ML081280095

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Supplement to License Amendment Request: Methodology for Determining Reactor Coolant System Pressure and Temperature and Low Temperature Over Pressure Limits
ML081280095
Person / Time
Site: Ginna Constellation icon.png
Issue date: 04/25/2008
From: John Carlin
Constellation Energy Group
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML081280095 (24)
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Text

John Carlin R.E. Ginna Nuclear Power Plant, LLC Site Vice President 1503 Lake Road Ontario, New York 14519-9364 585.771.5200 585.771.3943 Fax John.Carlin@constellation.com 0 Conso-etellation Energy' Nuclear Generation Group April 25, 2008 U. S. Nuclear Regulatory Commission Washington, DC 20555 ATTENTION:

Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 Supplement to License Amendment Request: Methodology for Determining Reactor Coolant System Pressure and Temperature and Low Temperature Over Pressure Limits

Reference:

(1) Letter to the USNRC Document Control Desk from John Carlin (Ginna),

License Amendment Request: Methodology for Determining Reactor Coolant System Pressure and Temperature and Low Temperature Over Pressure Limits, dated February 9, 2008 (2) Letter to Robert Mecredy (RG&E) from Pao-Tsin Kuo (NRC), License Renewal Safety Evaluation Report for the R.E. Ginna Nuclear Power Plant, dated March 3, 2004 (3) Letter to USNRC Document Control Desk from Mary Korsnick (Ginna LLC),

License Amendment Request Regarding Extended Power Uprate, dated July 7, 2005 On February 9, 2008, The R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC) submitted a request for a change in the methodology used to determine Reactor Coolant System (RCS) pressure and temperature limits (Reference 1). On March 5, 2008, a conference call was held between Ginna LLC personnel and members of the NRC Staff. During that call, Ginna LLC agreed to submit a draft Pressure Temperature Limits Report (PTLR) using the methodology described in Reference (1). Attachment (1) to this letter contains the requested draft PTLR.

Attachment (2) contains WCAP-15885, R.E. Ginna Heatup and Cooldown Limit Curves for Normal Operation, Revision 0, which provides the data used to develop Attachment (1).

N1, L1

-QA

Document Control Desk April 25, 2008 Page 2 WCAP-15885 was developed for Ginna's license renewal effort to demonstrate the ability for safe operation into the requested period of extended operation, using approved ASME Code Cases 588, 640 and 641. The NRC accepted the evaluation for that purpose, but did not approve the new curves for use, indicating that a future technical specification amendment would be required (see Reference 2, section 4.2.5). The above mentioned code cases were eventually incorporated into WCAP-14040A, Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves, Revision 4 (the methodology requested in Reference 1). Because WCAP-15885 predates WCAP-14040-A, Revision 4, Westinghouse and Ginna LLC have reviewed WCAP-15885 and determined that it is consistent with WCAP-14040-A, Revision 4 methodology. However, it should be noted that the curves in the draft PTLR differ from those in WCAP-15885 in that they contain instrument uncertainties.

Also, the PTLR curves are considered valid to only 47.3 EFPY due to the increased fluence resulting from Ginna's subsequent power uprate (see Reference 3, Attachment 5, section 2.1.2.2.5).

This submittal does not contain any new regulatory commitments.

Should you have questions regarding this matter, please contact Mr. Thomas Harding at (585) 771-3384 or Thornas.Iarding(iiwconstellation.com.

R.E. Ginna Nuclear Power Plant, LLC

Document Control Desk April 25, 2008 Page 3 STATE OF NEW YORK TO WIT:

COUNTY OF WAYNE I, John T. Carlin, being duly sworn, state that I am Vice President, R.E. Ginna Nuclear Power Plant, LLC (Ginna LLC), and that I am duly authorized to execute and file this request on behalf of Ginna LLC.

To the best of my knowledge and belief, the statements contained in this document are true and correct. To the extent that these statements are not based on my personal knowledge, they are based upon information provided by other Ginna LLC employees and/or consultants. Such information has been reviewed in accordance with company practice and I believe it to be reliable.

