ML17264A679

Rev 1 to RCS Pressure & Temperature Limits Rept (Ptlr)

ML17264A679
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Site:	Ginna
Issue date:	05/24/1996
From:	ROCHESTER GAS & ELECTRIC CORP.
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ML17264A623	List: ML17264A622 ML17264A678 ML17264A679
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NUDOCS 9610090394
Download: ML17264A679 (20)

Text

GINNA STATION PTLR Revision 1

RCS PRESSURE AND TEMPERATURE LIIMITS REPORT (PTLR)

R sponsible Manager s ~-(l~~

Effective Date Controlled Copy No.

9610090394 961004 PDR ADOCK 05000244 I

P PDR

R.E.

Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report Revision 1

Per letter from J.A. Hitchell, NRC, to R.C. Hecredy,

RGIIE,

Subject:

"R.E. Ginna - Acceptance for Referencing of Pressure Temperature Limits Report, Revision 1

(TAC No. H94770)," this revision of the PTLR is only valid until December 31, 1996 at which time NRC approval of a revised PTLR is required for continued operation.

Note:

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 2.2 RCS Pressure and Temperature Limits........................

Low Temperature Overpressure Protection System Enable Temperature..................

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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

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5 FIGURE 1

Reactor Vessel Heatup Limitations...........................

6 FIGURE 2 Reactor Vessel Cooldown Limitations 7

TABLE 1

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TABLE 2 TABLE 3 Surveillance Capsule Removal Schedule........

8 Comparison of Surveillance Material with RG 1.99 Predictions..

9 Calculation of Chemistry Factors Using Surveillance apsule Data................................................

C 10 TABLE 4 TABLE 5

~i TABLE 6 Calculation of ARTS at 22 EFPY..............................

12 Reactor Vessel Toughness Table (Unirradiated)

Reactor Vessel Surface Fluence Values at 17 and 32 EFPY......

11 PTLR Revision 1

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 3.4.6 3.4.7 3.4.10 3.4.12 RCS Pressure and Temperature (P/T) Limits RCS Loops -

NODE 4 RCS Loops -

NODE 5, Loops Filled Pressurizer Safety Valves Low Temperature Overpressure Protection (LTOP) System PTLR Revision 1

2.0 OPERATING LIMITS

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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 l. 1, "Oefinitions."

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, Section 2.7, is 60'F.

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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)

(Methodology of Reference 3, Attachment II, Section 3.4 using 1/4T RT>>

value from PTLR Figures 1 and 2) 2.2. 1 The enable temperature for the Low Temperature Overpressure Protection System is 328'F.

2.3 Low Tem erature Over ressure Protection S stem Set pints (LCO 3.4. 12) 2.3. 1 Pressurizer Power 0 crated Relief Valve Lift Settin Limits (Methodology of Reference 3, Attachment II as calculated in Reference 4, Attachment IV)

The lift setting for the pressurizer Power Operated Relief Valves (PORVs) is ~ 411 psig (includes instrument uncertainty).

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PTLR 3

Revision 1

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.

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, RTg>>

which i s determined in accordance with ASTH E208.

The empiri cal 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 ASHE Boiler and Pressure Vessel Code.

The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

As shown by References 6 and 7, 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.

3.

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

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 SUPPLEHENTAL DATA INFORHATION AND DATA TABLES

4. 1 The RTpyg value for Ginna Station limiting beltline material is 266.5 F

for 32 EFPY per References 6,

7, and 8.

4.2 Tables

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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.

PTLR Revision 1

l e

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 22 EFPY for the limiting reactor vessel material.

5.0 REFERENCES

2.

3.

Letter from R.C. Hecredy, RG&E, to A.R. Johnson, NRC,

Subject:

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

Pressure and Temperature Limits Report (PTLR)," dated April 22, 1996.

WCAP-14040, "Methodology Used to Develop Cold Overpressure Hitigating System Setpoi'nts 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.

Letter from R.C. Hecredy, RG&E, to A.R. Johnson, NRC,

Subject:

"Technical Specification Improvement

Program, Reactor, Coolant System (RCS)

Pressure and Temperature Limits Report (PTLR);" dated December 8, 1995.

4.

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6.

Letter from R.C. Hecredy, RG&E, to A.R. Johnson, NRC,

Subject:

"Application for Amendment to Facility Operating License, Hethodology for Low Temperature Overpressure Protection (LTOP) Limits," dated February 9,

1996.

WCAP-7254, "Rochester Gas and Electric, Robert E. Ginna Unit No.

1 Reactor Vessel Radiation Surveillance Program,"

Hay 1969.

WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R.E.

Ginna Reactor Vessel Radiation Surveillance Program,"

dated December 1993.

7.

8.

Letter from A.R. Johnson, NRC, to R.C. Hecredy, RG&E,

Subject:

"R.E.

Ginna Nuclear Power Plant - Pressurized Thermal Shock Evaluation (TAC No. M93827)," dated March 22, 1996.

Letter from Westinghouse Electric Corporation (SE-REA-96-072) to G.J.

Wrobel, RG&E,

Subject:

"Vessel Fluence Re-evaluation for R.E. Ginna,"

dated April 16, 1996.

PTLR Revision 1

l1

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Leak Tes t Limi t FOR 20 CFPY

-250 2000 L tMITING MATGAIAL-CI ACUMFCAENTIAL WCLO LIMITING 1 4T ATHOT 23 LIMITIHG 3/4T AT

= 208'F HOT 1750 1500 12SO a.

