CP-202100127, Transmittal of Pressure and Temperature Limits Report, ERX-07-003, Revision 6

Transmittal of Pressure and Temperature Limits Report, ERX-07-003, Revision 6

ML21075A112
Person / Time
Site:	Comanche Peak
Issue date:	03/16/2021
From:	Hicks J Luminant
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
CP-202100127, T 254.897.6725, TXX-21059
Download: ML21075A112 (21)

Text

m Jack C. Hicks Manager, Regulatory Affairs Comanche Peak Nuclear Power Plant (Vistra Operations Company LLC)

Luminant P.O. Box 1002 6322 North FM 56 Glen Rose, TX 76043 T 254.897.6725 CP-202100127 TXX-21059 March 16, 2021 U. S. Nuclear Regulatory Commission Ref 10 CFR 50.36(c)(5)

ATTN: Document Control Desk Washington, DC 20555-0001

Subject:

Comanche Peak Nuclear Power Plant (CPNPP)

Docket Nos. 50-445 and 50-446 PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR), ERX-07-003, REVISION 6

Dear Sir or Madam:

Enclosed is Comanche Peak Nuclear Power Plant (CPNPP), Pressure and Temperature Limits Report, ERX-07-003, Revision 6. This report is prepared and submitted pursuant to Technical Specification 5.6.6, Reactor Coolant System (RCS) PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR).

This communication contains no new commitments regarding CPNPP Units 1 and 2.

Should you have any questions regarding this submittal, please contact Garry Struble at (254) 897-6628 or Garry.Struble@luminant.com.

Sincerely,

Enclosure:

CPNPP Pressure and Temperature Limits Report, ERX-07-003, Revision 6 c- Scott Morris, Region IV [Scott.Morris@nrc.gov]

Dennis Galvin, NRR [Dennis.Galvin@nrc.gov]

John Ellegood, Senior Resident Inspector, CPNPP Uohn.Ellegood@nrc.gov]

Neil Day, Resident Inspector, CPNPP [Neil.Day@nrc.gov]

ERX-07-003, Revision 6 COMANCHE PEAK NUCLEAR POWER PLANT (CPNPP)

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

March 2021 Prepared . Date: 3(?-/20:2 I I I Parvez Salim Principal Engineer, Westinghouse Electric Co.

Reviewed: ~ ,)---~

°anCGutflrte Date: 3- Z - 2.o 2- I Principal Engineer, Westinghouse Electric Co.

Approved: Date: 3/2/2021 Kevin N. Roland Manager, Integrated Site Engineering, Texas/Kansas ERX-07-003, Rev. 6

DISCLAIMER The information contained in this report was prepared for the specific requirement of Vistra Operations Company LLC and may not be appropriate for use in situations other than those for which it was specifically prepared. Vistra Operations Company LLC 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, Vistra Operations Company LLC does not authorize its use by others, and any such use is forbidden except with the prior written approval of Vistra Operations Company LLC. Any such written approval shall itself be deemed to incorporate the disclaimers of liability and disclaimers of warranties provided herein. In no event shall Vistra Operations Company LLC 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 ERX-07-003, Rev. 6

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 PIT Limits for Heatup, Cooldown, lnservice 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 ERX-07-003, Rev. 6

LIST OF TABLES TABLE PAGE 2-1 Limiting Materials and Reference Temperatures for CPNPP Unit 1 and Unit 2 Reactor Vessels ... .......... .. ..... ...... .... .... ............... ........... ...... ..... ...... .. ........ ................. . 9 2-2 Calculation of Chemistry Factor Values for Unit 1 Surveillance Capsule Test Results...................................................................................................................... 10 2-3 Calculation of Chemistry Factor Values for Unit 2 Surveillance Capsule Test Results...................................................................................................................... 11 2-4 PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 1 with Delta-76 Steam Generators - Applicable Up To 36 EFPY ........................................ 12 2-5 PORV Setpoints for Low Temperature Overpressure (LTOP) System For Unit 2 with 05 Steam Generators -Applicable Up To 36 EFPY ................................................. 12 iv ERX-07-003, Rev. 6

LIST OF FIGURES FIGURE PAGE 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) ...................................................................... 13 2-2 Reactor Coolant System Cooldown Limitations for CPNPP Unit 1 and Unit 2 - Applicable for the First 36 EFPY (w/o Margins for Instrumentation Errors)................................................................... 14 V

ERX-07-003, Rev. 6

1.0 INTRODUCTION

This report presents the Reactor Coolant System (RCS) Pressure and Temperature (PIT) limits for Comanche Peak Nuclear Power Plant (CPNPP) Unit 1 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 (PIT) 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 O of this PTLR was submitted to the NRG in support of Operating License Amendment 132. The NRG reviewed the submittal and 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 were changed to those of Table 14 of Reference 5.

