ML102730623

R. E. Ginna - Transmittal of RCS Pressure and Temperature Limits Report (PTLR)

ML102730623
Person / Time
Site:	Ginna
Issue date:	09/22/2010
From:	Harding T Constellation Energy Nuclear Group, EDF Group
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
WPLNRC-1002339
Download: ML102730623 (19)

Text

Thomas Harding Director, Licensing R.E. Ginna Nuclear Power Plant, LL(

CENG a joint venture of 1503 Lake Road Ontario, New York 14519-9364 585.771.5200 585.771.3943 Fax thomas.hardinpqr(,cenqllc com Enem, i Cnsellation #% D I September 22, 2010 U.S. Nuclear Regulatory Commission Washington, DC 20555-0001 ATTENTION: Document Control Desk

SUBJECT:

R.E. Ginna Nuclear Power Plant Docket No. 50-244 Transmittal of RCS Pressure and Temperature Limits Report (PTLR)

REFERENCES:

(1) Letter from T. Harding, Ginna LLC to NRC Document Control Desk,

Subject:

Commitment Change Associated with the Submittal of a Revised Pressure Temperature Limits Report, dated February 18, 2010 (2) Letter from T. Harding, Ginna LLC to NRC Document Control Desk,

Subject:

Additional Information Associated with Revised Pressure Temperature Limits Report Commitment Change, dated April 13, 2010 In accordance with the R.E. Ginna Nuclear Power Plant Improved Technical Specification 5.6.6, which requires the submittal of revisions to the PTLR, the attached report is hereby submitted.

The commitment date for submitting the attached PTLR was revised to October 1, 2010 by Reference 1. Additional information to support the revised commitment date was verbally requested by the NRC staff and provided by Reference 2.

P'0OO

ýUPq P6 -10,P~93-39

Document Control Desk September 22, 2010 Page 2 There are no new commitments being made in this submittal. If you should have any questions regarding the information in this submittal, please contact Tom Harding at (585) 771-5219 or Thomas.HardingJr(cengllc.com.

Attachment:

Ginna PTLR, Revision 6 c: M. Dapas, NRC D.V. Pickett, NRC Resident Inspector, NRC (Ginna)

Attachment Ginna PTLR, Revision 6 R.E. Ginna Nuclear Power Plant, LLC September 22, 2010

PTLR a joi'nt ve6nture of L

,R.E. GINNA NUCLEAR POWER'PLANT RCS Pressure and Temperature Limits Report PTLR Revision 6 Responsible Manager; Effective Date:

R.E. Ginna Nuclear Power Plant PTLR-1 of 16 .Revision 6

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 o!f 16 ýRevision 6

. PTLR 2.0 OPERATING LIMITS The cycle-specific parameter limits for the specifications listed in Section 1.0 are presented in the following subsections. AIlchanges t6theSe, limits "mustbe developed usinig th'e NRC approved methodologies specified in Technical Specification 556.6. These limits have been determined suchthat'all applicable limits of the safety analysis are met. All, itemfis that' apper in capitalized type are defined in Technical Specification 1.1, Definitions. Reference1 Cailculates Pressure/

Temperature Limits out to 53 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'rmaximumrheatup of 600FP'er 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 ,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 minimumboltup temperature; using the methodology of Reference 4, Enclosure 2 is 60OF (Reference 12).

2.2' Low Temperature Overpressure Protection ,System Enable Temperature (Calculated In Reference 12)

(LCO 3.4.6)

(LCO 3.4.7)

(LCO314.10)

(LCO 3.4.12) 2;2.1 The enable tempe'raturelfor the Low Temperature Overpressure Protection System is 3220 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 Power Operated Relief Valves (PORVs) is

• <410 psig (includes instrument uncertainty).

R.E,,Ginna Nuclear Power Plant ,PTLR!*3,,of ý16 IRqvisiqn_.6

PTLR 3.0 REACTOR VESSEL MATERIAL SURVEILLANCE PROGRAM The reactor vessel material irradiation sureillance specimens shall be removed and,-_

,examined to0determine changes in materialjproper.ties., The removal-schedule is provided in Table PTLR 1. -The results of-these examinations shall be used~toupdate.

