ML20086P103

Application for Amend to License NPF-30,changing Tech Spec 3/4.4.9 Re Pressure/Temp Limits to Modify Plant Heatup & Cooldown Curves & Max Allowable PORV Setpoint Curve for Cold Overpressure Protection,Per Generic Ltr 91-01

ML20086P103
Person / Time
Site:	Callaway
Issue date:	12/18/1991
From:	Schnell D UNION ELECTRIC CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20086P105	List: ML20086P103 ML20086P602
References
GL-91-01, GL-91-1, ULNRC-2537, NUDOCS 9112260203
Download: ML20086P103 (9)

Text

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$ YONI ILLECTRIC er

'$3 December 18, 1991 U.S. Nuclear Regulatory Commission Attn:

Document Control Desk Mail Station P1-137 Washington, DC 20555 Gentlemen:

ULNRC-2 5 37 CALLAWAY PLANT DOCKET NUMBER 50-483 REVISION TO TECHNICAL SPECIFICATION 3/4.4.9 PRESSURE / TEMPERATURE LIMITS

References:

1)

Regulatory Guide 1.99, Revision 2, dated May, 1988 2)

NRC Generic Letter No. 88-11, dated July 12, 1988 3)

ULNRC-1867, dated November 30, 1988 4)

NRC Generic Letter 91-01, dated January 4, 1991 5) 10 CFR 50.61, Final Rule effective June 14, 1991 6)

ULNRC-2461, dated August 14, 1991 Union Electric Company herewith transmits an application for amendment to Facility Operating License Number NPF-30 for the Callaway Plant.

This amendment application requests that the plant heatup and cooldown curves and the maximum allowable PORV setpoint curve for cold overpressure protection, as found in Technical Specification Figures 3.4-2, 3.4-3, and 3.4-4 be modified.

Additionally, the react or vessel surveillance capsule removal schedule as given oy Technical Specification Table 4.4-5 would be removed from the Technical Specifications.

Following is a discussion of the requested amendments.

1)

Figure 3.4-2 is the heatup limitation curve.

This curve is revised to reflect the RT calculated from the surveillance capsulgD$ata.

The revised curve will be valid to 17 effective full power years (EFPY).

The curve is based on Regulatory Guide 1.99, Revision 2, and 10 CFR 50.61 (effective June 14, 1991).

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l U.S. Nuclctr Rrgulatory Commiccion December 18, 1991 Page 2 2)

Figure 3.4-3 is the cooldown limitation curve.

This curve is also revised to reflect the RTNDT calculated for 17 EFPY in the surveillance capsule report.

This curve is also based on Regulatory Guide 1.99, Revision 2 and 10 CFR 50.61.

3)

Table 4.4-5 is the surveillance capsule withdrawal schedule.

This table is removed from Technical Specifications in accordance with NRC Generic Letter 91-01.

This table will be included in the next revision of the Callaway FSAR and be maintained therein.

4)

Figure 3.4-4 is the maximum allowable PORV setpoint curve for cold overpressure protection.

This curve is revised due to the changes made in the heatup and cooldown limitation curves.

5)

Technical Specification Bases 3/4.4.9 is revised to reflect the change in the service life period and the removal of the surveillance capsule withdrawal schedule.

6)

Surveillance Requirement 4.4.9.1.2 is revised to reflect the removal of the surve131ance capsule withdrawal schedule, 7)

Table B 3/4.4-1 is revised to correct a typographical error.

This submittal satisfies the reporting requirements for NRC Generic Letter 88-11, 10 CFR 58.61, and 10 CFR 50, Appendix H.

NRC Generic Letter 88-11 requires licensees to evaluate Regulatory Guide (R.G.) 1.99, Revision 2 for applicability to their facilities.

A review of R.G.

1.99, Rev. 2 shows that tne exicting analysis for callaway's heatup, cooldown, and low temperature overpressure protection setpoints (Technical l

Specification Figures 3.4-2, 3.4-3, and 3.4-3, l

respectively) is boundod.

Union Electric commits to use the methodology presented in R.G.

1.99, Rev. 2 for future 10 CFR 50, Appendix G submittals (see Reference 3).

This i

letter confirms that the R.G.

1.99, Rev. 2 methodology has been used for this submittal.

10 CFR 50.61, " Fracture Toughness Requirements I

for Protection Against Pressurized Thermal Shock Events; I

effective June 14, 1991 requires licensees to submit the projected values of RT for the reactor vessel with the next update of the preEIbre/ temperature limits.

This l

l U.S.

Nuclear R gulatory Commiccion L

December 19,-

1991 Page 3 letter satisfies that requirement as the requested information was transmitted via Reference 6 and also contained herein as Attachment.i.

10 CFR 50, Appendix H requires submittal of the capsule withdrawal summary report within one year of capsule withdrawal.

This requirement was satisfied by the submittal of Reference 6.

This section of CFR also requires the submittal of revised Technical Specification figures for pressure / temperature limits, if changes are required.

This letter satisfies that requirement as the requested changes are contained herein as Attachment 2.

