ML20056E347

Amend 176 to License DPR-50,extending Effectiveness of TS Pressure Vs Temp Limits Until Plant Operation Has Reached 15.2 EFPY

ML20056E347
Person / Time
Site:	Crane
Issue date:	08/16/1993
From:	Stolz J Office of Nuclear Reactor Regulation
To:	Metropolitan Edison Co, Jersey Central Power & Light Co, Pennsylvania Electric Co, GPU Nuclear Corp
Shared Package
ML20056E348	List: ML20056E347 ML20056E349
References
DPR-50-A-176 NUDOCS 9308230217
Download: ML20056E347 (8)

Text

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UNITED STATES i

NUCLEAR REGULATORY COMMISSION Q.s,

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WASHINGT ON D.C. 2055%0001 METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY PENNSYLVANIA ELECTRIC COMPANY GPU NUCLEAR CORPORATION l

l DOCKET NO. 50-289 l

THREE MILE ISLAND NUCLEAR STATION. UNIT NO. 1 AMENDMENT TO FACILITY OPERATING LICENSE l

Amendment No.176 I

License No. DPR-50 1.

The Nuclear Regulatory Commission (the Comission) has found that:

A.

The application for amendment by GPU Nuclear Corporation, et al.

(the licensee) dated March 19, 1993, as supplemented on July 23, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission'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 Act, and the rules and regulations of the Comission; 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 Comission's regulations and all applicable requirements have been satisfied.

9308230217 930816 PDR ADOCK 05000289 P

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2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license i

amendment, and paragraph 2.c.(2) of Facility Operating License No.

DPR-50 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.176, are hereby incorporated in the license.

GPU Nuclear Corporation 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.

l FOR THE NUCLEAR REGULATORY COMMISSION

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Q YV dohnF.Stolz,Direcor l

7Jr ject Directorate I-4 Division of Reactor Projects - I/II i

Office of Nuclear Reactor Regulation

Attachment:

j Changes to the Technical Specifications Date of Issuance:

August 16, i993 j

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ATTACHMENT TO LICENSE AMENDMENT NO.176 FACILITY 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 vertical lines indicating the area of change.

Remove Insert 3-3 3-3 3-4 3-4 3-5 3-5 3-5a 3-5a 3-5b 3-5b 1

i l

3.1.2 PRESSURIZATION HEATUp AND C00LDOWN LIMITATIONS ApolicabilitY Applies to pressurization, heatup and cooloown of the reactor coolant system.

Obiective 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.

Specification 3.1.2.1 For operations until 15.2 effective full power years, the reactor coolant pressure and the system heatup and cooldown rates (with the exception of the pressurizer) shall be limited in accordance with Figure 3.1-1 and Figure 3.1-2 and are as follows:

Heatuo/Cooldown i

Allowable combinations of pressure and temperature shall be to the right of and below the limit line in Figure 3.1-1.

Heatup and j

cooldown rates shall not exceed those shown on Figure 3.1-1.

I Inservice 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.

i 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 l

l below 100'F.

3.1.2.3 The pressurizer 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 i

than 430*F.

3.1.2.4 Prior to exceeding 15.2 effective full power years of operation, Figures 3.1-1 and 3.1-2 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 propose <1 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 l

Amendment 29, 134,176 i

t 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 l

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 15'F step change has been included in the analysis with no additional soak time to accommodate decay heat initiation at approximately 252*F.

i The unirradiated reference nil ductility temperature (RT,) 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 the 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 RT with accumulated nuclear operations.

Theadjustedreferencetemperatures%7 ave been calculated as described in retarence No. 6.

The predicted RT was calculated using the respective predicted neutron fluence at 15.2 Nfective full pcwer years of operation and the procedures defined in Regulatory Guide 1.99, Rev. 2, Section C.1.1 for the plate metals and the provisions of Section C.2.1 for the weld metals.

The analysis of the reactor vessel material contained in the second Three Mile Island Nuclear Station Unit 1 surveillance capsule as well as B&WOG surveillance capsules confirmed that the current technique, as described in reference No. 6, for predicting the change in impact properties due to irradiation are conservative.

Analyses of the activation detectors in the THI-1 surveillance capsules have provided estimates of reactor vessel wall fast neutron fluxes for cycles 1 through 4.

Extrapolation of reactor vessel fluxes (average of cycles 8 &

9), and corresponding fluence accumulations, based on predicted future fuel cycle design conditions during 15.2 effective full power years of operation are described in References 5 and 6.

j 3-4 Amendment 29, 134, 157,176

Based on the predicted RT, after 15.2 effective full power years of i

operation, the pressure / temperature limits of Figure 3.1-1 and 3.1-2 have i

been established in accordance with the requirements of 10 CFR 50, Appendix G.

Also, see Reference 4.

The methods and criteria employed to establish the operating pressure and temperature limits are as described 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.

t The pressure limit lines on Figures 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.

l

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 NDTT for the shell.

REFERENCES (1) UFSAR, Section 4.1.2.4

" Cyclic Lottjs" (2) ASME Boiler and Pressure Code,Section III, N-415 (3) BAW-1901, Analysis of Capsule TMI-1C, 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 1 Pressure - Temperature Limits (5) BAW-2108, Rev.1, B&WOG Materials Committee Report " Fluence Tracking System" (6) GPUN Topical Report 095, Rev. O, " Summary Evaluation of the TMI-l Reactor Vessel Embrittlement for Operating Pressure / Temperature Limits" 3-5 Amendment 29, 134, 157,176

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