.Figure:PTLR - 1 and Figure PTLRr- 2., -

The pressure vessel steel surveillance program (Ref. 5 as modified by Ref. 10) is in compliance with Appendix H to 10 CFR 50, entitled, "Reactor Vessel Material Surveillance Program Requirements." Theý"ma6terial test re*quireffients 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, "Fra6ture Toughness Criteria for'Protettion Against Failure," to section Xl of the ASME Boiler andPressure Vessel .Code. The surveillance capsule removal schedule meets the requirements of ASTM E185-82.

As shown by Reference 10 (Appendix D), the reactor vessel material irradiation specimensinindicate surveillancepresented idisCUssioh Regulatory the surveillance thatGuide' data meets the credibility 1.99 Revision 2where:"

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 shelfenergy unambiguously.

3. The scatter of ARTNDT values are within the best fit scatter limits as shown on Table PTLR - 2 for the surveillance weld material. The scatter of ARTNDT values anr*hot within th scatter limits as sh*wn on.Table PTLR - 2for the-st sfi Intermediate and Lower Shell Forging materiaIs, Which use RG 1:99 Rev. 2 Regulatory Position 1.1.

4. The Charpy specimen irradiation temperature matches the reactor, vessel 0 . !"I" .

surface interface temperature within +/- 25 F.

5. The

,surveillance data falls withimthe scatter band of the material database.

R.E' Ginna Nuclear Power Plant PTLR-4 of 1'15 .. I Revision'6

PTLR 4.0 SUPPLEMENTAL DATA INFORMATION AND DATA TABLES, 4.1 The RTPTS value for 53 EFPY post-EPU for Ginna Station limiting beltline material is 275 0 F for welds and 143°F for forgings per Reference 1.

4.2 Tables Table PTLR - 1 contains the location and schedule for the removal of surveillance capsules.

Table PTLR'- 2*contains acomparison of measured surveillance material30 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 chdmistry 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.

Table PTLR - 6 shows examplecalculations of the ART values at 53 EFPY. for the limiting reactor vessel material.

5.0 REFERENCES

I1.. W C AP-1,72,14-NP,,Revision 0, "R. E.,,Ginna Heatup and.Cooldown Limit-Curves for Normal Operation and(Pressurized ThermalShock.Evaluation,"..dated July, 2010.

2. WCAP-14040-A, "Methodology Used to Develop Cold Overpressure Mitigating System Setpoints and RCS Heatup and Cooldown Limit Curves," Revision 4, 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,of 16 fRevislon,6

PTLR

6. Letter from R.C Me'credy',,RG&E'to Guy8. Vi'ssing,NRC, "Correc6tionrs-t6 P'oposed' Low Temperature Overpressure Protection System Technical Specification," October 8,.1997.

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

8. Letter from M. Korsnick, CEG, to US NRC Document Control Desk,

Subject:

R. E.

Ginna-Nuclear Power Plant, Licensee Amendment ReqUest Regarding Ektended Power Power Uprate. (Attachment 5 - Licensing Report), dated July 7, 2005.

9. CN-RCDA-04-149,. Revision 2,2"Ginna Extended Power, Uprate Program.Reactor Vessel Integrity Evaluations.",

10. WCAP-17036-NP, Revision 1, "Analysis of Capsule N from the R. E. Ginna Reactor Vessel Radiation Survelilance Program," dated September 2010.

11. BAW-1803, Revision 1, "Correlations for Predicting tle Effects of Neutron Radiation on Linde 80 Submerged-Arc Welds," dated May 1991.

12. DA-ME-08-020, Revision 2, "Pressure Temperature Limit Report (PTLR) Supporting Analysis;"!dated August'5,'201 0.'

13. Safety Evaluation by the Office of Nuclear Reactor Regulation Related to Amendment

-.;No. 97 to;Renewed Facility, Operating;License ,No.. DPR-18 R. E. Ginna Nuclear Power Plant, Docket No. 50-244.

14. LTR-AMLRS-10-26, Revision 0, "R. E. Ginna Surveillance Capsule P Withdrawal Recommendations," dated September 9, 2010. .