In Revision 2 of this PTLR, the LTOP System PORV setpoints for CPNPP Unit 1 with the /176 steam generators were changed to those of Table 12 of Reference 5.

1 ERX-07-003, Rev. 6

In Revision 3 of this PTLR, the heatup and cooldown PIT limit curves (Figures 2-1 and 2-2) for CPNPP Units 1 and 2 were changed to those of Reference 6.

In Revision 4 of this PTLR, the LTOP System PORV setpoints for CPNPP Unit 1 with t,,,. 76 steam generators were changed to those of Table 9 of Reference 10.

In Revision 5 of this PTLR, the LTOP System PORV setpoints for CPNPP Unit 2 with D5 steam generators are changed to those of Table 9 of Reference 11.

In Revision 6 of this PTLR, the latest capsule results from Unit 1 Capsule X in WCAP-16610-NP

[2] and Unit 2 Capsule Win WCAP-17269-NP [3] are reflected.

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 PIT 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 PIT 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 PIT limits that bound both CPNPP units.

The RCS PIT limits calculated for selected heatup and cooldown rates for CPNPP Unit 1 and Unit 2 are extracted from Reference 6.

2 ERX-07-003, Rev. 6

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 /J.76 steam generators) and those for CPNPP Unit 2 (with the D5 steam generators) are extracted from References 10 and 5, respectively.

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 (RT Nor) 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 Intermediate Shell Plate R-1107-1 plate material in the Unit 1 beltline was most limiting. Forecast properties of the limiting material were used to establish PIT limits for heatup and cooldown curves and LTOP setpoints.

The final reactor vessel specimen capsules were withdrawn and tested from Units 1 and 2 when the neutron fluence exceeded 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.

3 ERX-07-003, Rev. 6

For Unit 1, the required specimen capsules U, Y, and X 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 [12].

Two of the standby capsules (Capsules V and W) were withdrawn in 1RF09 and stored for later evaluation, if necessary. The third standby capsule (Capsule Z) was withdrawn during 1RF11 in the fall of 2005 and stored for later evaluation, if necessary.

For Unit 2, the required specimen capsules U, X, and W have been withdrawn and evaluated

[3]. The third required specimen capsule, Capsule W, was withdrawn during 2RF11 in the fall of 2009, with a fluence within the range of one-times to two-times the 54 EFPY Peak Fluence [3].

Two of the standby capsules (Capsules V and Y) were withdrawn in 2RF07 and stored for later evaluation, if necessary. The third standby capsule (Capsule Z) was withdrawn during 2RF11 and stored for later evaluation, if necessary.

Because all reactor vessel surveillance capsules have been withdrawn, and the requirements ASTM E185-82 have been met for Units 1 and 2 for the 40-year licensed period, capsule removal schedules are not required.

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 lnservice Leak and Hydrostatic Testing 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, lnservice 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 4

ERX-07-003, Rev. 6

presented in Table 2-1. These values were 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. With consideration of three surveillance capsules, these values remain conservative. All surveillance data for Unit 1 is credible. For Unit 2, the surveillance plate data (for the intermediate shell plate R3807-2) is not credible, while the surveillance weld data is credible.

The limiting reference temperatures for pressurized thermal shock (RT PTs) 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 RT PTs screening criteria, the conservative fluence projections for the CPNPP vessel materials, and the absence of a significant change in the projected values of RT PTs, the results of the Pressurized Thermal Shock reports in WCAP-13437 [8] for Unit 1 and WCAP-14345 [9] for Unit 2 have not been revised in this PTLR.

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 three 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.

5 ERX-07-003, Rev. 6

2.2.2 PIT Limits for Heatup. lnservice Leak & Hydrostatic Testing. and Criticality The PIT limits for heatup, inservice leak & hydrostatic testing, and criticality, based on the limiting material from the Unit 1 and Unit 2 reactor vessels, are extracted from Reference 6 and presented in Figure 2-1.

2.2.3 PIT Limits for Cooldown The PIT limits for cooldown, base_d on the limiting material from the Unit 1 and Unit 2 reactor vessels, are extracted from Reference 6 and presented in Figure 2-2.

2.3 LTOP System Setpoints (LCO 3.4.12)

The nominal PORV setpoints for use with the Low Temperature Overpressure (L TOP) System are shown in Table 2-4 and Table 2-5. The PORV setpoints in Table 2-4 are applicable to Unit 1 with 1176 steam generators and were extracted from Table 9 of Reference 10. The PORV setpoints in Table 2-5 are applicable to Unit 2 with 05 steam generators and were extracted from Table 9 of Reference 11.

