ML20217E531

Proposed Tech Specs Pages for Section 3.1.2 to Incorporate New Pressure Limits for Reactor Vessel IAW 10CFR50,App G for Period of Applicability Through 17.7 EFPY

ML20217E531
Person / Time
Site:	Three Mile Island
Issue date:	03/23/1998
From:	GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20217E519	List: ML20217E513 ML20217E531
References
NUDOCS 9803310068
Download: ML20217E531 (6)

Text

. ' 3.1,2 PRESSURIZATION HEATUP AND COOLDOWN LIMITATIONS APRliGahility

- Applies to pressurization, heatup and emidown of the reactor coolant system.

- Obiectives To assure 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 heatup and cooldown conditions and inservice leak and hydro test conditions, below values which may result in non-ductile failure.

Sp. sci 5sation 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 limited in accordance with Figure 3.1-1 arai Figure 3.1-2 r.nd are as follows:

Heatup/Cooldown 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 shown on Figure 3.1-1.

Inigtvicq_ Leak and Hy3kostaliglqging

. 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 De 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 presssurizer heatup and cooldown rates shall not exceed 100 F in any one hour. 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 submitted 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.

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- Amendment No. 297 434<4%

9803310068 980323 PDR ADOCK 05000289 '

P PDR ,

. - Bases All reactor coolant system components are designed to withstand the effects of cyclic loads due to system temperature and pressure changes (Reference 1). These cyclic loads are introduced by unit load transients, reactor trips, and 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. The maximum unit heatup and cooldown rates satisfy stress limits for cyclic operation (Reference 2). The 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 (NDIT).

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 the analysis with no additional soak time to accommodate decay heat initiation at approximately 240 F -

indicated RCS temperature.

The unirradiated reference nil ductility temperature (RTmtr) for the surveillance region materials were -

determined in accordance with 10 CFR 50, Appendixes G and H. For other beltline region materials and other reactor coolant pressure boundary materials, the unirradiated impact properties were estimated using he methods described in BAW-10046A, Rev. 2.

As a result of fast neutron irradiation in the beltline region of the core, there will be an increase in the RTurr with accumulated nuclear operations. The adjusted reference temperatures have been calculated as described in Reference No. 6.

The predicted RTurrr was calculated using the respective predicted neutron fluence at 17.7 effective l full power years of operation and the procedures defined in Regulatory Guide 1.99, Rev. 2, Section C.l.1 for the plate metals and for the limiting weld metals (SA-1526 & WF-25).

l Analyses of the activation detectors in the TMI-l surveillance capsules have provided estimates of q 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 j design conditions through 17.7 effective full power years of operation are described in l References 5 and 6. 4 l

_ 4' Amendment No. 29,13,157,176 i

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• Based on the predicted RTwara fter 17.7 effective full power years of operation, the pressure / temperature limits of Figure 3.1-1 and 3.1-2 have been established by FTl calculation, reference No. 7, in accordance with the requirements t 10 CFR 50, Appendix G. Also, see Reference 4. Tbe methods and criteria employed to establish the operating pressure and temperature limits are as descrFied in BAW-10046A, Rev. 2. The protection against nonductile failure is provided by maintaining the coolant pressure below the upper limits of these pressure temperature limit curves.

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

a. A 25 psi error in measured pressure.

b. A 12 F error in measured temperature.

c. System pressure is measured in either loop.

d. Maximum differential pressure between the point of system pressure measurement and the limiting reactor vessel region for the allowable operating pump combinations.

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 NDTf for the shell.

BliEERENGS (1) UFSAR, Section 4.1.2.4 "C3clic Loads" (2) ASME Boiler and Pressure Code, Section 111, N-415 (3) BAW-1901, Analysis of Capsule TMI-lC, GPU Nuclear, Three Mile Island Nuclear Station -

Unit 1, Reactor Vessel Materials Surveillance Program (4) BAW-1901, Supplement 1 Analysis of Capsule TMI-lC, GPU Nuclear, Three Mile Island Nuclear Station - Unit 1, Reactor Vessel Materials Surveillance Program, Supplement i Pressure - Temperature Limits. i I

(5) BAW-2108, Rev.1, B&WOG Materials Committee Report " Fluence Tracking System" l (6) GPU Nuclear calculation No. C-1101-221-E520-013 Rev. O "TMI-l Reactor Vessel Welds Fluence, RTrrs and RTsar per R.G. l.99 R-2, Pos. No.1 (7) FTl calculation No. 32-5001065-01,"TMI-l P/T Limits," March 1998.

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- UNITED STATES OF AMERICA-NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 GPU NUCLEAR INC. LICENSE NO. DPR-50 CERTIFICATE OF SERVICE This is to certify that a copy ofTechnical Specification Change Request No. 270 td Appendix A of the Operating License' for Three Mile Island Nuclear Station Unit 1, has, on the date given below, been filed with executives _of Londonderry Township, Dauphin County, Pennsylvania; - .

Dauphin County, Pennsylvania; and the Pennsylvania Department ofEnvironmental Resources,

. Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:

Mr. Darryl LeHew, Chairman Ms. Sally S. Klein', Chairman Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County R. D. #1, Geyers Church Road Dauphin County Courthouse Middletown, PA 17057 Harrisburg, PA 17120 Director, Bureau of Radiation Protection PA Dept. of Environmental Resources Rachael Carson State OfYice Building P.O. Box 8469 Harrisburg, PA 17105-8469 Att: Mr. Stan Maingi GPU NUCLEAR INC.

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BY: 'b Vice President and Director, TMI DATE: 3- a 3-93 4 4

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