ML20210K969

Application to Amend License DPR-50,consisting of Tech Spec Change Request 168,changing Tech Specs to Allow Nominal Increase in Power While Main Steam Safety Valve & Other Func Tional Tests Being Performed

ML20210K969
Person / Time
Site:	Three Mile Island
Issue date:	02/05/1987
From:	Hukill H GENERAL PUBLIC UTILITIES CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20210K924	List: ML20210K920 ML20210K969 ML20210L036
References
NUDOCS 8702120323
Download: ML20210K969 (8)

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METROPOLITAN EDIS0N COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMPANY THREE MILE ISLAND NUCLEAR STATION, UNIT 1 A

s Operating License No. DPR-50 Docket No. 50-289 Technical Specification Change Request No.168 This Technical Specification Change Request is submitted in support of Licensee's request to change Appendix A to Operating License No. DPR-50 for Three Mile Island Nuclear Station, Unit 1. As a part of this request, proposed replacement pages for Appendix A are also included.

GPU NUCLEAR CORPORATION BY: a Vice President & Director, TMI-l l

l Sworn and Subscribed l to before me this 50 l day of M>linjaAd ,1987.

l l

e AJAM f*

f Notary Puylic l

$DLRON P. BROWN. NOTARY PUBLIC i

L!IDDtfiOW:1 BORD DAUPHIN COUNTY 11Y C0tittlS$lCN LIFIPCS JUNE 12,1989 l

Member, Penns>ivania Associat.cn of Notaries l

I 8702120323 870205 hDR ADOCK 05000289 PDR

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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION IN THE MATTER OF DOCKET NO. 50-289 LICENSE NO. DPR-50 GPU NUCLEAR CORPORATION This is to certify that a copy of Technical Specification Change Request No.168 to 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 of Emironmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follows:

Mr. Jay H. Kopp, Chairman Mr. Frederick S. Rice, 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 Mr. Thomas Gerusky, Director PA. Dept. of Ewironmental Resources Bureau of Radiation Protection P.O. Box 2063 Harrisburg, PA 17120 GPU NUCLEAR CORPORATION BY:

Vice Pretfdent & Director, TMI-l

DATE

February 5, 1987

4 I. Technical Sp'ecification Change Request (TSCR) No.168 GPUN requests that the following pages be inserted into the existing Technical Specifications (assuming that TSCR No.162 has been approved):

Revised pages 3-26, 3-26a, 3-26b, and 3-26c These pages are attached to this Change Request.

II. Reason for Change Current Technical Specifications state that when the Reactor is between 250*F and Hot Shutdown, two (2) Main Steam Safety Valves (MSSVs) per Steam Generator shall be OPERABLE. When the Reactor is above Hot Shutdown, all eighteen (18) MSSVs shall be OPERABLE.

Current Technical Specifications do not provide for efficient inI place testing of MSSYs following Valve Maintenance during Cold Shutdown or ref ueling.

Technical Specifications have been modified to allow for a. nominal increase in power of no more than 5% while the MSSVs and other

' functional testing are being performed. ~

III. Safety Evaluation Justifying Change Plant specific analysis shows that two MSSVs per Steam Generator are more than sufficient to relieve Reactor Coolant Pump heat and stored

energy when the reactor is below 5% full power operation and had been subcritical by 1% A K/K for at least one hour since power operation above 5%. The calculations made the following assumptions

1. Steam conditions are dry saturated.

l 2. Atmospheric dump valves (MS-Y-4A/B) remain closed.

3. The plant has been subcritical for at least one hour since power operation above 5% power.

4. The RPS trip setpoint is no higher than 5% full power.

Based on these assumptions, the limiting criteria for MSSV operability is their ability to relieve 5% full power. In all cases, the decay heat generated following a reactor trip under these restrictions is bounded by the 5% full power condition. This Tech. Spec. change will require two MSSVs on each Steam Generator to be conservative which is more than sufficient for this 5% full power condition.

A review of the FSAR accidents has not identified any additional requirements for MSSV operability beyond the relief capacity criteria.

The FSAR accidents looked at are as follows: ,

1. Uncompensated Operating Reactivity Changes -- Not affected.

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III. Safety Evaluation Justifying Change (Cont'd.)