2.4 Reactor Vessel Material Surveillance Program A withdrawal schedule for Units 1 and 2 are not necessary, because all Units 1 and 2 surveillance capsules have been withdrawn from the reactor vessel.

6 ERX-07-003, Rev. 6

3.0 REFERENCES

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

2. "Analysis of Capsule X from the TXU Electric Comanche Peak Unit 1 Reactor Vessel Radiation Surveillance Program," WCAP-16610-NP, Revision 0, September, 2006.

3. "Analysis of Capsule W from the Comanche Peak Unit No. 2 Reactor Vessel Radiation Surveillance Program," WCAP-17269-NP, Revision 0, September, 2010.

4. "Comanche Peak Units 1 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-16994, June 28, 2007, VL-07-001465.

6. "Luminant Comanche Peak Nuclear Power Plant Unit 1 and 2 Reactor Vessel Pressure-Temperature Limits," WPT-17774, March 13, 2014, VDRT-4804676.

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.

10. "Comanche Peak Unit 1 LTOPS PORV Setpoint Revision due to MSIP and Legacy Errors," LTR-SCS-19-7, Revision 0, dated March 28, 2019, VDRT-5732489.

7 ERX-07-003, Rev. 6

11. "Comanche Peak Unit 2 LTOPS PORV Setpoint Revision," LTR-SCS-19-12, Revision 0, dated May 1, 2019, Transmitted via WPT-18170 VDRT-5749607.

12. "Comanche Peak Nuclear Power Plant Stretch Power Uprate Licensing Report,"

WCAP-16840-NP, Revision 0, August 2007.

8 ERX-07-003, Rev. 6

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

Limiting Material Adjusted Reference Unit Pressurized Thermal for P-T Curves Temperature (ART)

Shock (RT-PTS) 1/4t 3/4t R-1107-1, 1 Intermediate 92°F 80°F 100°F Shell Plate R-3807-2, 2 Intermediate 84°F 69°F 94°F Shell Plate NOTE: For Unit 1, the limiting RT Prs from Reference 8 is for the Lower Shell Plate R-1108-1; while Intermediate Shell Plate R-1107-1 was identified as the limiting material for the 1/4t and 3/4t locations for the P-T curves in Reference 4. For Unit 2, Intermediate Shell Plate R-3807-2 is limiting for both P-T limit curves and RT PTs as shown in References 4 and 9.

9 ERX-07-003, Rev. 6

Table 2-2: Calculation of Chemistry Factor Values using Unit 1 Surveillance Capsule Test Results Material Capsule F(a) FF(b) ~RTNDT(c) FF X ~RT NOT FF 2 Lower Shell u 0.318 0.685 6.7 4.592 0.470 R1108-2 y 1.49 1.11 7.0 7.773 1.23 (Longitudina0 X 3.24 1.31 48.6 63.610 1.71 Lower Shell u 0.318 0.685 21.5 14.735 0.470 R1108-2 y 1.49 1.11 25.7 28.538 1.23 (Transverse)

X 3.24 1.31 35.6 46.595 1.71 SUM 165.8 6.832 CFR11oa-2 =L( FF x ~RTNoT) + L( FF2) =165.8 + 6.832 =24.3°F Weld Metal u 0.318 0.685 O.O(d,e) 0.0 0.470 (Heat# 88112) y 1.49 1.11 16.5(d) 18.322 1.23 X 3.24 1.31 27.0(d) 35.339 1.71 SUM 53.66 3.416 CFwEL = I( FF x ~RT NOT) + L( FF 2) =53.66 + 3.416 =15. 7°F Notes:

(a) F = Calculated Fluence (10 19 n/cm 2 , E > 1.0 MeV).

( b) FF = Fluence Factor= F< 0 0.1

• log F)

( c) All available data is from Comanche Peak Unit 1l2l. Therefore, no temperature adjustment is required.

( d) The measured fiRTNOT values for the weld metal have been adjusted by a ratio of 0.955 (CF vessel weld

+ CF surveillance weld).

( e) The measured fiRT NOT value of -13.4°F was observed. 0.0°F was used in the calculation for conservatism.

NOTE: WCAP-16610-NP identifies the Unit 1 plate surveillance material as non-credible because one of the measured data points has a scatter of greater than one standard deviation (CT), 17°F, from the Regulatory Guide 1.99, Revision 2, Positions 2.1 fitted chemistry factor. However, from a statistical point of view, +/- 1CT would be expected to encompass 68% of the data. The total number of data points that meet the+/- 1CT criterion (5 out of 6 or 83%) is well within these bounds. Therefore, the Comanche Peak Unit 1 surveillance plate data is treated as credible in this PTLR.

