ML20154F840

Amend 208 to License DPR-50,revising TS 3.1.2 of TMI-1 TS to Incorporate New P/T Limits for Rv Pressurization Heatup, Cooldown & Inservice Leak & Hydrostatic Test,To Be Effective for Period of 17.7 EFPYs

ML20154F840
Person / Time
Site:	Crane
Issue date:	10/05/1998
From:	Thomas C NRC (Affiliation Not Assigned)
To:
Shared Package
ML20154F844	List: ML20154F840
References
NUDOCS 9810090395
Download: ML20154F840 (8)

Text

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1. 1 UNITED STATES j

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 3066H001 Se...../

METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY j

PENNSYLVANIA ELECTRIC COMPANY l

GPU NUCLEAR. INC.

DOCKET NO. 50-289 THREE MILE ISLAND NUCLEAR STATION. UNIT NO.1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No208 License No. DPR-50 1.

The Nuclear Regulatory Commission (the Commission or NRC) has found that:

A.

The application for amendment by GPU Nuclear, Inc., et al. (the licensee) dated, March 23,1998, as supplemented June 30,1998, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, the provisions of the i

Act, and the rules and regulations of the Commission; C.

There is reasonable assurance: (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

l 9810090395 981005 l

PDR ADOCK 05000289 l

P PDR l

1 2-2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.c.(2) of Facility i

Operating Ucense No. DPR-50 is hereby amended to read as follows:

(2)

Technical Soecifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 208, are hereby incorporated in the license. GPU Nuclear, Inc., shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance, to be implemented within 60 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION 1

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Cecil O. Thomas, Director l

Project Directorate 1-3 Division of Reactor Projects - 1/II l

Office of Nuclear Reactor Regulation l

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Attachment:

Changes to the Technical j

Specifications i

Date of issuance: October 5, 1998 I

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ATTACHMENT TO LICENSE AMENDMENT NO.208 l

FAClLITY OPERATING LICENSE NO. DPR-50 DOCKET NO. 50-289 Replace the following pages of the Appendix A, Technical Specifications, with the attached pages. The revised pages are identified by Amendment number and contain verticallines indicating the areas of change.

Remove Insert 3-3 3-3 3-4 3-4 3-5 3-5 3-Sa 3-Sa 3-5b 3-5b l

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3.1.2 PRESSURIZATION HEATUP AND COOLDOWN LIMITATIONS i

Aoolicability Applies to pressurization, heatup and cooldown of the reactor coolant system.

Obiectives l

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To assure that temperature and pressure changes in the reactor coolant system do not cause cyclic loads in excess of design for reactor coolant system components.

To assure that reactor vessel integrity by maintaining the stress intensity as a result of operational plant l

heatup and cooldown conditions and insenice leak and hydro test conditions below values which may l

result in non-ductile failure.

Specification 3.1.2.1 For operations until 17.7 effective full power years, the reactor coolant pressure and the l

system heatup and cooldown rates (with the exception of the pressurizer) shall be l

limited in accordance with Figure 3.1-1 and Figure 3.1-2 and are as follows:

l Heatuo/Cooldown l

Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-1. Heatup and cooldown rates shall not exceed those showm on Figure 3.1-1.

I Insenice Leak and Hydrostatic Testina Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-2. Heatup and cooldown rates shall not exceed those shown on Figure 3.1-2.

3.1.2.2 The secondary side of the steam generator shall not be pressurized above 200 psig if the temperature of the steam generator shell is below 100*F.

3.1.2.3 The pressurizer heatup and cooldown rates shall not exceed 100 F in any one hour.

j The spray shall not be used if the temperature difference between the pressurizer and the spray fluid is greater than 430'F.

3.1.2.4 Prior to exceeding 17 7 effective full power years of operation, Figures 3.1-1 and 3.1-2 l

shall be updated for the next service period in accordance with 10 CFR 50, Appendix G, Section V.B. The highest predicted adjusted reference temperature of all the beltline materials shall be used to determine the adjusted reference temperature at the end of the service period. The basis for this prediction shall be submitte<! for NRC staff review in accordance with Specification 3.1.2.5.