Subsc ibed and sworn before me, a Notary Public in and for the State o ew York and County of t n.-

, this cd

" day of,r-l

,2 8.

WITNESS my Hand and Notarial Seal:

My Commission Expires:

RICHARD A. JOHNSON NOTARY PUBLIC, STATE OF NEWYORK O*/*.*

    • r'lO*No.

01J06082344

'OvQUALIFIED IN WAYNE COUNTY I

Date MY COMMISSION EXPIRES &f--

JC/MR

Attachment:

(1) Draft PTLR (2) WCAP-15885 cc:

S. J. Collins, NRC D.V. Pickett, NRC Resident Inspector, NRC (Ginna)

P.D. Eddy, NYSDPS J. P. Spath, NYSERDA R.E. Ginna Nuclear Power Plant, LLC Draft PTLR

PTLR Constellation Energy R.E. Ginna Nuclear Power Plant RCS Pressure and Temperature Limits Report PTLR Revision%

_.U

//

be Responsible Manager, A, P" George Wrobef Effective Date:

6/29/05 Controlled Copy No.

Record Cat.# 4.43.3 4 5

RevisionI R.E. Ginna Nuclear Power Plant PTLR-1

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 (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 R.E. Ginna Nuclear Power Plant PTLR-2 Revision

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.

Reference 1 calculates Pressure/Temperature Limits out to 52 EFPY pre-Extended Power Uprate (EPU). Reference 9 determines the data is Reference 1 is valid out to 47.3 EFPY post-EPU. The titles and labels in the PTLR will show the 47.3 EFPY.

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 60°F per hour.

b.

A maximum cooldown of 100°F per hour.

2.1.2 The RCS PIT limits for heatup and cooldown are specified by Figure PTLR - 1 and Figure PTLR - 2, respectively. These curves are based on Reference 1 as modified in Reference 12 to include instrument errors.

2.1.3 The minimum boltup temperature, using the methodology of Reference 4, is 600 F(Reference 12)r 2.2 Low Temperature Overpressure Protection System Enable Temperature\\

(LCO 3.4.6)

(Calculated in Reference 12)

(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"2°F.

2.3 Low Temperature Overpressure Protection System Setpoints (LCO 3.4.12) 2.3.1 Pressurizer Power Operated Relief Valve Lift Setting Limits\\

(See Reference 12)

The lift setting for the pressurizer ower Operated Relief Valves (PORVs) is

< 1 (includes instrument uncertainty).

R10r R.E. Ginna Nuclear Power Plant PTLR-3 Revision

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 edule meets the requirements of ASTM E185-82.

As shown by ReferencA (spe--iically 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.

RG 1.99 Rev. 2 Regulatory Position 1.1.

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 PTLR-4 Revision

PTLR 4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES 4.1 The RTp 1 value fbr Cini--

Station lim-iting belt'ine ma.terial is 266.6 0F f"_O Z2 EFP3 Y per Rfeteie-+t The RTPTs value for Ginna Station limiting beltline material is 273.1°F for welds and 128.2°F for forgings per Reference 12.

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.

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

5.0 REFERENCES

1 WG,-A P

  • ,-14 84 "r.EG- -,:___

tu and Limit Norm*,Oeal-e*,

16.

WCAP-15885, Revision 0. "R. E. Ginna Heatup and Cooldown Limit Curves for Normal Operation," dated July 2002.

2.

WCAP-14040-TW-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision', Jantuey-96.May 2004.

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

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 RC&^"

Design Analysis A-M 97-031, "Eval.. ato. of G.n.. ReCS Coant Temperatro, to Support LnTOPSITD Requirements," Rovi*i*o 0.