The Callaway Plant On-Site Review Committee and the Nuclear Safety Review Committee have reviewed and approved this amendment application.

Attachmento 1 through 5 provide the RT values, Technical Specification changes, S$I$ty Evaluation, Significant Hazards Evaluation, and Environmental Consideration respectively.

It has been determined that this amendment application does not involve any unreviewed safety questions as defined in 10 CFR 50.59, nor a significant hazard consideration as determined per 10 CFR 50.92.

If you have any questions concerning this amendment application, please contact us.

Very truly yours, (fbthe

\\]

Donald F.

Schnell WEK/plh Attachments:

1.

Revised RT Values (10 CFR 50.61 Submittal) PTS 2.

Proposed Technical Specification Changes 3.

Safety Evaluation 4.

Significant Hazard Evaluation 5.

Environmental Consideration i

STATE OF MISSOURI )

)

SS

. CITY OF ST. LOUIS )

Donald F. Schnell, of lawful age, being first duly sworn upon oath says that he is Senior Vice President-Nuclear and an officer of Union Electric Company; that he has read the foregoing document and knows the content thereof;-that he has executed the same for and on behalf of said company with. full power and authority to do so; and that the facts therein stated are true and correct to the best of his knowledge, information and b6ief,

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By r/511& '.' /

/m Donald F. '$chnell

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Senior Vice President Nuclear SUBSCRIBED and sworn to before me this

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- day of A/6tel/v 1991.

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,i T' NOTARf FUSLIC. SIATE Of MinCua M/ CCM.'. TBS 10N UFIRES APGlL 22, tra ST. LOUIS COUNTY l

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T. A;LBaxter,LEsq.:

Shaw, Pittman,-Potts & Trowbridge

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Washington,-D'.C.'20037 Dr.

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O. Cermak

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18225-A Flower Eill Way Gaithersburg, MD 20879-5334 R. C. Knop Chief, Reactor Project Branch _1 U.S. Nuclear Regulatory Commission Region III 799' Roosevelt Road Glen'Ellyn, Illinois 60137 Bruce Bartlett

' Callaway-Resident Office

- U;S.! Nuclear ~ Regulatory Commission RR#1-Steedman, Missouri 65077' J. R. - Hall-1-(2 )

Office of Nuclear Reactor Regulation-U.S._ Nuclear; Regulatory Commission 1? White? Flint,-North, Mail Stop 13E21

- 11555)Rockville Pike Rockville,,1MD 20852 Manager, Electric Department

- Missouri Public Service Commission P.O. Box 360 Jefferson City, MO 65102

.Ron Kucera Department of Natural Resources P.O. Box 176 Jefferson City,sMO 65102 l

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ULNRC-2537 CALLAWAY PIANT RESPONSB TO 10 CFR 50.61 "FRACIURE TOUGHNESS REQUIREMENTS FOR PROTECI' ION AGAINST PRESSURIZED THERMAL SHOCK EVENTS' 1.

Introduction The purpose of this report is to determine the reference temperatures for pressurized thermal shock (RT values for the Callaway Plant reactor vessel to-addrebbg)he t

Pressurized Thermal Shock (PTS) Rule.

(The expiration date of the Callaway Operating License is October 18, 2024.)

The PTS Rule requires that the PTS submittal be updated whenever there are changes in core loadings, surveillance measurements or other information that indicates a significant change in projected values.

In 1985 Nuclear Regulatory Commission (NRC) issued a formal ruling on pressurized ther:..al shock.

It established screening criterion on pressurized water reactor (PWR) vesselembrittlementasmeasuredby(penil-ductility g

reference temperature, termed RT RT screening values were set for beltline axi$$ welds, fbEhingsand plates and for beltline circumferential weld seams for

-end-of-life plant operation.

The screening criteria were determined using conservative fracture mechanics analysis techniques.

All PWR vessels in the United States have been required to evaluate vessel embrittlement in accordance with the criteria through end-of-life.

The Nuclear Regulatory

-Commission has amended its regulations for light water nuclear power plants to change the procedure for calculating radiation embrittlement.

The revised PTS Rule was published intheFederalRyg{ ster,May15, 1991 with an effective date l

of June 14, 1991 This amendment makes the procedure for calculating RT values consistent yjyli the methods given in Regulatory bdkde 1.99, Revision 2 2.

Core Loading Patterns Predictions of fluence during Callaway Cycle 1 were based on design values of 3411 MWt and an out-in loading pattern which has inherently high leakage.

The Cycle 2 fluence calculation was based upon best estimate calculations assuming 3411 MWt and a low leakage pattern j

(checkerboard pattern).

The low leakage loading pattern was E

employed at Callaway for Cycle 2 onward.

Page 1 of 4

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j!i*I ULNRC 2537

. Cycle 3_and 4_were transition cores to Vantage S fuel and-c design values : of 3565 MWt_ -(Uprated: conditions) ' were _used in-the fiuence calculations.

~

The reactor core-power distributionfused in the Callaway.

adjoint = calculations was_ taken f rom the fuel _ cycle design-reports for the_first fourioperating cycles.