3.1.2.5 The updated proposed technical specifications referred to in 3.1.2.4 shall be submitted for NRC review at least 90 days prior to the end of the service period. Appropriate additional NRC review time shall be allowed for proposed technical specifications submitted in accordance with 10 CFR 50, Appendix G, Section V.C.

3-3 Amendment No. 29,134, iWr,208

i Bases All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes (Reference 1). Dese cyclic loads are introduced by unit load transients, reactor trips,had unit heatup and cooldown operations. The number of thermal and loading cycles used for design purposes are shown in Table 4.1-1 of the UFSAR. He maximum unit heatup and cooldown rates satisfy stress limits for cyclic operation (Reference 2). He 200 psig pressure limit for the secondary side of the steam generator at a temperature less than 100'F satisfies stress levels for temperatures below the Nil Ductility Transition Temperature (NDTT).

The heatup and cooldown rate limits in this specification are based on linear heatup and cooldown ramp rates which by analysis have been extended to accommodate 15'F step changes at any time with the appropriate soak (hold) times. Also, an additional temperature step change has been included in

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the analysis with no additional soak time to accommodate decay heat initiation at approximately 240'F indicated RCS temperature.

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He unirradiated reference nil ductility temperature (RTurr) for the surveillance region materials were j

determined in accordance with 10 CFR 50, Appendixes G and H. For other beltline region ma:erials j

and other reactor coolant pressure boundary materials, the unirradiated impact properties were estimated using he methods described in BAW-10046A, Rev. 2.

1 is a result of fast neutron irradiation in the beltline region of the core, there will be an mereate m the l

RTurr with accumulated nuclear operations. He adjusted reference temperatures have been calculated as described in Reference No. 6.

The predicted RTun was calculated using the respective predicted neutror fluence at 17.7 effective i

full power years of operation and the procedures defined in Regulatory Guide 1.99, Rev. 2, Section i

C.I.1 for the plate metals and for the limiting weld metals (SA-1526 & WF-25).

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Analyses of the activation detectors in the TMI l surveillance capsules have provided estimates of j

reactor vessel wall fast neutron fluxes for cycles I through 4. Extrapolation of reactor vessel fluxes (average of cycles 8 and 9), and corresponding fluence accumulations, based on predicted fuel cycle 1

design conditions through 17.7 effective full power years of operation are described in References 5 and 6.

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3-4 Aniendment NocEW_4,157, !?6 208

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' Based on the predicted RTm after 17.7 effective full power years cf operation, the.

pressure / temperature limits of Figure 3.1-1 and 3.1-2 have been established by FTI calculation, reference No. 7,in accordance with the requirements of 10 CFR 50, Appendix G.- Also, see Reference 4. De methods and criteria employed to establish the operating pressure and temperature limits are as described in BAW-10046A, Rev. 2. De protection against nonductile failure is provided by maintammg the coolant pressure below the upper limits of these pressure temperature limit curves.

De pressure limit lines on Figure 3.1-1 and 3.1-2 have been established considering the following:

A 25 psi errorin measured pressure.

a.

b. A 12*F error in measured temperature.

System pressure is measured in either loop.

c.

d. Maximum differential pressure between the point of system pressure measurement and the -

limiting reactor vessel region for the allowable operating pump combinations.

1 The spray temperature difference restriction, based on a stress analysis of spray line nozzle is imposed to maintain the thermal stresses at the pressurizer spray line nozzle below the design limit.

Temperature requirements for the steam generator correspond with the measured NDTT for the shell.

REFERENCES (1)

UFSAR, Section 4.I'.2.4 " Cyclic Loads" (2)

ASME Boiler and Pressure Code,Section III, N-415 (3)

B AW-1901, Analysis of Capsule TMI-lC, GPU Nuclear, Three Mile bland Nuclear Station -

Unit 1, Reactor Vessel MaterLis Surveillance Program (4)

BAW-1901, Supplement 1, Analysis of Capsule TMI-lC, GPU Nuclear, nree Mile Island Nuclear Station - Unit 1, Reactor Vessel Materials Surveillance Program, Supplement 1 Presrure -Temperature Limits.

