ML051930019

R. E. Ginna Nuclear Power Plant, Transmittal of RCS Pressure and Temperature Limits Report (PTLR)

ML051930019
Person / Time
Site:	Ginna
Issue date:	07/01/2005
From:	Korsnick M Constellation Energy Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML051930019 (16)

Text

Maria Korsnick 1503 Lake Road Vice President Ontario, New York 14519-9364 585.771.3494 585.771.3943 Fax maria.korsnick~constellation.com Constellation Energy I t R.E. Ginna Nuclear Power Plant July 1, 2005 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 Transmittal of RCS Pressure and Temperature Limits Report (PTLR)

In accordance with the R.E. Ginna Nuclear Power Plant Improved Technical Specification 5.6.6, which requires the submittal of revisions to the PTLR, the attached report is hereby submitted.

There are no new commitments being made in this submittal. Should you have questions regarding the information in this submittal, please contact George Wrobel at (585) 771-3535 or george.wrobeleconstellation.com.

Attachment cc: S. J. Collins, NRC P. D. Milano, NRC Resident Inspector, NRC

/06I5q7 61-

PTLR Constellation Energye R.E. Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report PTLR Revision 4 Responsible Manager A rGeorge Wro Effective Date: -

6/29/05 Controlled Copy No.

Record Cat.# 4.43.3 (

R.E. Ginna Nuclear Power Plant PTLR-1 Revision 4

PTLR 1.0 RCS Pressure and Temperature Limits Report (PTLR)

This Pressure and Temperature Limits Report (PTLR) for the R.E. Ginna Nuclear Power Plant has been prepared in accordance with the requirements of Technical Specification 5.6.6.

Revisions to the PTLR shall be provided to the NRC after issuance.

The Technical Specifications addressed in this report are listed below:

3.4.3 RCS Pressure and Temperature (PIT) Limits 3.4.6 RCS Loops - MODE 4 3.4.7 RCS Loops - MODE 5, Loops Filled 3.4.10 Pressurizer Safety Valves 3.4.12 Low Temperature Overpressure Protection (LTOP) System R.E. Ginna Nuclear Power Plant PTLR-2 Revision 4

PTLR 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. All changes to these limits must be developed using the NRC approved methodologies specified in Technical Specification 5.6.6. These limits have been determined such that all applicable limits of the safety analysis are met. All items that appear in capitalized type are defined in Technical Specification 1.1, Definitions.

2.1 RCS Pressure and Temperature Limits' (LCO 3.4.3)

(LCO 3.4.12) 2.1.1 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 600F per hour.

b. A maximum cooldown of 100F per hour.

2.1.2 The RCS P/T limits for heatup and cooldown are specified by Figure PTLR - 1 and Figure PTLR - 2, respectively.

2.1.3 The minimum boltup temperature, using the methodology of Reference 4, Enclosure 2 is 600F.

2.2 Low Temperature Overpressure Protection System Enable Temperature 2 (LCO 3.4.6)

(LCO 3.4.7)

(LCO 3.4.10)

(LCO 3.4.12) 2.2.1 The enable temperature for the Low Temperature Overpressure Protection System is 322 0F.

2.3 Low Temperature Overpressure Protection System SetpoInts (LCO 3.4.12) 2.3.1 Pressurizer Power Operated Relief Valve Lift Setting Limits3 The lift setting for the pressurizer Power Operated Relief Valves (PORVs) is

< 411 psig (includes instrument uncertainty).

R.E. Ginna Nuclear Power Plant PTLR-3 Revision 4

PTLR 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation surveillance specimens shall be removed and examined to determine changes in material properties. The removal schedule is provided in Table PTLR - 1. The results of these examinations shall be used to update Figure PTLR - 1 and Figure PTLR - 2.

The pressure vessel steel surveillance program (Ref. 5) is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Radiation Surveillance Program."

The material test requirements and the acceptance standard utilize the reference nil-ductility temperature, RTNDT, which is determined in accordance with ASTM E208.

The empirical relationship between RTNDT and the fracture toughness of the reactor vessel steel is developed in accordance with Appendix G."Protection Against Non-Ductile Failure," to section III of the ASME Boiler and Pressure Vessel Code. The surveillance capsule removal schedule meets the requirements of ASTM El 85-82.

As shown by Reference 1 (specifically its Reference 51), the reactor vessel material irradiation surveillance specimens indicate that the surveillance data meets the credibility discussion presented in Regulatory Guide 1.99 Revision 2 where:

1. The capsule materials represent the limiting reactor vessel material.

2. Charpy energy vs. temperature plots scatter are small enough to permit determination of 30 ft-lb temperature and upper shelf energy unambiguously.

