ML17264A776

Provides Revised Pressure & Temperature Limits Rept to Complete All Outstanding Util Commitments

ML17264A776
Person / Time
Site:	Ginna
Issue date:	12/30/1996
From:	Mecredy R ROCHESTER GAS & ELECTRIC CORP.
To:	Vissing G NRC (Affiliation Not Assigned), NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NUDOCS 9701140197
Download: ML17264A776 (25)

Text

CATEGORY 1 I

REGULATO INFORMATION DISTRIBUTION STEM (RIDS)

ACCESSION NBR:9701140197 DOC.DATE: 96/12/30 NOTARIZED:

NO DOCKET FACIL:50-244 Robert Emmet Ginna Nuclear Plant, Unit 1, Rochester G

05000244 AUTH.NAME AUTHOR AFFILIATION MECREDY,R.C.

Rochester Gas 6 Electric Corp.

RECIP.NAME RECIPIENT AFFILIATION VISSINGPG.S.

SUBJECT:

Provides revised pressure

& temperature limits rept to complete all outstanding util commitments.

DISTRIBUTION CODE:

AOOID COPIES RECEIVED:LTR L ENCL 1

SIZE:

TITLE: OR Submittal: General Distribution NOTES:License Exp date in accordance with 10CFR2,2.109(9/19/72).

05000244 C

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N NOTE TO ALL "RIDS" RECIPIENTS:

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4ND ROCHESTER GASANDELECTRIC CORPORATION ~ 89 EASTAVENUE, ROCHESTER, N. Y 146d9-OOOI AREA CODE716 5'.27OO ROBERT C. MECREDY Vice President Nvcteor Operotions U.S. Nuclear Regulatory Commission Document Control Desk Attn:

Guy S. Vissing Project Directorate I-l Washington, D.C. 20555 December 30, 1996

Subject:

Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR)

Rochester Gas &Electric Corporation

- R.E. Ginna Nuclear Power Plant Docket No. 50-244

References:

(a)

Letter from R.C. Mecredy, RG&E, to G.S. Vissing, NRC, Reactor Coolant System (RCS) Pressure and Temperature Limits Report (PTLR), dated December 2, 1996.

(b)

Letter from G.S. Vissing, NRC, to R.C. Mecredy, RG&E, KIT. Ginna Ginna

- Acceptance ofRequest to Extend TimeforApproval ofRevision ofPressure and Temperature LinIitsReport (PTLR) (TACNo. M97313), dated December 10, 1996.

(c)

Letter from R.C. Mecredy, RG&E, to G.S. Vissing, NRC, Request to Use ASME Code Case N-514 in the DeteI'nIination of Low TeInperature Overpressure Protection (LTOP), dated December 18, 1996.

(d)

Letter from R.C. Mecredy, RG&E, to G.S. Vissing, NRC, Applicationfor Amendment to Facility Operating License, Revision to Reactor Coolant System (RCS) Pressure and TenIperature Limits Report (PTLR), dated September 13, 1996.

Dear Mr. Vissing By Reference (a), RG&E requested an extension to the NRC's acceptance of the current Ginna Station PTLR from December 31, 1996 to July 1, 1997, for reasons as stated in the letter. The NRC approved this request in Reference (b) provided that RGkE submit the followingby December 31, I

1996:

a.

Request to allow use ofASME Code Case N-514; and b.

Revised PTLR incorporating all NRC comments received to date.

I r 970ii40i97 9hi230 PDR ADOCK 05000244 P

PDR The request to use ASME Code Case N-514 was provided by Reference (c). Therefore, the purpose ofthis letter is to provide the revised PTLR to complete all outstanding RG&E commitments.

The revised PTLR is attached.

There is only one change to the PTLR since that provided in Reference (d). This change revises the 3/4-T ARTvalue listed in Figures 1 and 2 and Table 6 to 196.9'F from 196'F. This is due to a round-off error and does not affect the heatup/cooldown curves. Allother sections of the PTLR remain unchanged.

The cover sheet and pages 6, 7, and 12 are denoted as Rev. 2, 12/96. Allother pages are identical to the September 13, 1996 submittal.

Please contact George Wrobel, Manager ofNuclear Safety and Licensing at (716) 724-8070 ifyou have further questions.

Very t ly yours, Robert C. Mecredy MDF<898 xc:

U.S. Nuclear Regulatory Commission Mr. Guy S. Vissing (Mail Stop 14C7)

PWR Project Directorate I-1 Washington, D.C. 20555 U.S. Nuclear Regulatory Commission Region I 475 Allendale Road King ofPrussia, PA 19406 Ginna Senior Resident Inspector

GINNA STATION PTLR Revision 2, 12/96 RCS PRESSURE AND TEMPERATURE LIMITS REPORT

{PTLR)

Responsible Hanager Effective Date Controlled Copy No.

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 TEHPERATURE LIMITS REPORT...................-....

2 2o0 OPERATING LIMITS o ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ ~

~

o 3

2. 1

"'RCS Pressure and Temperature Limits..........................

3 2.2 Low Temperature Overpressure Protection System Enable emperature................................................

T 3

2.3 Low Temperature Overpressure Protection System Setpoints.....

3 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM......................

• 4 4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES.......................