(5)

BAW-2108, Rev.1, B&WOG Materials Committee Report "Fluen'ce Tracking System" (6)

GPU Nuclear calculation No. C-1101-221-E520-013 Rev. O, "TMI-l Reactor Vessel Welds Fluence, RTrrs and RTm per R.G.1.99 R-2, Pos. No.1 (7)

FTI calculation No. 32-5001065 01,'TMI-l P/r Limits," March 1998.

3-5 Amendment No. 29,134,157,175 208

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g Figure 3.1.1 Reactor Coolant System Combined Heatup/Cooldown Limituns

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[ Applicable through 17.7 EFPY]

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ASSil*Go RTamw.*F K

2500 -

BELTLINE 1/4T - 214 p

NOTES:

d BELTLINE 3/4T - 160 g-g 1 -Tems.^dma; CLOSURE HEAD-60

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All temperaturcs'are the indicated values in the operating REGION

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RC Pump (s) RCS Cold Leg.

OUTLET NOZZLE - 60 EL 2000 - Except:

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E When the DHR System is operating without any RC Pumps S

O-operating, then the indicated DHR retum temperature to the S

Reactor Vessel shall be used.

Y 2 - Heatuo:

J Point Jemo.

Press.

E E 3 21500

~ 50FMr or 1SF/18 Min. Steps A

60 250 oeo B

90 335

. 3 - Cooldown:

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g C

120 335 O

tr, T>270,100FMr or 15F/9 Min.

D 200 370 i

y 270>T>250,50FMr or 15F/18 Min.

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E 245 485 g

1000 - 250>T> Approx. 240 [DHR init.],10FMr or 10F/60 Min.

F 265 575 f.,

- Temp. Steps Due to DHR Initiation H

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g 273

,3, Approx. 235 [After DHR Init.] >T>70, 30FMr or 15F/30 Min.

H 300 780 l

3 4 - RC Pump Combinations:

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500 -. T > 195 F: ALL ; T < 195 F : 1-1,1-0,0-1

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J 375 1675 V

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405 2350 l'

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467 2500 k B\\

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50 100 150 200 250 300 350 400 450 500 indicated Reactor Coolant inlet Temperature, [*F]

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Figure 3.1.2 Reactor Coolant inservice Leak and Hydrostatic Test y;

[ Applicable through 17.7 EFPY]

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RC Pump (s) RCS Cold Leg.

Point Temp.

Press.

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A 65 380 0'

e When the DHR System is operating without any RC Pumps B

95 470 n.

2000 - operating, then the indicated DHR retum temperature to the C

120 470 E

Reactor Vessel shall be used.

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197 490 g

a 2 - Heatus:

E 217 490 y

50FMr or 15F/18 Min. Steps F

222 600 I

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3 - Cooldown:

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265 812 h

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T>270,100FMr or 15F/9 Min.

H 275

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270>T>250,50FMr or 15F/18 Min.

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to 250>T> Approx. 240 [DHR init.],10FMr or 10F/60 Min.

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340 1578 O

o Temp. Steps Due to DHR Initiation K

375 2278 j

1000 - Approx. 235 [After DHR Init] >T>70, 30FMr or 15F/30 Min.

L 383 2500 l

fE 4 -RC Pums Combinations:

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J T > 195 F: ALL ; T < 195 F : 1-1,1-0,0.1 ASSUMED RTann.*F W

G BELTLINE 1/4T - 214 y

500 g

p BELTLINE 3/4T - 160 5

I CLOSURE HEAD - 60 REGION A

B LC D d E

OM NOME - 60 0

0 50 100 150 200 250 300 350' 400 450 500 Indicated Reactor Coolant inlet Temperature, [*F]