1000 atd 750 OK Unacceptable Operation 60 'F/hr Acceptable Operation 500 250 Criticality Limit based on 1nservice Hydro-static Test Temp.

0 50 100 150 200 250 300 350 400

. 450 500 (NOICATEO TEMPERATURE: (OCG.F) i FIGURE I RCACTOR VCSSEL HEATUP LIMITATIONS APPLICADLC FOR THE'IRST 22 Cl PY PTLR Revision I

2500 p(

LIHITING HAT>RIAL 2250 2000 C IRCVHFERCNTIAL WELO LIHITING 1/4T RTOT= 2 LfHITING 3/4T RT

= 208'F HOT 17SO 1SOO 12SO c

1000 D

EJ 7SO OK 500 250 CooIdown Rates F/hr 0

, '20~

QO60~

100 Vnacceptab1e Oper ation Acceptable Operation 0

50 100 150 200 2SO 300 3SO tNOICATCO TEMPERATURE (OEG.F)

FIGURE 2

REACTOR VESSEL COOLOOWN LIHITATIONS APPLICABLE FOR THE FIRST 22 EFPY Revision 1

Table 1

Surveillance Ca sule Removal Schedule Vessel Location Capsule (deg.)

Capsule Lead Factor Removal Schedule" Capsule Fluence E19 (n/cm')"

I 77' R

257' 67 S

57 N

237' 247 2.99 3.00 1.85 1.74 1.74 1.9 1.6 (removed) 2.7 (removed) 7 (removed) 17 (removed)

Teo"'tandby 0.612 1.242 2.128 4.180 Teo'b'/A

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NOTES:

(a)

Effective Full Power Years (EFPY).

(b)

To be determined, there is no-current requirement for removal.

(c)

Reference 6 with ENDFB-VI adjustment per Reference 8.

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PTLR Revision 1

TABLE 2 Surveillance Haterial 30 ft-lb Transition Tem erature Shift 30 lb-ft Transition Temperature Shift Haterial Lower Shell Intermediate Shell Weld Hetal HAZ Hetal Capsule Fluence (x 10" n/cm',

E ) 1.0 Hev)"

0.612 1.242 2.128 4.180 0.612 1.242 2.128 4.180 0.612 1.242 2.128 4.180 0.612 1.242 2.128 4.180 Predi cted" OF 26 32 37 37 52 50 135 168 191 218 Heasured" OF 25 25 30 60 140 165 150 205 90 100 95

( F) 37 52 13 (a)

Reference 6 with EHDFB-VI adjustment per Reference 8.

TABLE 3 Calculation of Chemistry Factors Using Surveillance Capsule Data Material Capsule Fluence (x 10" n/cm',

E > 1.0 MeV)<

wRT

(

o F)FN)

FF*~RTwov

('F)

FF Intermediate Shell Forging 05 (Tangential) 0.612 1.242 2.128 4.180 0.862 1.060 1.205 1.366 25 30 42 21.55 0.743 26.5 36.15 1.124 1;452 57.37 1.866 Sum:,

141. 57 5.185 Chemistry Factor - 136.39'F Intermediate Shell 0.612 1.242 0.862 0

0 0.743 1.060 0

0 1.124 2.128 1.205 0

0 1.452 4.180 1.366 60 81.96 1.866 Sum:

81. 96

5. 185 Chemistry Factor 15.81'F Weld Metal 0.612 1.242 2.128 4.180 0.862 158.2 136.37 0.743 1.060 186.45 197.64 1.124 1.205 169.5 204.25 1.452 1.366 231.65 316.43 1.866 Sum:

854.69

5. 185 Chemistry Factor

= 164.8'F NOTES:

(a)

Reference 6 with ENDFB-VI adjusted per Reference 8.

(b)

~RT>> for weld material is the adjusted value using the

1. 13 r atioing factor per Reference 7 applied to the measured values of Table 2.

PTLR 10'ycle 25, Revision 0

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TABLE 4 Reactor Vessel Toughness Table (Unirradiated)"

Haterial Description Intermediate Shell Lower Shell Circumferential Meld (a)

Per Reference 7.

Cu

(%)

.07

.05

.25 Ni (%)

.69

.69

.56 Initial RT>>( F) 20 4P

-4.8 TABLE 5 Reactor Vessel Surface Fluence Values at 17 and 32 EFPY" x 10'n/cm, E > 1.0 MeV)

EFPY 17 32 po 2.26 3.94 15o 1.38 2.44 30'.997 1.81 45o 0.886 1.65 (a)

Reference 6 with ENDFB-VI adjustment per Reference 8.

PTLR Cycle 25, Revision 0

4'cb 1

TABLE 6 Calculation of Adjusted Reference Temperatures at 22 EFPY for the Limiting Reactor Vessel Haterial Parameter 0 eratin Time Hateri al Location Chemistry Factor (CF),

F"'luence (f), ~10" n/cm'E

> 1.0 HeV)"

Fluence Factor FF hRTgpi' CF x FF y F

Initial RT>> (I),

F Hargin (H),

F" ART - I + (CFxFF)

+ H, F""

Circ. Weld 1/4-T 164.8 1.91 1.177 193.97

-4.8 48.3 238'F Values 22 EFPY Circ. Weld 3/4-T 164.8 0.875 0.963 158.70

-4.8 48.3 203'F NOTES'a)

Value calculated using Table 5 values.

(b)

Values from Table 3.

(c)

ART value is a conservative approximate per adjustment of ENDFB-VI.

PTLR 12 Cycle 25, Revision 0

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