4

5.0 REFERENCES

5 FIGURE 1

FIGURE 2 TABLE 1 TABLE 2 TABLE 3 Reactor Vessel Heatup Limitations............................

6 Reactor Vessel Cooldown Limitations..........................

7 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 TABLE 6 Calculation of ARTS at 24 EFPY..............................

12 Reactor Vessel Toughness Table (Unirradiated)

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

11 PTLR Revision 2

IP

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 -

MODE 4 RCS Loops -

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

0

2.0 OPERATING LIHITS 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 tate-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.

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)

(Hethodology of Reference 3, Attachment II, Section 3.4 using 1/4T Ropy value from Reference 1).

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 (Hethodology of Reference 3, Attachment II as calculated in Reference 4, Attachment IV)

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

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 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, 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 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 a,RT>> values are within the best fit scatter limits as shown on Table 2.

The only exception is with respect to.

th'e Intermediate Shell which is not the limiting reactor vessel material.

5.

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

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

4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES

4. 1 The RTpyg value for Ginna Station limiting beltline material is 259. 1 F

for 32 EFPY per Reference I.

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.

j PTLR Revisi'on 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

2.

3.

WCAP-14684, "R.E. Ginna Heatup and Cooldown Limit Curves for Normal Oper ation," dated June 1996.

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

Letter from R.C. Mecredy, RG8E, 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.

5.

Letter from R.C. Hecredy, RGLE, 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,"

May 1969.

PTLR Revision 2

MATERIALPROPCRTY BASIS LIMITINGMATCRIAL: CIRCUMFCRENTIALWELD SA-047 LIMITINGART VALUES AT 24 CFPY:

1/4T, 232 F 3/4T, 3.96. 9" F 2500 j

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FIGURE I REACTOR VESSEL HEATUP LIMITATIONS APPLICAI3LE FOR THE FIRST 24 EFPY Revision 2

PTLR 50 100 1.50 200 250 300 350 400 450'00

MATEBIALP BOP CBTY BASIS LIMITINGMATEBIAL: CIBCUMFCBCNTIALWELD SA-847

. LIMITINGABT VALUES AT 24 EFPY:

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FIGURE 2

REACTOR VESSEL COOLDOMH LIMITATIONS APPLICABLE FOR THE FIRST 24 EFPY Revision 2

50 100 150 200 250 300 350 400 450 500 Indicated.Temperature (Deg.pj

Table 1

Survei.llance Ca sule Removal Schedule

.Vessel Location Capsule (deg.)

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

77 257 67o 57'37'47'.99 3.00 1.85 1.74 1.74 1.9 1.6 (removed) 2.7 (removed) 7 (removed) 17 (removed)

Teo'b'tandby

.5028 1.105 1.864 3.746 Teo~bi N/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 Material 30 ft-lb Transition Tem erature Shift 30 lb-ft Transition Temperature Shift Material Lower Shell Intermediate Shell Meld Metal HAZ Metal Capsule S

Fluence (x 10" n/cm',

E > 1.0 MeV)"

.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" (oF 26 32 37 42 37 46 52 59 135 168 191 218 Measured" of) 25 25 30 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 Usin Surveillance Capsule Data Material Intermediate Shell Forging 05 (Tangential)

E Capsule S

Fluence (x 10" n/cm',

E) 1.0

~eV)<>

.5028 1.105 1.864 3.746 FF

.8081 1.0279 1.1706 1.3418

~RT

(

o F)PN) 25 25 30 42 Sum:

FF*hRTgp7

( F) 20.2 25.7 35.1 56.4 137.4

-FF

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

.5028

.8081 0

0 S

3.746 1.3418 60

-80.5 Sum:

80.5 Chemistry Factor 16.5'F R

1.105 1.0279 0

T 1.864 1.1706 0

0

.6530 1.0566 1.3703 1.8004 4.8803 Weld Netal

.5028

.8081 149.7 121.0

.6530 1.105 1.0279 176.4 181.3 1.0566 1.864 1.1706 160.4 187.8 1.3703 3.746 1.3418 219.1 294.0 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 Tabl'e 2.

PTLR 10 Revision 2

TABLE 4 Reactor Vessel Toughness Table (Unirradiated)"

Material Description Intermediate Shell Lower Shell Circumferential Meld Cu

(%)

.07

.05

.25 Ni (1o)

.69

.69

.56 Initial RTgQ7( F)-

20 40

-4.8 (a)

Per Reference 1.

TABLE 5 Reactor Vessel Surface Fluence Values at 19.5 and 32 EFPY'"

x 10'n/cm',

E > 1.0 MeV)

EFPY 19.5 32 0o 2.32 3.49 15'.47 2.20 30'.05 1.56 45'969 1.45 (a)

Reference l.

PTLR Revision 2

4 l,

TABLE 6 Calculation of Adjusted Reference Temperatures at 24 EFPY for the Limiting Reactor Vessel Material Parameter 0 crating Time Material Location Chemistry Factor (CF), 'F'"

Fluence (f), 10" n/cm (E > 1.0 MeV)"

Fluence Factor FF hRTgpy CF x FF >

F Initial RT(I), 'F Margin (M), 'F" ART I + (CFxFF)

+

M F""'OTES:

(a)

Value calculated using Table 5 values.

(b)

Values from Table 3.

(c)

Reference l.

Circ. Weld 1/4-T 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, 12/96

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