CPSES-200601971, Submittal of Additional Information Related to License Amendment Request 05-007, Revision to Technical Specification 5.6.6 Regarding Pressure and Temperature Limits Report

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Submittal of Additional Information Related to License Amendment Request 05-007, Revision to Technical Specification 5.6.6 Regarding Pressure and Temperature Limits Report
ML062790150
Person / Time
Site: Comanche Peak  Luminant icon.png
Issue date: 09/29/2006
From: Madden F
TXU Power
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
00236, CPSES-200601971, TAC MC9500, TAC MC9501, TXX-06168
Download: ML062790150 (21)
v • d • e


Text

6,LJ TXU Powe r TXU Power Mike Blevins Ref: 10CFR50.90 Comanche Peak Steam Senior Vice President &

Electric Station Chief Nuclear Officer P. O. Box 1002 (E01)

Glen Rose. IX 76043 Tel: 254 897 5209 Fax: 254 897 6652 mike.blevins@txu.com CPSES-200601971 Log # TXX-06168 File # 00236 September 29, 2006 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 SUBMITTAL OF ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST 05-007, REVISION TO TECHNICAL SPECIFICATION 5.6.6 REGARDING PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

TAC NOS. MC9500 AND MC9501 REF: 1) TXU Power letter, logged TXX-05198, from Mike Blevins to the U. S. Nuclear Regulatory Commission, dated December 16, 2005.

2) TXU Power letter, logged TXX-06146, from Mike Blevins to the U. S. Nuclear Regulatory Commission, dated August 31, 2006.

Dear Sir or Madam:

In Reference 1, TXU Generating Company LP (TXU Power) submitted a proposed amendment which would revise the CPSES Technical Specifications (TS) requirements consistent with the NRC-approved Revision 0 to Technical Specification Task Force (TSTF) Standard Technical Specification Change Traveler, TSTF-419, "Revise PTLR Definition and References in ISTS 5.6.6, RCS PTLR."

Based on comments identified by the NRC reviewer during a conference call on September 21, 2006, TXU Power hereby submits, as an enclosure to this letter, a revised sample PTLR. This sample PTLR replaces those previously submitted in References 1 and 2, and has been revised to specifically address the comments identified by the reviewer.

A member of the STARS (Strategic Teaming and Resource Sharing) Alliance

-AOO(

Callaway

  • Comanche Peak o Diablo Canyon 9 Palo Verde
  • South Texas Project 9 Wolf Creek

TXX-06168 Page 2 of 2 This communication contains no new or revised commitments.

Should you have any questions, please contact Mr. Bob Kidwell at (254) 897-5310.

In accordance with I OCFR50.9 I(b), TXU Power is providing the State of Texas with a copy of this proposed amendment.

I state under penalty of perjury that the foregoing is true and correct.

Executed on September 29, 2006.

Sincerely, -

TXU Generation Company LP By: TXU Generation Management Company LLC Its General Partner Mike Blevins By: ki 2(7I'd

/Fred W. Madden Director, Oversight and Regulatory Affairs RJK Enclosure Sample Pressure and Temperature Limits Report (PTLR) c - B. S. Mallett, Region IV (w/o Encl)

M. C. Thadani, NRR Resident Inspectors, CPSES (w/o Encl)

Ms. Alice Rogers Bureau of Radiation Control Texas Department of Public Health 1100 West 49th Street Austin, Texas 78756-318

ERX-06-00x, Draft Rev. C LIP COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)

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

Month 200x Prepared: Date:

Engineer Safety Analysis Approved: Date:

Manager Safety Analysis Manager Approved: Date:

Manager Technical Programs Manager

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE DISCLAIMER The information contained in this report was prepared for the specific requirement of TXU Power, and may not be appropriate for use in situations other than those for which it was specifically prepared. TXU Power 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, TXU Power does not authorize its use by others, and any such use is forbidden except with the prior written approval of TXU Power. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall TXU Power 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.

ii

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE TABLE OF CONTENTS DISCLAIMER ............................................................................................................................ ii TABLE OF CONTENTS ............................................................................................................. iii LIST OF TABLES ..................................................................................................................... iv LIST OF FIGURES ................................................................................................................... v SECTION PAGE

1.0 INTRODUCTION

......................................................................................................... 1 2.0 OPERATING LIMITS ..................................................................................................... 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 Criticality (LCO 3.4.3) ....................................................................................... 4 2.3 LTOP System Setpoints (LCO 3.4.12) ............................................................. 6 2.4 Reactor Vessel Material Surveillance Program ................................................ 6

3.0 REFERENCES

.......................................................................................................... 7 iii

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE LIST OF TABLES TABLE PAGE 2-1 Limiting Materials and Reference Temperatures for CPSES Unit 1 and Unit 2 R eactor Vessels ......................................................................................................... 8 2-2 Calculation of Chemistry Factor Values for Unit 1 Surveillance Capsule Test Results ........................................................................................................................ 9 2-3 Calculation of Chemistry Factor Values for Unit 2 Surveillance Capsule Test Results ...................................................................................................................... 10 2-4 PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 1 Replacement Steam Generators - Applicable Up To 36 EFPY ................................ 11 2-5 PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 1 and Unit 2 with Original Steam Generators - Applicable Up To 36 EFPY ................ 11 2-6 Unit 2 Reactor Vessel Material Surveillance Program - Withdrawal Schedule ........ 12 iv

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE 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

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE

1.0 INTRODUCTION

This report presents the Reactor Coolant System (RCS) Pressure and Temperature (P/T) limits for Comanche Peak Steam Electric Station (CPSES) 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 CPSES Unit 1 and Unit 2 operation for 36 Effective Full Power Years (EFPY).

I

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE 2.0 OPERATING LIMITS RCS P/T Limits The RCS P/T 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 CPSES 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 CPSES units.