7. WCAP-14684. "R. E. Ginna Heatup and Cooldown Limit Curves for Normal Operation," dated June 1996.1
8. Letter from M. Korsnick, CEG, to US NRC Document Control Desk,

Subject:

R.E. Ginna Nuclear Power Plant, Licensee Amendment Request Regarding Extended Power Uprate, (Attachment 5 - Licensing Report), dated July 7, 2005. CMIS Record ID 1001353.

19. CN-RCDA-04-149, Revision 2, "Ginna Extend Power Uprate Program Reactor Vessel Integrity Evaluations."
10. WCAP-13902, "Analysis of Capsule S from the Rochester Gas and Electric Corporation R. E. Ginna Reactor Vessel Radiation Surveillance Program," dated December 1993.
11. BAW-1 803, Revision 1, "Correlations for Predicting the Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds," dated May 1991.
12. DA-ME-08-020, Revision 0, "Pressure Temperature Limit Report (PTLR) Supporting Analysis," dated April 21, 2008.

5 RevisionI R.E. Ginna Nuclear Power Plant PTLR-6

PTLR MATERIAL PROPERTY RARIA ING MATERIAL. CIRCUMFERENTIAL WELD SA-847 ING ART VALUES AT 32 EFPY:

1/4T, 241"F 3/4T. 207"F 2500

  • o 2250 2000 1750 4)

L.

1500 1250 1000 co 750 500 250 0

100 1e0 260 250 360 dica ted Temperatur R. E.

acR Figure PTLR-1 eactor Coolant System Heatup Limitations (Heatup Rates of 6N Applicable to 32 EFPY (Without Margins for Instrumentation Errors' 100-F)

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

Material Property Basis Limiting Material: Inter. to Lower Shell Forging Girth Weld and Inter. Shell Forging Limiting ART Values at 47.3 EFPY:

1/4T, 256F (Circ Flaw ART), 112F (Axial Flaw ART) 3/4T, 223F (Circ Flaw ART), 103F (Axial Flaw ART)

HU 60F/hr -

-HU 10OF/hr -

-- -60 Critical Limit -

-100 Critical Limit......

LeakTest I 2500 2250 2000 1750 1500 1250 1000 750 500 250 0

-I I

II

-(----*

-e-

-I--

Unacceptable Operation

/I 4"

/

Acceptable III I

Il1 Operationi I---

-J I

'I

~--

i

]

I I

I I-i 0

50 100 150 200 250 300 350 Temperature (F) 400 450 500 Figure PTLR-1 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates up to 100 F/Hr) Applicable for the First 47.3 EFPY (Including Normal Instrument Errors) (Reference 12)

R.E. Ginna Nuclear Power Plant I

I eiion 5

PTLR MATERIAL PROPERTY BASIS LIMITING MATERIAL: CIRCUMFERENTIAL WELD SA-847 IMITING ART VALUES AT 32 EFPY:

1/4T. 241'F 3/4T. 207"F 2500"-'r" 2250 Cf 2000 1750 1 1 1[i

/

l ltI I IOPERATION

=

1500 0'1250 cr*

1250ACCEPTABLIZ J

OP XR AT 10N 1000 10 1 See Next Page 750ooo==

. 250 1 f1

-T-0 50 160.

150 200 250 300 350 400 450 500 I ieated Temperatu (Deg.F)

Figure PTLR - 2 R. E. Ginna eactor Coolant System Cooldown Limitations (Cooldown Rates of 20 40 60 and 1 'F/hr) Applicable to 32 EFPY (Without Marains for Instrumentation FF nr-R.E. Ginna Nuclear Power Plant PTLR-8 Revision

Material Property Basis Limiting Material: Inter. to Lower Shell Forging Girth Weld and Inter. Shell Forging Limiting ART Values at 47.3 EFPY:

1/4T, 256F (Circ Flaw ART), 112F (Axial Flaw ART) 3/4T, 223F (Circ Flaw ART), 103F (Axial Flaw ART)

I-CD OF/hr -

CD 20F/hr - -----

CD 40F/hr -

-CD 60F/hr -

CD 100F/hr I 2500 2250 I I I I 11111111111 2000 Unacceptable Operation 1750 1500 A;il-