3. The scatter of ARTNDT values are within the best fit scatter limits as shown on Table PTLR - 2. The only exception is with respect to the Intermediate Shell which is not the limiting reactor vessel material.

4. The Charpy specimen irradiation temperature matches the reactor vessel surface interface temperature within +/- 250F.

5. The surveillance data falls within the scatter band of the material database.

R.E. Ginna Nuclear Power Plant PTLR4 Revision 4

PTLR 4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES 4.1 The RTpTS value for Ginna Station limiting beltline material is 256.60F for 32 EFPY per Reference 1.

4.2 Tables Table PTLR - 1 contains the location and schedule for the removal of surveillance capsules.

Table PTLR - 2 contains a comparison of measured surveillance material 30 ft-lb transition temperature shifts and upper shelf energy decreases with Regulatory Guide 1.99, Revision 2 predictions.

Table PTLR - 3 shows calculations of the surveillance material chemistry factors using surveillance capsule data.

Table PTLR - 4 provides the reactor vessel toughness data.

Table PTLR - 5 provides a summary of the fluence values used in the generation of the heatup and cooldown limit curves.

Table PTLR - 6 shows example calculations of the ART values at 32 EFPY for the limiting reactor vessel material.

5.0 REFERENCES

1. WCAP-14684. "R.E.Ginna Heatup and Cooldown Limit Curves for Normal Operation,"

dated June 1996.

2. WCAP-14040-NP-A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 2, January 1996.

3. Letter from R.C. Mecredy, RG&E, to Guy S Vissing, NRC,

Subject:

"Application for Amendment to Facility Operating License, Revision to Reactor Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR) Administrative controls Requirements," dated September 29, 1997.

4. Letter from R.C. Mecredy, RG&E, to Guy S. Vissing, NRC, "Clarifications to Proposed Low Temperature Overpressure Protection System Technical Specification," dated June 3, 1997.

5. WCAP-7254, Rochester Gas and Electric, Robert E. Ginna Unit No. 1 Reactor Vessel Radiation Surveillance Program," May 1969.

R.E. Ginna Nuclear Power Plant PTLR-5 Revision 4

PTLR

6. Letter from R.C Mecredy, RG&E, to Guy S. Vissing, NRC, "Corrections to Proposed Low Temperature Overpressure Protection System Technical Specification," October 8, 1997.

7. RG&E Design Analysis DA-ME-97-031, "Evaluation of Ginna RCS Coolant Temperature to Support LTOPS Requirements," Revision 0.

R.E. Ginna Nuclear Power Plant PTLR-6 Revision 4

PTLR MATERIAL PROPERTY BASIS LMITING MATERIAL, RCUMFERENTIAL WELD 8A4847 UMING ART VALUES AT 32 EFPY: 1t4T. 241-F 3/4T. 207F 2500 L *1E 4B942966985435 ma 2250 7

IE t &13

=. 2000 V 1 750 LB]

I it

= 1500 n2 1250 tQ PATY UAIT acZRtATI If I 1000 I 1 is **t1 /IT.

JTn V0

.i- 750 S

hXATVP -- TI IORATE It. I

'It 500 f

250 I

msasPRal P ga 0.1ar 0 .l I ...- .

0 50 1o 6 Io 200 250 360 35O 400 4 0 S1 a Indicated Temperature ( De g . F )

Figure PTLR - 1 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates of 60 and 100°F)

I Applicable to 32 EFPY (Without Margins for Instrumentation Errors)

R.E. Ginna Nuclear Power Plant PTLR-7 Revision 4

PTLR MATERIAL PROPERTY BASIS UMITING MATERIAL CIRCUMFERENTIAL WEL SA-847 UMITING ART VALUES AT 32 EFPY: 1/4T, 241T 314T, 20TF 2 500 .j 2250

° 2000 1750 IUKACCzTi V I Owiaa:TCII L-500 0,

5-tn m1 1250 ACCIPTADLt Iclps EI HR

0. I 1000 750 13LT3 1 /Br.

0O 500S W

4 41 JI&!

250 0

U so i6o ito 260 2to 360 3to 46o 450 s5o Indicated Temperature (Deg.F)

Figure PTLR-2 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates of 0, 20 40 60 and 1OO 0 F/hr) Applicable to 32 EFPY (Without Margins for Instrumentation Errors)

I R.E. Ginna Nuclear Power Plant PTLR-8 Revision 4

PTLR Table PTLR - 1 Surveillance Capsule Removal Schedule Capsule Vessel Location (deg.) Capsule Lead Factor Removal Schedule(8) Capsule Fluence -

E19(n/cm2)(b)

V 770 2.99 1.6 (removed) .5028 R 2570 3.00 2.7 (removed) 1.105 T 670 1.85 7 (removed) 1.864 S 570 1.74 17 (removed) 3.746 N 2370 1.74 TBD() TBD(C)

P 2470 1.9 Standby N/A (a) Effective Full Power Years (EFPY).