The RCS P/T limits calculated for selected heatup and cooldown rates for CPSES Unit I 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.

2

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE The LTOP System PORV setpoints for CPSES Unit I (with the replacement A76 steam generators) are extracted from Reference 5. The LTOP System PORV setpoints developed for CPSES Unit I (with the original D4 steam generators) and Unit 2 are extracted from Reference 6.

REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reduction in toughness that results from neutron radiation is measured as an increase in the Nil Ductility Reference Temperature (RTNDT) and reduction of the upper-shelf energy of reactor vessel beltline materials, including welds. At CPSES, 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 I 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 1 RF1 1 in the fall of 2005, with a fluence within the range of one-times to two-times the 52 EFPY Peak Fluence [21, 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 3

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE 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 2RFl 1 in the spring of 2010 and stored for later evaluation, if necessary.

2.1 RCS Temperature Rate-of-Change Limits (LCO 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 100°F in any 1-hour period.

2.1.3 Maximum Temperature Change During Inservice Leak and Hydrostatic Testinq During inservice leak and hydrostatic testing operations above the heatup and cooldown limit curves, the RCS temperature change limit is 10°F 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 plate data (for the intermediate shell plate R3807-1) is not credible, while the surveillance weld data is credible.

4

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE The limiting reference temperatures for pressurized thermal shock (RTPTS) values for each unit's reactor vessel were previously docketed in accordance with IOCFR50.61 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 CPSES 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 CPSES vessel materials, and the 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 I 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.

5

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

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

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 the replacement (Model A76) steam generators and were extracted from Section 4.3.3 of Reference 5. The PORV setpoints in Table 2-5 are applicable to Unit 1 with original (Model D4) steam generators and for Unit 2 with original (Model D5) steam generators and were extracted from Reference 6. The replacement steam generator design has a larger RCS volume and a larger primary-to-secondary heat transfer area than the original 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

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE

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 I Reactor Vessel Radiation Surveillance Program," WCAP-15144-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 1 and 2 Heatup and Cooldown Limit Curves for Normal Operation," WCAP-1 6346-NP, Revision 0, October 2004.
5. "Comanche Peak Unit 1 Replacement Steam Generator Project NSSS Engineering Report," WCAP-16469-P, Revision 1, June, 2006.
6. Revised COMS Setpoints for Unit 1 and Unit 2 OSG, WPT-16748.
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-1 3437, 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

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

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

Unit Limiting Material Temperature (ART) Pressurized Sok(TPS Thermal Shock (RT-PTS) 114t 314t R-1107-1, 1 Intermediate 92 0F 80°F 100I F Shell Plate R-3807-2, 2 Intermediate 84 0 F 690 F 94 0 F Shell Plate 8

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE 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 0.683 6.6 4.521 0.469 Ri 108-2 (Longitudinal) Y 1.49 1.11 6.9 7.66 1.23 Lower Shell U 0.318 0.683 21.3 14.591 0.469 RI 108-2 (Transverse) Y 1.49 1.11 25.3 28.08 1.23 SUM 54.852 3.398 CFR1108-2 = X( FF x ARTNDT) - *( FF 2 ) = 54.852 - 3.398 = 16.1°F Weld Metal U 0.318 0.683 0 .0 (de) 0.0 0.469 (Heat # 88112) Y 1.49 1.11 1 7 .6(d) 19.54 1.23 SUM 19.54 1.699 CFWELD = ,( FF x ARTNDT) + Z( FF 2 ) = 19.54 + 1.699 = 11.5°F Notes:

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

0 28 0 1 (b) FF = Fluence Factor = F( - . logF)

(c) All available data is from Comanche Peak Unit 1r2!. 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 0F. 0.0°F was used in the calculation for conservatism.

NOTE: The Chemistry Factor from the previous analysis in Reference 2 was 15.7 0F for the surveillance lower shell plate and 10.7°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.

9

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE 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 CFRI108-2 = X( FF x ARTNDT) + Z( FF 2 ) = 83.025 - 3.852 = 21.6 0 F Weld Metal U 0.315 0.683 3. 7 4 (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 = i( FF x ARTNDT) + .( FF 2 ) = 63.21 + 1.926 = 32.8 0 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 = fO.28-0.1,logf).

(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

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

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

(OF) 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 I Original Steam Generators and Unit 2 with Original Steam Generators - Applicable Up To 36 EFPY Adjusted RCS Temperature PORV #1 Setpoint (psig) PORV #2 Setpoint (psig)

(OF) 70 375 375 150 375 375 200 440 440 220 440 440 250 580 580 380 580 580 470 2335 2335 II

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

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

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

SAMPLE SAMPLE SAMPLE Figure 2-1 Reactor Coolant System Heatup Limitations for CPSES Unit I and Unit 2 -

Applicable for the First 36 EFPY (w/o Margins for Instrumentation Errors) 2500 Leak Test Limit J - loperlim Version:5.2 Run:16045 Operlim.xts Version: 5.:

2250 . .... Critical C Limit 20 Deg. F/Hr.I 2000 Critical Limit 1750 ...... .- 60 Deg. F/Hr Heatup RateI 20 Deg. CF/HrILii 1750 I . .10 Deg. F/HrI Heatup Rate C" - 60 Deg. F/Hr 1500

- Heatup Rate A 100 Deg. FIHr Acceptable 1250

[ Opera t ion 01000---

0 Unacceptable a Operation 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 13

Pressure and Temperature Limits Report for Comanche Peak Steam Electric Station (Applicable Up To 36 EFPY)

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

Applicable for the First 36 EFPY (w/o Margins for Instrumentation Errors) 2500 2250 2000 1750 0 1500

  • 1250 IL U 1000 750 500 250 0

0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 14

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