2. Startup Accident -- The startup accident would be terminated by the 5% high flux trip for the range of analyzed reactivity additions.

In all cases, the thermal reactor power lags neutron power. The maximum thermal power remains less than 5% FP.

3. Rod Withdrawal Accident at Rated Power Operation -- Not applicable when 5% FP trip setpoint is in place.

4. Moderator Dilution Accident -- Would terminate on high flux at

< 5% FP.

5. Cold Water Accident -- Would be terminated by 5% FP trip. Also it is unlikely that less than 4 pumps would be operating during testing.

6. Loss of Coolant Flow -- More than adequate flow is provided by 1 pump per loop. With less than 1 pump per loop, Reactor trip will be initiated by pump / power monitors.

7. Stuck-Out, Stuck-In, or Dropped Control Rod Accident -- Fuel conditions at 5% FP are not affected by MSSV operability. No change on analysis.

8. Loss of Electric Power -- Requires that MSSV relieve core thermal 1

power. Sufficient MSSVs need to be operable to relieve core energy at 5% full power trip setpoint. Accident not otherwise affected by MSSV.

i 9. Steam Line Break -- Overcooling accident which does not rely on MSSV operability.

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10. Steam Generator Tube Rupture -- Dose is not affected by MSSV l provided sufficient valves to relieve core energy are available.

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11. Fuel Handling Accident -- Not affected.

12. Rod Ejection Accident -- Neutron power significantly leads thermal power, actual thermal power will therefore be less than 5% FP.

l 13. Large Break LOCA -- No requirements for OTSG energy removal.

14. Small Break LOCA -- MSSV need to relieve core energy prior to trip and decay heat thereafter. Core energy limited to 5% FP by trip function. Decay heat value is bounded by core energy case.

l 15. Maximum Hypothetical Accident -- No requirement for OTSG energy removal. Accident results not affected by MSSV.

16. Waste Gas Tank Rupture -- Not affected.

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[

III. Safety Evaluation Justifying Change (Cont'd.)

17. Loss of Feedwater Accident -- MSSV need capacity to remove core energy which is limited by overpower trip setpoint of 5% FP and subsequent decay heat removal. Decay heat generation is bounded by 5% FP.

In summary, there is justification to allow startup and low power testing prior to the completion of the MSSV testing. See Attachment I for calculations.

IV. No Significant Hazards Considerations GPUN has determined that this Technical Specification Change Request poses no significant hazards as defined by the NRC in 10 CFR 50.92.

A. Operation of the facility in accordance with the proposed amendment would not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

The Technical Specification changes are to allow for testing of the MSSVs during Power Operation below 5% power with RPS trip setpoints at less than 5% power. The results of this change will not impact the events analyzed in Chapter 14 of the TMI-l FSAR and the TMI-l Reload Reports will remain bounding.

Therefore, the Technical Specification change for MSSV testing does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

B. Operation of the facility in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated. Analyses have been performed to ensure that technical specification limits are not exceeded. Results show that MSSV operability requirements are conservatively bounded by the existing analysis in all cases.

Therefore, it is concluded that the Technical Specification change i for MSSV operability does not create the possibility of a new or i different kind of accident from any accident previously evaluated. i l

e C. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety.

All safety criteria as described in the Technical Specification bases are preserved by the additional MSSV operability i nformation. Therefore, it is concluded that the Technical Specification change for MSSV operability does not involve a significant reduction in a margin of safety.

V. Implementation It is requested that the amendment authorizing this change become effective immediately upon issuance. l VI. Amendment Fee (10 CFR 170.21)

Pursuant to the provisions of 10 CFR 170.21, attached is a check for

$150.00. -

ATTACMENT I Page 1 of 3 MAIN STEAM SAFETY VALVE CALCULATIONS DURING LOW POWER PHYSICS TESTING I. Assumptions

1. Steam conditions are dry, saturated.

2. Atmospheric Dump Valves (MS-V-4A/B) remain closed.

3. The plant is below 5% full power operation but had been subcritical 1% AK/K for at least one hour since power operation above 5% full power.

