ML17264A400

Rev 0 to RCS Pressure & Temp Limits Rept.

ML17264A400
Person / Time
Site:	Ginna
Issue date:	02/24/1996
From:	ROCHESTER GAS & ELECTRIC CORP.
To:
Shared Package
ML17264A399	List: ML17264A398 ML17264A400 ML17264A401
References
NUDOCS 9603140435
Download: ML17264A400 (15)

Text

GINNA STATION PTLR Revision 0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

Responsible Hanager Effective Date Controlled Copy Ho.

9b03i40435 9'b03ii 05000244 PDR ADOCK PDR P

R.E. Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report Revision 0 This report is not part oF the Technical Specifications. This report is referenced in the Technical Specifications.

TABLE OF CONTENTS 1 ~ 0 RCS PRESSURE AND TEMPERATURE LIMITS REPORT ........................ 2 2.0 OPERATING LIMITS ................................................... 3

2. 1 RCS Pressure and Temperature Limits .......................... 3 2.2 Low Temperature Overpressure Protection System Enable T emperature .............. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ 3 2.3 Low Temperature Overpressure Protection System Setpoints ~ ~ ~ ~ ~ 3 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM.

4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES....................... 4 5 .0 REFERENCES ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ t ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 FIGURE 1 Reactor Vessel Heatup Limitations ............................ 6 FIGURE 2 Reactor Vessel Cooldown Limitations ......... . .. ....... 7 TABLE 1 Surveillance Capsule Removal Schedule. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 8 TABLE 2 Comparison of Surveillance Material with RG 1.99 Predictions.. 9 TABLE 3 Calculation of Chemistry Factors Using Surveillance Capsule Data. ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 10 TABLE 4 Reactor Vessel Toughness Table (Unirradiated)

TABLE 5 Reactor Vessel Surface Fluence Values at 7 and 21 EFPY......

TABLE 6 Calculation of ARTS at 21 EFPY...........; 12 PTLR Revision 0

R.E. Ginna Nuclear Power Plant Pressure and Temperature Limits Report 1.0 RCS Pressure and Tem erature Limits Re ort PTLR This Pressure and Temperature Limits Report (PTLR) for Ginna Station 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 (P/T) 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 PTLR Revision 0

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 Tem erature Limits (LCO 3.4.3 and LCO 3.4. 12)

(Reference 1)

2. 1. 1 The RCS temperature rate-of-change limits are:

a. A maximum heatup of 60'F per hour.

b. A maximum cooldown of 100 F per hour.

2.1.2 The RCS'P/T limits for heatup and cooldown are specified by Figures 1 and 2, respectively.

2. 1.3 The minimum boltup temperature, using the methodology of Reference 2, is 60'F.

2.2 Low Tem erature Over ressure Protection S stem Enable Tem erature (LCOs 3.4.6, 3.4.7, 3.4.10 and 3.4.12)

(Reference 1) 2.2. 1 The enable temperature for the Low Temperature Overpressure Protection System is 330'F.

2.3 Low Tem erature Over ressure Protection S stem'et pints (LCO 3.4. 12) 2.3. 1 Pressurizer Power 0 crated Relief Valve Lift Settin Limits (Reference 1)

The lift setting for the pressurizer Power Operated Relief Valves (PORVs) is ~ 424 psig.

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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 1. The results of these examinations shall be used to update'Figures 1 and 2.

The pressure vessel steel surveillance program (Ref. 3) 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, RTg>> which is determined in accordance with ASTM E208. The empirical relationship between RT>> 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 E185-82.

As shown by Reference 4, 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 ~RT>> values are within the best fit scatter limits as shown on Table 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 + 25'F limits.

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

4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES 4.1 The RT>>~ value for Ginna Station is 275.2 F for 32 EFPY per Reference

5. (Note - these values are based on Capsule T. The new revised RTpys values based on Capsule S will be implemented following NRC review of these values).

4.2 Tables Table 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.

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

Table 4 provides the reactor vessel toughness data.

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

Table 6 shows example calculations of the ART values at 21 EFPY for the limiting reactor vessel material.

5.0 REFERENCES

Letter from A.R. Johnson, NRC, to R.C. Hecredy, RGKE,

Subject:

"Issuance of Amendment No. 48 to Facility Operating License No. DPR-18, R.E. Ginna Nuclear Power Plant (TAC No. H79828)," dated March 6,'992.

2. WCAP-14040, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and,RCS Heatup and Cooldown Limit Curves," Revision 1, December 1994 as approved by letter from C. I. Grimes, NRC to R.H.

Newton, WOG,

Subject:

"Acceptance for Referencing of Topical Report WCAP-14040, Revision 1", (TAC No. H91749), dated October 16, 1995.

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

4. WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R.E. Ginna Reactor Vessel Radiation Surveillance Program,"

dated December 1993.

5. Letter from George E. Lear, NRC to Roger Kober, RGE E, "Safety Evaluation by Office of Nuclear Reactor Regulation Regarding Projected Values of Material Properties for Fracture Toughness Requirements for Protection Against Pressurized Thermal Shock Events, Rochester Gas and Electric Company, R.E. Ginna Nuclear Power Plant Docket No. 50-244 (TAC No.

59956)," dated November 17, 1986.

