ML17265A136

Rev 2 to RCS Pressure & Temp Limits Rept (Ptlr)

ML17265A136
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Site:	Ginna
Issue date:	12/23/1997
From:	ROCHESTER GAS & ELECTRIC CORP.
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ML17265A135	List: ML17265A134 ML17265A136
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NUDOCS 9801140119
Download: ML17265A136 (15)

Text

GINNA STATION PTLR Revision 2 RCS PRESSURE AND TEMPERATURE LIMITS REPORT (PTLR)

R ponsible Nanager

/Z-z 3 P7 Effective Date Controlled Copy No.

9801140119 980106 pDR hooch osoooaea P

PDR

R.E.

Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report Revision 2

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 Temperature....................

3 2.3 Low Temperature Overpressure Protection System Setpoints 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM.........

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4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES.................

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5.0 REFERENCES

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5 FIGURE 1

FIGURE 2 TABLE 1 TABLE 2 TABLE 3 TABLE 4 TABLE 5 TABLE 6 Reactor Vessel Heatup Limitations...........

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6 Calculation of Chemistry Factors Using Survei Capsule Data.................................

llance

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10 Reactor Vessel Toughness Table (Unirradiated)

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11 Reactor Vessel Surface Fluence Values at 19.5 and 32 EFPY......

11 Calculation of ARTS at 24 EFPY..............................

12 Reactor Vessel Cooldown Limitations...............'...........

7 Surveillance Capsule Removal Schedule...................

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B Comparison of Surveillance Material with RG 1.99 Predictions..

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PTLR Revision 2

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 -

MODE 5, Loops Filled Pressurizer Safety Valves Low Temperature Overpressure Protection (LTOP) System PTLR Revision 2

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 l. 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 4, Enclosure 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)

(Methodology of Reference 3, Attachment VI and Reference 6,

as calculated in Reference 7).

2.2.1 The enable temperature for the Low Temperature Overpressure Protection System is 322'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 VI and Reference 6,

as calculated in Reference 3, Attachment VII).

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

I PTLR Revision 2

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 I 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, Ropy which is determined in accordance with ASTH 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 ASHE Boiler and Pressure Vessel Code.

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

As shown by Reference I (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:

l.

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 r RT>>y 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 mater'ial.

4.

The Charpy specimen irradiation temperature matches the reactor vessel surface interface temperature within + 25 F.

5.

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

4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES 4.I The RTpyg value for Ginna Station limiting beltline material is 256.6'F for 32 EFPY per Reference l.

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.

PTLR Revision 2

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 24 EFPY for the limiting reactor vessel material.

5.0 REFERENCES

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

dated June 1996.

2.

3.

4, 5.

6.

7.

WCAP-14040-NP-A, "Hethodology Used to Develop Cold Overpressure Hitigating System Setpoints and RCS Heatup and Cooldown Limit Curves,"

Revision 2, January 1996.

Letter from R.C. Hecredy, RG8E, 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.

Letter from R.C. Hecredy, RGI%E, to Guy S. Vissing, NRC, "Clarifications to Proposed Low Temperature Overpressure Protection System Technical Specification," dated June 3, 1997.

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

1 Reactor Vessel Radiation Surveillance Program,"

Hay 1969.

Letter from R.C. Hecredy, RGI%E, to Guy S. Vissing, NRC, "Corrections to Proposed Low Temperature Over pressure Protection System Technical Specification," October 8, 1997.

RGIIE Design Analysis DA-HE-97-031, "Evaluation of Ginna RCS Coolant Temperature to Support LTOPS Requirements,"

Revision 0.

I PTLR Revision 2

~ MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALVVELP SA-847 LIMITINGART VALUES AT 24 EFPY.

1/4T, 232 F 3/4T, 196 F 2500 i

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UNACCEPTABLE OPERATION HZATUP RATE UP 20 00 F/Hr.

HHATUP RATE UP 20 I00 ?/Hr.

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POB TBB 5BBTICB PBRIOD UP TO 24

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I 100 150 200 250 300 350 400 450 500 Indicated Temperature (Deg.F.)