ý4##

1 1-1 ý-I I I I I 1250 1000 750 5O0 250 0~

I.

t 1 ~

Acceptable I

Operation

-I-0 50 100 150 200 250 300 350 400 450 500 Temperature (F)

Figure PTLR-2 R. E. Ginna Reactor Coolant System Cooldown Limitations (Cooldown Rates up to 100 F/Hr) Applicable for the First 47.3 EFPY (Including Normal Instrument Errors) (Reference 12)

R.E. Ginna Nuclear Power Plant I eiion 5

PTLR Table PTLR - 1 Surveillance Capsule Removal Schedule (a)

Effective Full Power Years (EFPY).

(b)

Reference 1.

(c)

To be le eur,,nn quir,,

fOf rein, oval.

Capsule N will be removed at shortly after receiving a fast neutron fluence equivalent to operation to 2029 (60 year license). The fluence on Capsule N will be between 1 and 2 times the peak end of life fluence. Removal is scheduled for the Spring Outage of 2008.

(d) Capsule P will be removed shortly following receiving a fast neutron fluence equivalent to operations to 2049. The specific withdrawal EFPY and fluence will be determing following the analysis of Capsule 5

RevisionI R.E. Ginna Nuclear Power Plant PTLR-9

PTLR Table PTLR - 2 Surveillance Material 30 ft-lb Transition Temperature Shift (b) Using equations of RG 1.99 Revision 2, with material chemistry of Table PTLR-4, plus 2 standard deviations of ARTNDT (1 7F for forges, 28F welds) per Generic Letter 96-03 Reviewer Note 7.

(c) Reference 11 R.E. Ginna Nuclear Power Plant PTLR-1 0 Revision 4

PTLR Table PTLR - 3 Calculation of Chemistry Factors Using Surveillance Capsule Data Intermn ate Shell Forgi 05 (Tangential)

Intermediate Shell Weld Metal (a)

R erence 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-11I Revision 4

TABLEA/L]

Calculation of Chemistry Factors using R.E. Ginna, Turkey Point & Davis Besse Surveillance Capsule Data Material Capsule Capsule f(a)

FF(b)

ARTNDT(C)

FF*ARTNDT FFa Lower Shell V

0.587 0.851 25 21.275 0.724 Forging 125P666 R

1.02 1.006 25 25.150 1.012 T

1.69 1.144 30 34.320 1.309 S

3.64 1.335 42 56.070 1.782 SUM:

136.815 4.827 CFLsF125p666 = 7(FF

F 2) = (136.815) + (4.827) 28.30F Intermediate Shell V

0.587 0.851 0

0 0.724 Forging 125S255 R

1.02 1.006 0

0 1.012 T

1.69 1.144 0

0 1.309 S

3.64 1.335 60 80.1 1.782 SUM:

80.1 4.827 CF 1ss=

12 (FF

  • RTNrDT) + *( FF 2) = (80.1) + (4.827) = 16.61F Ginna Surveillance V

0.587 0.851 149.8 (140) 127.480 0.724 Weld Metal R

1.02 1.006 176.6 (165) 177.660 1.012 (Heat # 61782)

T 1.69 1.144 160.5(150) 183.612 1.309 S

3.64 1.335 219.4 (205) 292.899 1.782 SUM:

781.651 4.827 CFHt # 6178 2 = X(FF

  • RTNDT) + X( FE2) = (781.651) + (4.827) = 161.9 0F Turkey Point Davis 2.956 1.287 221 (215) 284.427 1.656 Surveillance Weld T (TP3) 0.699 0.900 163 (166) 146.700 0.810 Material(")

V (TP3) 1.484 1.109 176 (179) 195.184 1.230 (Heat # 71249)

T(TP4) 0.673 0.889 208 (211) 184.912 0.790 SUM:

811.223 4.486 CF Ht#71249 =- (FF

FE2) = (811.2230F) + (4.486) = 180.80F See Next Page for Notes

)

I I

IR. E. Ginna Nuclear Power Plant[

I Po,,i~inrn RI Liii Lo.~, IZI

Notes:

(a) f= fluence. See Table 3, (x 1019 n/cm2, E > 1.0 MeV).

