surveillance Capsule Test Results Material Capsule F(a) FF(b) ARTNDT(c) FF x ARTNDT FF 2 Inter. Shell R3807-2 U 0.315 0.683 1.6 1.093 0.466 (Longitudinal) , X 2.20 1.21 1.6 1.94 1.46 Inter. Shell R3807-2 U 0.315 0.683 23.4 15.982 0.466 (Transverse) X 2.20 1.21 52.9 64.01 1.46 SUM 83.025 3.852 CFRllO8-2 = ( FF x ARTNDT) + YZ(FF 2 ) = 83.025 3.852 = 21.6°F Weld Metal U 0.315 0.683 3.74(d) 2.55 0.466 (Heat # 89833) X 2.20 1.21 5 0 .1 3 (d) 60.66 1.46 SUM 63.21 1.926 CFWELD = ,,( FF x ARTNDT) ÷ 1( FF 2 ) = 63.21 + 1.926 = 32.8°F Notes:

(a) F = Calculated Fluence. Units are x 1019 n/cm 2 (E > 1.0 MeV). See Table 2-2 of Reference 4.

(b) FF.= Fluence Factor = F(028 " 0.log F).

(c) All available data is from Comanche Peak Unit 2131. Therefore, no temperature adjustment is required.

(d) The measured ARTNDT values for the weld metal have been adjusted by a ratio of 1.04.

NOTE: For Unit 2, the surveillance plate data (for the intermediate shell plate R3807-1) is not credible, while the surveillance weld data is credible.

10 ERX-07-003, Rev. 1

Table 2-4: PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit I with Delta-76 Steam Generators - Applicable Up To 36 EFPY Adjusted RCS PORV #1 Setpoint PORV #2 Setpoint Temperature (psig) (psig)

(°F)ý 70 389 389 150 389 389 200 447 447 220 447 447 250 573 573 380 573 573 470 2335 2335 Table 2-5: PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 2 with D5 Steam Generators - Applicable Up To 36 EFPY Adjusted RCS PORV #1 Setpoint PORV #2 Setpoint Temperature (psig) (psig)

( 0°F) 70 375 375 150 375 375 200 440 440 220 440 440 250 580 580 350 580 580 405 2335 2335 11 ERX-07-003, Rev. 1

Table 2-6: Unit 2 Reactor Vessel Material Surveillance Program - Withdrawal Schedule CAPSULE VESSEL LEAD WITHDRAWAL WITH DRAWAL NUMBER LOCATION FACTOR TIME OUTAGE U 58.50 3.93 1 st Refueling 1st Refueling x 238.50 4.15 8.83 EFPY 2RF07 W 121.50 4.11 13 EFPY 2RF11 z 301.50 4.11 Standby 2RF1l V 61.00 3.87 Standby 2RF07 Y 241.00 3.87 Standby 2RF07 12 ERX-07-003, Rev. 1

Figure 2-1 Reactor Coolant System Heatup Limitations for CPNPP Unit 1 and Unit 2 -

Applicable for the First 36 EFPY (w/o Margins for Instrumentation Errors) 2500 2250 2000 1750 1500 I 1250 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 13 ERX-07-003, Rev. 1

Figure 2-2 Reactor Coolant System Cooldown Limitations for CPNPP Unit I and Unit 2 -

Applicable for the First 36 EFPY (w/o Margins for Instrumentation Errors) 2500 lOperlim Version:5.2 Run:16045 Operlim.xls Version: 5.2 2250 Unacceptable Operation 2000 LAcceptable 1750 Operation 1500 1 250 -_ Cooldown n_____ _ _____

Rates, FIH r steady-state

-20 1000

-60 i-100 750 500 Bo Ituap 250 * -----------

j - Temperature, 60 F .- _ _ __

0 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 14 ERX-07-003, Rev. 1