15. Safety Evaluatio6nby'the Office of Nuclear Reactor Regulation Related t6'Amendment I1.ý '106to Rehen"edFa6clit*'0perating Licnhse"No ::DPR-18' "R. E. Ginna:Nuclear Power Plant, LLC, Docket No. 50-244," U. S. NRC, February 23, 2009.

R.E. Girnna Nuclear Power Plant PTLR-&of 16 Revision 6

PTLR aterial Property Basis (Reference 1)

Limiting Material: Inter to Lower Shell Forging Girth Weld and Lower Shell Forging Limiting ART Values at 53 EFPY: 1/4T, 262°F (Circ Flow ART), 136*F (Axial Flaw ART) 3/41T, 231'F {Cfrc Flaw ART), 1Z7*F (Axial Flaw ART)

HI-lU60F/hr ---- HU 100/he - - 60 Critical Limit ...... lOOCriticaI Limit - -LekTest 2500 7- 1 T t ~ t 1 -

- "7- z r 2250 SUNACCEPTABLE

._j 7o OPERATION 1500 -;'- *

... 4-J .............

1250 ,'-* 4OPERATION 1000 ]--- ..

750 t

-T I- I-.-

0 0 500 4

0 0 50 100 150 200 250 300 350 400 450 500 Temporaturef 'F)

Figure PTLR -1 R. E. Ginna Reactor Coolant System Heatup Limitations (Heatup Rates up to 100°F/hr) Applicable for the First 53 EFPY (Including Normal Instrument Errors) (Reference 12)

R.E. Ginna Nuclear Power Plant PTLR-7 of 16 .Revision 6

PTLR Material Property Basis (Reference 1)

Limiting Material: Inter to Lower Shell Forging Girth Weld and Lower Shell Forging Limiting ART Values at 53 EFPY! 1/&T, 262°F (CIrc Flaw ART), 136°F (AxialFlaw ART)

.3/4T, 231°F (Circ Flaw ART), 127'F (Axial Flaw ART) 7 - CDOF/hr ........ CD 20F/hr -

• C- 40F/hr ..... CO 60F/hr CD 100F/hr 2500 2250 I

2000 1750 1500 1250 1000 750 500 250 0 50 100 150 200 250 300 350 400 450 500 Temperature ( F)

Figure PTLR- 2 R. E. Ginna Reactor CoolanttSystem Cooldown Limitations (Cooldown Rates of up to 100°F/hr)

I Applicable for the First 53 EFPY (Including Normal' Instriument Errors) (Reference 12)

R.E., Ginna Nuclear Power Plant OTLR-8 of 16 Revision 6

PTLR Table PTTLR - I Surveillance Capsule Removal Schedule(a)

Capsule, Vessel Location (deg.) Capsule Lead Factor (b) Removal Schedule Capsule Fluence 2

EFPY (c) E19(n/cm )(b)

,V 770 2.96 1.4 (removed) 0.587 R -257 2.97.. . 2;6 (removed) ... 1.02 T 670 1.82 6.9 (removed) 1.69 S- 570 1.79 17 (removed) 3.64 N 2370 1.82 30.5 (removed) 5.80 x 1019 P 2470 1.90 (d) (d)

(a) Reference 10.

(b) Updated.in Capsule N dosimetry analysis (c) EFPY from plant startup (d) The latest Capsule P should be removed is shortly after the vessel accumulates a fluence of

.39.9 EFPY, which corresponds to a maximum 80 year fluence of 76 EFPY for the Capsule.