FIGURE I REACTOR VESSEL HEATUP LIHITATIONS APPLICABLE FOR THE FIRST 24 EFPY (WITHOUT HARGIN FOR INSTRUHENT ERRORS)

PTLR Revision 2

MATERIALPROPERTY BASIS LIMITINGMATERIAL: CIRCUMFERENTIALWELD SA-847 LIMITINGART VALUES AT 24 EFPY:

1/4T, 232 F 3/4T, 196'F 2500

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FIGURE 2

REACTOR VESSEL COOLDOWN LIMITATIONS APPLICABLE FOR THE FIRST 24 EFPY (WITHOUT MARGIN FOR INSTRUMENT ERRORS)

Revision 2

PTLR 50 100 150 200 250

~00 350 400 450 500

Table 1

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

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

770 257'7'7'37'47'.99 3.00 1.85 1.74 1.74 1.9 1.6 (removed) 2.7 (removed) 7 (removed) 17 (removed)

TBD' Standby

.5028 1.105 1.864 3.746 T8D'b'/A NOTES:

(a)

Effective Full Power Years (EFPY).

(b)

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

(c)

Reference l.

PTLR Revision 2

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

E > 1.0 HeV)"

.5028 1.105 1.864 3.746

.5028 1.105 1.864 3.746

.5028 1.105 1.864 3.746

.5028 1.105 1.864 3.746 Predicted"

('F) 26 32 37 42 37 46 52 59 135 168 191 218 Heasured"

('F) 25 25 30 42 60 140 165 150 205 90 100 95

( F) 37 46 52 41 13 (a)

Reference 1 (including its Reference 51).

TABLE 3 Calculation of Chemistry Factors Using Surveillance Capsule Data Haterial Intermediate Shell Forging 05 (Tangential)

Capsule Fluence (x 10'~

n/cm'j E >1.0 HeV)"

.5028 1.105 1.864 3.746 FF

.8081 1.0279 1.1706 1.3418 aRT (o F)PN) 25 25 30 42 Sum:

FF*aRTNov

( F) 20.2 25.7 35.1 56.4 137.4 FF2

.6530 1.0566 1.3703 1.8004 4.8803 Intermediate Shell Chemistry Factor - 28.2'F

.5028

.8081 0

0 1.105 1.0279 0

0 1.864 1.1706 0

0 3.746 1.3418 60 80.5 Sum:

80.5 Chemistry Factor

= 16.5'F

.6530 1.0566 1.3703 1.8004 4.8803 Weld Hetal

.5028

.8081 149.7 121.0

.6530 1.105 1.0279 176.4 181.3 1.0566 1.864 3.746 1.1706 160.4 1.3418 219.1 187.8 1.3703 294.0 1.8004 NOTES:

(a)

Reference 1.

Sum:

- 854.69 4.8803 Chemistry Factor

= 160.7'F (b) zRT>>~ for weld material is the adjusted value using the 1.069 ratioing factor per Reference 1 applied to the measured values of Table 2

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PTLR 10 Revision 2

TABLE 4 Reactor Vessel Toughness Table (Unirradiated)"

Material Description Intermediate Shell Lower Shell Circumferential Weld CU

(%)

.07

.05

.25 Ni {%)

.69

.69

.56 Initial RT>>{ F) 20 40

-4.8 (a)

Per Reference l.

TABLE 5 Reactor Vessel Surface Fluence Values at 19..5 and 32 EFPY" x 10" (n/cm~,

E > 1.0 HeV)

EFPY 19.5 32 po 2.32 3.49 15 1.47 2.20 30 1.05 1.56 45'969 1.45 (a)

Reference l.

PTLR Revision 2

TABLE 6 Calculation of Adjusted Reference Temperatures at 24 EFPY for the Limiting Reactor Vessel Haterial Parameter Operatin Time Haterial Location Circ. Weld 1/4-T Chemistry Factor (CF),

F"i Fluence (f), 10'/cm (E > 1.0 HeV)~o>

Fluence Factor FF zRT>>

CF x FF, F

Initial RTNpy (I)

F Hargin (H), 'F" ART - I + (CFxFF)

+ H, F""

NOTES:

(a)

Value calculated using Table 5 values.

{b)

Values from Table 3.

(c)

Reference l.

160.7 1.85 1.17 188

-4.8 48.3 232 Values 24 EFPY Circ. Weld 3/4-T 160.7

.851

.955 153.4

-4.8 48.3 196.9 PTLR 12 Revision 2