4. Overpower trip setpoint is less than 5% full power.

5. Reactor core heat at 5% power is 126.75 MWt.

6. Reactor coolant pump heat is conservatively 5 MWt for each of the four purr.ps, for a total of 20 MWt of pump heat.

7. Multiply ANS decay heat value by 1.2 to conservatively bound all projected fuel cycle schemes for THI-1.

8. The Nuclear Instrumentation (NIs) is calibrated such that the NI's reflect the total core heat, both nuclear heat and fission product decay heat per standard calibration methods.

9. Feedwater heating using Auxiliary Boilers is being utilized.

Resultant feedwater temperature at inlet to OTSG is 200*F.

II. Calculations The total heat load on the OTSG when critical is made up of core thermal heat, feedwater thermal heat and reactor coolant pump heat. When subcritical, the total heat load is the residual decay heat and feedwater thermal input. The decay heat value was calculated by assuming that the plant had tripped at 100% power and had been subcritical for one hour. Decay heat following subsequent operation at 5% full power is bounded by the 5% full power thermal heat load condition.

When critical at less than 5% full power, the maximum heat load is:

4 Total = 4 5% full power + 9 RCP heat + 9 feedwater heat i

4 Total = 126.75 MWt + 20 MWt + 25 MWt = 171.75 MWt Conservatively assume 175 MWt.

4 Total = (175 MWt) (1000 Kw) (3412.12 Btu ) = 597.1 x 106 Btu Mw Kw-br hr l

l 9 Total = 597,100,000 BTU hr

ATTACHMENT I Page 2 of 3 II. Calculations (Cont'd)

When subcritical following a trip from less than 5% full power (See Assumption 3), the heat load on the OTSG's is less than that for the critical case given above. The critical condition at less than 5% full power is the limiting case. Decay heat based on 1.2 time ANSI 5.1-1979 one hour af ter a trip from 100% full power is 40.8 MWt. Decay heat after subsequent 5% full power operation, decay heat remains below 5%

full power in all cases.

MAIN STEAM SAFETY VALVES SPECIFICATIONS Capacity (1bs/hr)

Set Point (psig) at 3% Accumulation MS-V-17A/D 1050 792,610 MS-Y-20A/D 1050 792,610 MS-Y-18A/D 1060 799,990 MS-Y-19 A/D 1080 814,955 MS-V-208,C 1092.5 824,265 MS-V-21A,B 1040 194,900 The comparison of the heat load to the MSSV capacities :nust be done using the smallest MSSV, MS-Y-21 A/B. The setpoint of this valve is at 1040 psig. It is fully open at 3% accumulation.

P1 = (1040 psig + 14.7 psi) (1.03) = 1086.3 psia hg = 1189.6 BTU /lb QValve = (rn) (hg)

QValve = (194,900 lbs/hr) (1,189.6 BTU /lb)

QValve = 231,853,040 BTU /hr ~ 231.9 x 106 BTU /hr The next smallest MSSVs are MS-V-17 A/D and MS-y-20 A/D. The setpoint of these valves are at 1050 psig. It is fully open at 3% accumulation.

P1 = (1050 psig + 14.7 psi) (1.03) = 1096.64 psia hg = 1189.2 BTU /lb hValve = (r'n) (hg) bValve = (792,610 lbs/hr) (1189.2 BTU /lb)

QValve = 942,570,000 BTU /hr - 942.6 x 106 BTU /hr

ATTACMENT I l Page 3 of 3 l l

l II. Calculations (Cont'd) l The heat load may now be compared to the heat load for the valves.

hTotal = 597.1 x 106 BTU /hr

, hValve =

231.9 x 106 BTU /hr (for each MS-V-21 A/B) bValve =

942.6 x 106 BTU /hr (for each MS-V-17 A/D and MS-V-20 A/D)

The total minimum valve relief capacity for each steam generator is 1174.5 x 106 BTU /hr or nearly twice the total reactor coolant system heat using the two smallest valves per OTSG. With two MSSVs per generator, the relief capacity is almost four times the potential heat load.

III. Conclusion Four MSSVs are required to relieve the 5% full power heat with two MSSVs on each steam generator.

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