The earliest withdrawal for Capsule P should be shortly after the vessel accumulates a fluence of 33.9 EFPY. This correlates to acceptable withdrawal for Capsule P at EOC 36, 37, 38, 39, or 40 in order to fulfill the commitment of Reference 13 to pull the final capsule shortly following accumulation of 80 years offluence. (Reference 10 and Reference 14)

R.E. Ginna Nuclear Power Plant :PTLR_9 of 1& evision 16

PTLR Table PTLR - 2 Surveillance Material 30ft-lb TransitionTemperature Shift 30 lb-ft Transition Temperature Shift (ARTNDT)

Fluence 19 2 Measured(b)

(x'10 n/cm , Predicted(a)

Material Capsule E > 1.0 MeV) (IF) (IF)

V 0.587 26.4 34.7 R 1.02 31.2 57.5 Lower Shell T 1.69 35.5 33.6 S 3.64 41.4 45.8 N 5.8 44.3 91.1 V 0.587 37.4 0.0 (c)

R 1.02 44.2 20.1 Intermediate Shell. T 1.69 50.4 0.0 (c)

S 3.64 58.8 76.8 N 5.8 6276.4 V 0.587 1356.2 146.7 R 1.02 159.7 156.2 Weld Metal T 1.69 181.8 149.7 S 3.64 212.1 212.2 N 5.8 227.2 216.9 V 0.587 - 30.7 R 1.02 58.6 HAZ Metal T 1.69 41.0 S 3.64 - 38.9 N 5.8 - 107.7 R.E. Ginna Nuclear Power Plant PTLRAO of 16  !ýý:Rdvision 6

PTLR (a) Based on Regulatory Guide. 1.99, Revision 2, methodology using the mean "weight percent valus of copper and nickel of t'esurveillance material.

I  ;(b) 'Calculatedin Appendix C of Reference 10.

.(c) Measured ARTNDT value wasdetermined-to be negative, but physicallya.

reduction should not occur, therefore a conservative value of zero is used.

R.E. Ginna Nuclear Power Plant PTLR-,l 1 of.1 6 tReylision 6

PTLR TablePTLR-3 ,

I Calculation of' Chemistry Factors using, R. E.,Ginna and T"ukey Point Surveillance Capsule Data Material Capsule Capsule FF(b) ARTNDT(C) Adjusted,, FF*ARTNDT FF 2 f(s) ARTNDT(d)

. 0;587 0.851 0.0 (e), 0 0.724 R 1.02 1.006 20.10 20.2 1.011 T 1,69 1.144 .0(e - 0 1.31 Intermediate Shell Forging 125S255 S 3.64 1.335 76.80 - 102.6 1.783 (L-C)

N 5.8 1.430 76.40 -- 109.3 2.046 Sum: 232.1 6.875 CF 125 S255 = _(FF

• ARTNDT) + Y(FF 2 ) = (232.1) + (6.875) = 33.8 0F V 0.587 0.851 34.70 -- 29.5 0.724 R 1.02 1.006 57.50 --- 57.8 1.011 T 1.69 1.144 33.60 -- 38.5 1.31 Lower Shell Forging S 3.64 1.335 45.80 - 61.2 1.783 125P666 (L-C)

N 5.8 1.430 91.10 - 130.3 2.046 Sum: 317.3 6.875 CF 125 P666 = E(FF

• ARTNDT) + E(FF 2 ) = (317.3) + (6.875) = 46.2 0F V 0.587 0.851 146.70 157.0 133.6 0.724 R 1.02 1.006 156.20 167.1 168.1 1.011 T 1.69 1.144 149.70 160.2 183.3 1.31 Ginna Surveillance Weld Metal (Heat # S 3.64 1.335 212.20 227.1 303.2 1.783 61782)

N 5.8 1.430 216.90 232.1 332 2.046 Sum: 1120.2 6.875 CFHt. #61782 = YE(FF

• ARTNoT) + I(FF 2 ) = (1120.2) + (6.875) = 162.9 0 F RE'.Ginna Nuclear Power Plant OTLR-12 6fit :Revis,io n"6

PTLR Table PTLR- 3 Calculation of Chemistry Factors using R. E. Ginna a Turkey Point Surveillance Capsule Data Material Capsule Capsule FF(b) ARTNDTMC, Adjusted FF*ARTNDT FF2 (a) (d)

.ARTNDT"d)

T 0.599 0.856 163.87 147.50 126.3 0.734 Turkey Point V 1.223 1.056 180.77 162.09 171.2 1.115 Surveillance Weld X 2.897 1.282 191.06 170.98 219.2 1.644 I Materiali (Heat # ..... ..