10 ERX-07-003, Rev. 6

Table 2-3: Calculation of Chemistry Factor Values using Unit 2 Surveillance Capsule Test Results Material Capsule F(a) FF(b) LiRTNDT(c) FF X LiRTNDT FF 2 Inter. Shell R3807-2 u 0.317 0.685 1.6 1.10 0.469 (Longitudina~ X 2.16 1.209 1.6 1.93 1.462 w 3.38 1.319 23.2 30.59 1.739 Inter. Shell R3807-2 u 0.317 0.685 23.4 16.02 0.469 (Transverse) X 2.16 1.209 52.9 63.96 1.462 w 3.38 1.319 74.4 98.11 1.739 SUM 211.71 7.339 CFR3807-2 = L( FF X LiRTNDT) + L( FF 2) = 211.71 + 7.339 = 28.8°F Weld Metal u 0.317 0.685 3.7(d) 2.56 0.469 (Heat# 89833) X 2.16 1.209 50.1 (d) 60.61 1.462 w 3.38 1.319 87_4(d) 115.20 1.739 SUM 178.37 3.669 CFwELD = L( FF x LiRTNor) + I( FF 2) = 178.37 + 3.669 = 48.6°F Notes:

(a) F = Calculated Fluence. Units are x 1019 n/cm2 (E > 1.0 MeV).

(b) FF= Fluence Factor= po.zs-o.1*1ogF).

(c) All available data is from Comanche Peak Unit 21 31. Therefore, no temperature adjustment is required.

(d) The measured LiRTNDT 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.

11 ERX-07-003, Rev. 6

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

(OF) 60 374 374 180 374 374 185 440 440 230 440 440 240 568 568 350 568 568 405 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)

(°F) 60 374 374 180 374 374 185 432 432 230 432 432 240 577 577 350 577 577 405 2335 2335 12 ERX-07-003, Rev. 6

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 ILeak Test Limit I

,' f4- ----- !Critical Limit 20Deg. F/Hr 1 J j~ r---_

2000 !Critical Limit

__,,-/' 7

(!)

u, 1750 -

--teatup Rate 20 Deg. F/Hr

/

, , l~~

f'4 J 60 Deg. F/Hr

!Critical Limit 100 Deg. F/Hr

--1Heatup Rate fl. 1500 - 60 Deg. F/Hr I

i,,

~

I

/

ti) 1250 7Heatup Rate ti) 100 Deg. F/Hr

~

Acceptable fl.

1000 '"'" Unacceptable Operation "C

s

.!!! Operation

I 750

£ ca (J

500 Boltup Criticality Limit based on Temp. inservice hydrostatic test 250 60°F V

--- temperature (152°F) for the service period up to 36 EFPY 0

i I

~

r--- ~The lower limit for RCS pressure is-14.7 psig I I I I I I I I I I I I I I I I I I 0 50 100 150 200 250 300 350 400 450 500 550 Moderator Temperature (Deg. F) 13 ERX-07-003, Rev. 6

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

Applicable for the First 36 EFPY (w/o Margins for Instrumentation Errors) 2500 -**-

2250 I

Unacceptable 2000 *'- Operation C)

U) 1750 Acceptable Operation

a. 1500 Q)

, ~7 u,

u, 1250 Cooldown Rates, °F/Hr Q) a.

"O 1000 --- steady-state

-20

.e

-40

-60

, 750 ~ ..____,,"

-100

£ cu

(.) -

500 Boltup 250 ~--*- Temperature, 60°F --

V 0

~ ~The lower limit for RCS pressure is -14.7 psig I I I I I

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

14 ERX-07-003, Rev. 6

EFFECTIVE PAGE LIST (EPL)

Unit 1 & Unit 2, PTLR Page No. Revision Title Page ERX-07-003, Rev. 6 ii ERX-07-003, Rev. 6 iii ERX-07-003, Rev. 6 iv ERX-07-003, Rev. 6 V ERX-07-003, Rev. 6 1 ERX-07-003, Rev. 6 2 ERX-07-003, Rev. 6 3 ERX-07-003, Rev. 6 4 ERX-07-003, Rev. 6 5 ERX-07-003, Rev. 6 6 ERX-07-003, Rev. 6 7 ERX-07-003, Rev. 6 8 ERX-07-003, Rev. 6 9 ERX-07-003, Rev. 6 10 ERX-07-003, Rev. 6 11 ERX-07-003, Rev. 6 12 ERX-07-003, Rev. 6 13 ERX-07-003, Rev. 6 14 ERX-07-003, Rev.6