(b) Reference 1.

(c) To be determined, there is no current requirement for removal.

R.E. Ginna Nuclear Power Plant PTLR-9 Revision 4

PTLR Table PTLR - 2 Surveillance Material 30 ft-lb Transition Temperature Shift 30 lb-ft Transition Temperature Shift Fluenice (X1019n/cM 2, Predictedt () Measured(")

Material Capsule E> 1.0 MeV)(8) (OF) (IF) (OF)

V .5028 26 25 1 R 1.105 32 25 7 T 1.864 37 30 7 Lower Shell S 3.746 42 42 0 V .5028 37 0 37 R 1.105 46 0 46 T 1.864 52 0 52 Intermediate Shell S 3.746 59 60 1 V .5028 135 140 5 R 1.105 168 165 3 T 1.864 191 150 41 Weld Metal 5 3.746 218 205 13 V .5028 - 0 _

R 1.105 90 T 1.864 100 HAZ Metal 5 3.746 95 (a) Reference 1 (including its Reference 51).

R.E. Ginna Nuclear Power Plant PTLR-1 0 Revision 4

PTLR Table PTLR - 3 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Fluence FF ARTNDT FF*ARTNDT FF2 19 2 (x 10 n/crn , (OF)(B)b) (OF)

E > 1.0 MeV)(a)

Intermediate V .5028 .8081 25 20.2 .6530 Shell Forging 05 (Tangential) R 1.105 1.0279 25 25.7 1.0566 T 1.864 1.1706 30 35.1 1.3703 S 3.746 1.3418 42 56.4 1.8004 Sum: 137.4 4.8803 Chemistry Factor = 28.20F Intermediate V .5028 .8081 0 0 .6530 Shell R 1.105 1.0279 0 0 1.0566 T 1.864 1.1706 0 0 1.3703 S 3.746 1.3418 60 80.5 1.8004 Sum: 80.5 4.8803 Chemistry Factor = 16.5 0F Weld Metal V .5028 .8081 149.7 121.0 .6530 R 1.105 1.0279 176.4 181.3 1.0566 T 1.864 1.1706 160.4 187.8 1.3703 S 3.746 1.3418 219.1 294.0 1.8004 Sum: 854.69 4.8803 Chemistry Factor = 160.70F (a) Reference 1.

(b) ARTNDT for weld material is the adjusted value using the 1.069 ratioing factor per Reference 1 applied to the measured values of Table PTLR - 2.

R.E. Ginna Nuclear Power Plant PTLR-1 1 Revision 4

PTLR Table PTLR - 4 Reactor Vessel Toughness Table (Unirradiated) (8)

Material Description Cu (%) Ni (%) Initial RTNDT (IF)

Intermediate Shell .07 .69 20 Lower Shell .05 .69 40 Circumferential Weld .25 .56 -4.8 (a) Per Reference 1.

R.E. Ginna Nuclear Power Plant PTLR-12 Revision 4

PTLR Table PTLR - 5 19 2 Reactor Vessel Surface Fluence Values at 19.5 and 32 EFPY(a) x 10 (n/cm , E > 1.0 MeV)

EFPY 00 150 30° 450 19.5 2.32 1A7 1.05 .969 32 3.49 2.20 1.56 1.45 (a) Reference 1.

R.E. Ginna Nuclear Power Plant PTLR-1 3 Revision 4

PTLR Table PTLR - 6 1 Calculation of Adjusted Reference Temperatures at 32 EFPY for the Limiting Reactor Vessel Material Parameter Values Operating Time 32 EFPY Material Circ. Weld Circ. Weld Location 1/4-T 314-T Chemistry Factor (CF), OF(a) 160.7 160.7 Fluence (f), 1019 n/cm 2 (E > 1.0 MeV)(b) 2.36 1.08 Fluence Factor (FF) 1.23 1.02 ARTNDT = CF x FF, OF 197.7 163.9 Initial RTNDT (I), OF -4.8 -4.8 Margin (M), OF(b) 48.3 48.3 ART = I + (CFxFF) + M, OF(bXc) 241 207 (a) Values from Table PTLR - 3.

(b) Value calculated using Table PTLR - 5 values.

(c) Reference 1.

R.E. Ginna Nuclear Power Plant PTLR-14 Revision 4

PTLR END NOTES

1. (Reference 1)

2. (Methodology of Reference 3, Attachment VI and Reference 6, as calculated in Reference 7.)

3. (Methodology of Reference 3, Attachment VI and Reference 6, as calculated in Reference 3, Attachment VII.)

R.E. Ginna Nuclear Power Plant PTLR-1 5 Revision 4