71249) Sum: 516.8 3.493 I

CFHt.#71249 = _(FF

• ARTNDT) + 7(FF 2) = (516.80F) + (3.493) 147.9 0 F I

(a) f = fluence (x,101 9n/cm 2, E > 1.0 MeV) from Table 6-6 of Reference 10 for Ginna data and Table D-5 of Reference 10 for Turkey Point Data.

(b) FF= fluence factor = f(°0. 8

-0.1 log (f))

(c) ARTNDT OF values are the measured 30 ft-lb shift values taken from Table 5-10 of I Reference 10 for Ginna data and Table D-5 of'Reference 10 for Turkey Pointdata.

(d).. To address the difference in chemistry factor between the surveillance weld and the QGinna vessel weld of the same heat,jthe surveillance weld metal ARTNDT values have been adjusted in accordance with Appendix D ofiReference 10 using the chemistryfactor ratio of: -.1.07 for Heat #61782, and 0.86 for Heat #71249, JThe chemistry factor ratio for Heat #61782 is derived from Table 2-3 of Reference 14, and for Heat #71249 the ratio is shown in Appendix D of Reference 10. Also, adjustments were made to the measured ARTNDT Turkey Point data to account for the operating temperature differences between the Ginna and Turkey Point vessels.

(e) Measured ARTNDT value was determined to be negative, but physically a reduction I should not occur, therefore a conservative value of zero is used.

R.E. Ginrna Nuclear Power Plant PTLM3,ý of 16 .ReVilsion. 6

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

Material Description Cu (%) NI (%) Initial RTNDT ('F)

Reactor Upper Closure r/a n/a 0 Head Flange Intermediate Shell .07 .69 20 Lower Shell .05 .69 40 IS to LS Circumferential .25 .56 -4.8 I Weld Vessel Flange n/a n/a -52

______ 4 4 1

- 068 30 Nozil~Shdll

..... .*....9*-- 'NStO IS Circumferenrtial Weld

. . O.23 0.59 10 (a) Per Reference ITable 2-1 and Table 2-2 (b) The nozzle shell forging weight-percent copper value of 0.17 was taken from Reference 15.

Section 3.3, P-T Limits: Staff Evaluation, of Reference 15 states: "The staff determined that

.an appropriate Cu value for Ginna RPV nozzle forging should be close to 0.17 percent, the highest Cu content for the RPV shell and nozzle forgings of the entire domestic fleet based on the RVID."

R.E Ginna Nuclear Power Plant PTLR-14 of 16 Revisi on 6

PTLR Table PTLR - 5 Reactor Vessel Surface Fluence Values at 30.5 and 53 EFPY(a) x 10 19 (n/cm 2, E > 1.0 MeV)

EFPY 00 150 300 450 30.5 3.20 2.01 1.45 1.31 53 5.56 3.42 2.46 2.30 (a) Reference 10 Table 6-2A R.E. Ginna Nuclear Power Plant PTLR-15 of 16 Revision 6

PTLR Table PTLR - 6 Calculation of Adjusted Reference Temperatures at 53 EFPY for the Limiting Reactor Vessel Material(a)

Parameter Values Operating Time 53 EFPY I Material Inter. to Lower Inter. to Lower Lower Shell Lower Shell Shell Circ. Shell Circ.

I Weld Weld Location 1/4-T 1/4-T 3/4-T 3/4-T Chemistry Factor (CF), OF 162.9 46.2 162.9 46.2

-- F-luence'(f)*!09---n1cm2.(*-t.e) -3.7.64- -3.764- -1..226- -- J-,7.26_

- Fluence-Factor.(FF) ~-- i+/-: -1.3429- ý,113429 - -< ~~Sl- - 0 H ARTNDT = CF x FF, OF 218.8 62 187.3 53.1 Initial RTNDT (I), OF -4.8 40 -4.8 40 I

Margin (M), OF *48.3 34 48.3 34 H

ART = I + (CFxFF) + M, OF 262 136 231 127 (a) Per Reference I Table 4-2 and 4-3 R.E. Ginna Nuclear Power Plant PTLR-16 of 16 Revision 6