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2,500 2,250 2,000 1,750 C$

Unaccgptable 1,500 g ~

Operatfon 11~

60 F/hr eV 1,0QO K

0 0

0 50 100 150 200 250 300 350 400 450 500 RCS COLD LEG TEMPERATURE ('F)

FIGURE 1 REACTOR VESSEL HEATUP LIMITATIONS APPLICABLE FOR THE FIRST 21 EFPY PTLR Revision 0

UMING MATERIAL CIRCVMFEREN11AL WELD UMHlNG1/4T RT~r= 240'F UMITING3/4T RT~ 178'F 2000

<750 Vnacceptable Operatton C9

~ 1500 co 1250 co 1000 O

Cooldown Rates Acceptabte 760 /Hr OperatIon 0

20 500 40 66

'0O~

250 0 50 100 150 200 250 300 350 400 450 500 RCS COLO LEG TEMPERATURE ('F)

FIGURE 2 REACTOR VESSEL COOLDOWN LIHITATIONS APPLICABLE FOR THE FIRST 21 EFPY PTLR Revision 0

Table 1 Surveillance Ca sule Removal Schedule Vessel Capsule Capsule Location Lead Fluence Capsule (deg.) Factor Removal Schedule" E19 (n/cm')

77o 2.99 1.6 (removed) 0.556 257 3.00 2.7 (removed) 1.15 1.85 7 (removed) 1.97 67'7'370 1.74 17 (removed)" 3.87 1.74 Too<'>

TOD"'tandby 247 1.9 . N/A NOTES (a) Effective Full Power Years (EFPY).

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

(c) Currently under NRC review, not included in current heat up/cooldown curve.

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TABLE 2 Surveillance Material 30 ft-lb Transition Tem erature Shift 30 lb-ft Transition Temperature Shift Fluence (x 10" n/cm', E > 1.0 Predicted" Measured Haterial Capsule 'F) ('F) ('F)

HeV)"'703 26 25 1.01 32 25 Lower Shell 1.75 37 30

.703 37 37

~ 1.01 46 Intermediate Shell 1.75 52 52

.703 135 140 1.01 168 165 Weld Hetal 1.75 191 150

.703 1.01 90 HAZ Hetal 1.75 100 (a) Based upon Reg. Guide 1.99, Revision 2 predictions (b) Letter from 'A. Johnson (NRC) to R. Hecredy (RG&E) "Issuance of Amendment No. 48 to Facility Operating License No. DPR-18, R.E. Ginna Nuclear Power Plant (TAC No. H79828)," dated March 6, 1992 especially Reference (5) of Section 3:1.2.

TABLE 3 Calculation of Chemistry Factors Usin Surveillance Capsule Data Fluence (x 10'/cm',

E >1.0 h Ropy FF*~RTwov Haterial Capsule HeV) ( F) (oF) FF Intermediate .703 0.901 25 22.5 .812 Shell Forging 05 1.01 1.003 25 25.1 1.006 (Tangential) 1.75 1.154 30 34.6 1.332 Sum: 82.2 3.15 Chemistry Factor" 26. 1 Intermediate .703 0.901 0 .812 Shell 1.01 1.003 0 1.006 1.75 1.154 0 1.332 Sum: 0.0 3.15 Chemistry Factor" 0.0 Weld Hetal .703 .901 140 126.1 ..812 1.01 1.003 165 165.5 1.006 1.75 '1.154 150 173.1 1.332 Sum: 464.7 3. 15 Chemistry Factor" = 147.5'F NOTES:

(a) Letter from A. Johnson (NRC) to R. Hecredy (RGEE) "Issuance of Amendment No. 48 to Facility Operating License No. DPR-18, R.E. Ginna Nuclear Power Plant (TAC No. H79828)," dated Harch 6, 1992 especially Reference (5) of Section 3. 1.2.

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TABLE 4 Reactor Vessel Toughness Table (Unirradiated)"

Haterial Description CU (%) Ni Po) Initial Ropy( F)

Intermediate Shell .07 .69 20 Lower Shell .05 .69 40 Circumferential Weld .25 Letter from "

(a) A. Johnson (NRC) to R. Hecr edy (RG&E) Issuance of Amendment No. 48 to Facility Operating License No. DPR-18, R.E. Ginna Nuclear Power Plant (TAC No. H79828)," dated Harch 6, 1992 especially Reference (5) of Section 3. 1.2.

TABLE 5 Reactor Vessel Surface Fluence Values at 7 and 21 EFPY" x 10'n/cm', E > 1.0 HeV)

EFPY Oo 30 44.5' 14.5'538

.866 .359 .310 21 2.32("1 1.42 0.991 .893 (a) WCAP-11026 "R.E. Ginna Reactor Vessel Fluence and RT>>s Evaluations,"

dated December 1985 Table II.2-3 through II.2-6, as referenced from License Amendment No. 48.

(b) Letter from A. Johnson (NRC) to R. Hecredy (RG8E) "Issuance of Amendment No. 48 to Facility Operating License No. DPR-18, R.E. Ginna Nuclear Power Plant (TAC No. H79828)," dated Harch 6, 1992 especially Reference (5) of Section 3. 1.2.

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TABLE 6 Calculation of Adjusted Reference Temperatures at 21 EFPY for the Limiting Reactor Vessel Haterial Parameter Values 0 crating Time 21 EFPY Haterial Circ. Weld Circ. Weld Location 1/4-T 3/4-T Chemistry Factor (CF), 147.5 147.5 (f), ~10'~ n/cm~ (E F"'l.uence

) 1.0 HeV)" 1.57 .720 Fluence Factor (FF)" 1.125 .908 hRTgpy CF x FFi F 165.9 133.93 Initial RT~ (I), F Hargin (H),

ART - I + (CFxFF) + H, F" 210 178'F NOTES:

(a) Value calculated from Table 4 note" reference.

(b) Values calculated from Table 3 and Table 4.

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