In the evaluation of_the_-future exposure of-the raactor pressure vessel the--average ~ exposure rates derived-from cycles 1 through 4.were employed._ In completing these coverage exposure rates,-the calculated averages were-also scaled by the average measurement / calculation ratios observed from evaluations of_ dosimetry from Capsules Y and U.

Cycle 5 is a_ full Vantsge 5 core based upon design values at 3565 MWt.

^

Since we anticipate core design changes which will only reduce leakage, ro additional margin has-been added to account ~for future design changes.

3.

-Neutron Fluence Values The calculated fast _ neutron fluence (E>1MeV) at the inner surface of the Callaway reactor vessel was projected-using theresyfys'eftheCapsuleYradiationsurveillance program and are presented in Table 1.

_The peak fluence (25 degrees location) was used for all PTS calculations.

Using_the prescribed' PTS Rule methodology, RT values were generated for all beltline region materials oETEhe Callaway reactor vessel as a function of end-of-life (32 EFPY) and 48 EFPYffluence values.

The fluence data were generated b9gyd on the most recent surveillance capsule program results Table 2 provides a summary of the RT values for all beltline region materials for the enSIbf-life (32 EFPY) and 48-EFPY using-the PTS Rule.

The PTS Rule requires that each plant _ assess the RT values based on plant specific surveillance capsul $Tbata under certain conditions.

These conditions are:

Plant specific surveillance data has been deemed credible as defined in Regulatory Guide 1.99, Revision 2, and RT values change significantly.

(Changes to PTS RT values are considered significant if the va$db determined with RT equations (1) and (2),

or that using capsule daE$b or both, exceed the Page 2 of 4

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v;"1 ULWRC-2537 screening criteria prior to the expiration _of the-operating license, including any renewed term,-if applicable, for the' plant.)

I or.Callaway, the use--of plant--specific-surveillance-capsule F

data-is acceptable because there have-been two capsules removed from the reactor vessel, hence the data is credible

_per Regulatory Guide 1.99, Revision 2.

TABLE 1 NEUIRON EXEOSLR3 FROJECTIOG AT KEY LOCATICtB Of 'ITE CALIM&Y PRESSLEE VESSEL CLAD / BASE FEIAL IN1'ERFACE FOR 32 EFPY

'T O'

i 15' 25' 35*

45' 19 2

Fluence x 10 n/an 1,50 2.16 2.39 2.02 2.38 (E > 1 MeV)

TABLE 2 CALIERY RFAT.R VESSEL BELTLINE RH3 ION 1%TERIAL PROPERTIES CPU NI I-R'INDT Screening RT. _ M ca _

.hhterial Descriptian

(%)

(%)

(*F)

Criteria 32 EF W 48 EFPY Internudiate Shell, R2707-1 0.04 0.57 40 270 106 109 Internediate Shell, R2707-2 0.05 0.59 10 270 82 85 Internediate Shell, R2707 0.06 0.61

-10 f.70 70 73 Lower Shell, R2708-1 0.07 0.59 50 270 138 143 (116)

(119)

Lower Shell, R2708-2 0.05 0.57 10

-270 82 85 lower Shell, R2708-3 0.07 0.59 20 270 108 113 Internediate and Lower Shell-0.04 0.06

-60 300-33-

-36 '

icngitudinal Welds, G2.03 Circumferential Weld 0.06 0.07

-60 300 30 45 (65)

(71)

( )= indicates nunters were calculated using surveillance capsule data.

i-Page 3 of 4

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ULNRC-2 5 3 7 -

4.:

' Conclusions "As shownLin Table-2, all the RT

NRCscreeningvaluesforPTSus[$hvaluesremainbelowthe

^

the projected ~ fluence values - for -both theJ end-of-lif e -- (32 ~ EPPY) - and 48E EFPY. - The highest RT values.for the upper bound fluence case at 32 EFPY-and 4$T$FPY for the lower shellu plate,--(R2708-1) are 138: degrees F and.143_ degrees F,.respectively.

Neither of the two' limiting materials in1the beltline 1 region of the

.Callaway_ reactor vessel, the lower shell plates, R2708-1 andz R2708-3, are; expected-to exceed the' screening l criteria-based

-on'the current fluence projections.

5..

References

[1] 110CFR Part 50, " Analysis of Potential Pressurized

' Thermal Shock Eve..ts," July 23, 1985.

[2) _ 10CFR Part 50, " Fracture Soughness Requirements for Protection'Against Pressurized Thermal Shock Events',

June 14, 1991.

-(PTS Rule)

[3]

Regulatory Guide _1.99, Revision 2,

" Radiation Embrittlement-of Reactor Vessel Materials, "U.S.

Nuclear Regulatory Commission, May 1988.

[4]

WCAP-12946, " Analysis of Capsule Y from the Union 21ectric Company Callaway Unit 1, Reactor Vessel Radiation Surveillance Program,"

E.~Terek, et al.,

~ April 1991.

[5]

WCAP-12498, " Evaluation of-Pressurized Thermal Shock for'Callaway," J. M. Chicots, et al., May 1991.

'[6]

ULNRC-1244, dated January 21, 1986.

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