(b)

FF = fluence factor = &,2-08log (c)

ARTNDT values are the measured 30 ft-lb shift values taken from the following documents:

- Ginna Plate and Weld... WCAP-14684.

- Turkey Point & Davis Besse... WCAP-15092 R.3 of the inlet temperature for each capsule that was removed (d)

Ginna operates with an average in approximately 5490F, Turkey Point 3&4 operate with an average inlet temperature of approximately 546*F, and Davis Besse operates with an average inlet temperature of approximately 5550F. The measured ARTNDT values from the Turkey Point 3&4 surveillance program were adjusted by subtracting 31F to each measured ARTNDT and the Davis Besse surveillance program data was adjusted by adding 60F to the measured ARTNDT value before applying the ratio procedure.

The surveillance weld metal ARTNDT values have been adjusted by a ratio factor of:

Ratio Ginna = 1.07, Ratio Turkey Point = 1.0 (conservative), Ratio Davis Besse = 1.0 (conservative)

The pre-adjusted values are in parenthesis. Since Turkey Point and Davis Besse material is similar to Gi nna's, this is acceptable.

.)

R. E.

Ginna Nuclear Power Planti'-R*

Rvo*

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

Material Description Cu (%)

Ni (%)

Initial RTNDT (IF)

Intermediate Shell

.07

.69 20 Lower Shell

.05

.69 40 Circumferential Weld

.25(c

.56 H'I

-4.8 (b)

(a)

Per Reference 1.

(rReference 11.

\\

(c) For use in Table PTLR-2, material for the Circumferential Weld is based on Table 1 of Reference 1: Cu 0.23%

and Ni 0.53%

Vessel Flange I

n/a I

n/a I

-52 PTLR->e[3 5

RevisionI R.E. Ginna Nuclear Power Plant

PTLR

[32--Table PTLR-5 Reactor Vessel Surface Fluence Values at"'.

and EFPY(a) x 10 19 (n/cm 2, E > 1.0 MeV)

EFPY

0.

150 300 450 (a 9.5 2."5 1.48 32 49-*45-6-30-66 2.26

-+*5-.

2005 (a)

Reference 1.

PTLR-Iýn14 5

RevisionI R.E. Ginna Nuclear Power Plant

PTLR Table PTLR - 6 Calculation of Adjusted Reference Temperatures a FPY for the Limiting Reactor Vessel Material Parameter Values Operating Time 47.3 4 3 EFPY Inter.

Inter.

Material Circ. Weld Circ. Weld h*erl ShellShl Location 1/4-T 114-T 3/4-T Chemistry Factor (CF), OF(a) 4I' 6e-.7 H

E 4160e-.-7 Fluence (f), 1019 n/cm2 (E > 1.0 MeV)(b) 42.I

[13.-

4!-68 Fluence Factor (FF)

B 01 H

1.*O.2 1.

ARTNDT = CF x FE, IF 5

Initial RTNDT (I), IF

-4 (d

-4.

(Ld)

Margin (M), OFb

48. (d
48. (dL ART = I + (CVxle)

+ M, OF(b)lc) PTL-24-1 23 (a) Values from Table PTLR - 3.

(b)

Value calculated using Table PTLR - 5 values.

(c)

Reference 1.

d) Per Referee 11.

PTLR-

[1]5 5

RevisionI R.E. Ginna Nuclear Power Plant

PTLR END NOTES

  • k (Reference 1 )
2.

(Methodology of Reference 3, Attachment VI and Reference 6, as feference 7.)

3.

( ethodology of Reference 3, Attachment VI and Reference 6, as Re rence 3, Attachment VII.)

.E. Ginna Nuclear Power Plant PTLR-1 5 Revision 4 WCAP-15885

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