ML20151T047

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Application for Amend to License DPR-50,consisting of Tech Spec Change Requets 184 Re Increase in Rated Power from 2,535 Mwt to 2,568 Mwt
ML20151T047
Person / Time
Site: Three Mile Island Constellation icon.png
Issue date: 04/18/1988
From: Hukill H
GENERAL PUBLIC UTILITIES CORP.
To:
Shared Package
ML20151T037 List:
References
NUDOCS 8804280566
Download: ML20151T047 (10)
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METROPOLITAN EDISON COMPANY JERSEY CENTRAL POWER & LIGHT COMPANY AND PENNSYLVANIA ELECTRIC COMDANY THREE MILE ISLAND NUCLEAR STATI0tl, UNIT 1 3 Operating License No. DPR-50 Docket No. 50-289

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Technical Specification Change Request No.184 l

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 i pages for Appendix A are also included.

GPU NUCLEAR CORPORATION I

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BY: A Vice President & Director, TMI-l Sworn and Subscribed to before me this /f d day of d & i l , 1988.

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UNITED STATES OF AMERICA NUCLEAR REGULATORY C0FNISSION IN THE MATTER OF DOCKET NO. 50-289 6PU NUCLEAR CORPORATION LICENSE NO. DPR-50 CERTIFICATE OF SERVICE This is to certify that a copy of Technical Specification Change Request No.184 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; end the Pennsylvania Department of Environmental Resources, Bureau of Radiation Protection, by deposit in the United States mail, addressed as follo'ws:

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Mr. Kenneth E. Witmer, Chairman Ms. Sally S. Klein, Chairman Board of Supervisors of Board of County Commissioners Londonderry Township of Dauphin County 25 Roslyn Road Dauphin County Courthouse Elizabethtown, PA 17022 Harrisburg, PA 17120 Mr. Thomas Gerusky, Director PA Dept. of Environmental Resources Bureau of Radiation Protection P.O. Box 2063 Harrisburg, PA 17120 GPU NUCLEAR CORPORATIO BY:

Vice Pres'Ident & Director, TMI-1 DATE: April 18, 1988 1

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l I. TECHNICAL SPECIFICATION CHANGE REQUEST (TSCR) NO. 184  :

GPUN requests that the following changed replacement page be inserted into the existing Technical Specifications:

Revised page: 1 -1 Also, the following changed replacement page should be inserted into i' existing Facility Operating License No. DPR-50:

Revised page: 5 These pages are attached to this change request.

II. REASON FOR CHANGE  :

This change is requested to allow an increase in the rated power for TMI-1 from 2535 MWt to 2568 MWt.

III. SAFETY EVALUATION JUSTIFYING CHANGE Cycle 7 has been designed to include a 1.3% rated core power upgrade to 2568 MWt from the 2535 MWt power level of previous cycles. TMI-1 Cycle 7 Reload Report (BAW-2015, March 1988) and Technical Specification Change Request No.182 (GPUN letter C311-88-2033) incorporate Technical ,

Specifications which bound operation of TMI-1 at a "ated power of 2568 MWt. All potentially affected plant systems and safety analyses were reviewed, as were relevant operational and environmental considerations, i to determine any adverse effects of the power increase. The evaluation identified no adverse nuclear safety effects.

A. System Review l

Plant systems listed in Table 1 were reviewed to determine if the components will perform acceptabl.- .

  • 2568 MWt. Each systein was  ;

evaluated considering the slightly pure demanding operating j conditions at the higher power, including system setpoints and  ;

response characteristics. Also, some TMI-1 systems were compared l against similar systems in B&W 177-Fuel Assembly plants presently operating at 2568 MWt or greater. The evaluations concluded that all safety related systems and components will perform within their !

design conditions at 2568 MWt. Non-safety related systems have i

also been evaluated at 2568 MWt. There will be no adverse impact  !

on nuclear safety associated with the operation of these systems.  !

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1 B. Transient and Safety Analyses  ;

The design basis safety analyses listed in Table 2 were reviewed to confirm their applicability for a rated power of 2568 MWt. For each event controlling analysis prameters were assessed for ,

changes due to the higher power. Many of the existing analyres were originally done at 2568 MWt and, thus, are unaffected by the change.

Certain FSAR events previously analyzed at 2535 MWt have been reevaluated for the Cycle 7 reload at 2568 MWt and determined to be acceptable (Reference TMI-1 Cycle 7 Reload Report, BAW-2015, March 1988). These include:

Startup Accident Rod Withdrawal Accident Moderator Dilution Accident Uncompensated Operating Reactivity Change Cold Water Accident Stuck-Out, Stuck-In or Dropped Control Rod Accident Rod Ejection Accident Waste Gas Tank Rupture Fuel Cask Drop Accident The remainder of events or responses not performed at 2568 MWt listed on Table 2 are discussed below.

1. HPI Flow Split / Cooling Capacity The B&W generic Small Break LOCA analysis was performed at a rated power of 2772 MWt and used an HPI flow split of 70% to the core and 30% out a cold leg discharge break. The TMI-1 Restart Report justified a TMI-1 flow split of 64% - 36% based on the reduced rated power of 2535 MWt. The 64% - 36% flow split has been reevaluated for a rated power of 2568 MWt. It ,

was concluded that during the period which the HPI flow split is of concern (i.e., prior to core flood tank discharge) the TMI-1 HPI system will have delivered as much water to the core as the generic SBLOCA analysis assumed. The evaluation for the power level upgrade considered the installation of cavitating venturies in the HPI lines, and the improved RCS pressure response of the revised B&W ECCS analyses. Thus, THI-1 has sufficient HPI capacity at a rated power of 2568 MWt to be bounded by the B&W generic analyses.

2. Condensate Storage Tank Capacity THI-1 has sufficient inventory to support a cooldown time in excess of 100 hours0.00116 days <br />0.0278 hours <br />1.653439e-4 weeks <br />3.805e-5 months <br />, which is greater than the estimated

' natural circulation cooldown time of 22 hours2.546296e-4 days <br />0.00611 hours <br />3.637566e-5 weeks <br />8.371e-6 months <br /> (at 10*F/hr).

This large margin assures that the adequacy of the condensate-grade feedwater supply to support a natural circulation cooldown will not be affected by a small change in

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rated power.

3. Intermediate Building Flood Level s

, The flow rate used in the analysis was greater than that expected to support operation at 2568 MWt. In addition, the flood level is limited by the amount of water available to be pumped into the building. The upgraded power level will not change the available water inventory. Therefore, the maximum predicted flood level will not change due to the upgraded power level.

4 Reactor Building Flood Level The flood level is limited by the amount of primary / secondary water available to be pumped into the building. The upgraded power level will not cause a change in either the primary system or secondary system available water inventory.

Therefore, the maximum predicted flood levels from either a primary or secondary break wil! not change due to the upgraded power level.

Based on the results of the design basis safety analyses evaluations discussed above it was concluded that all TMI-1 analyses remain conservative at 2568 MWt.

Cycle 7 bounding radiological dose projections for FSAR accidents were done at 2568 MWt as discussed in the Cycle 7 reload submittal (GPUN letter C311-88-2033 TSCR #182). Results for all events are well below the acceptance criteria of 10CFR100.

C. _0perational Considerations l Cycle 7 will be operated within Technical Specification limits developed for a full power level of 2568 MWt as described in the Cycle 7 reload, referred to above. TMI-1 fuel and core mechanical, thermal-hydraulic and nuclear designs were originally performed and approved at the 2568 MWt or higher power ratings . An evaluation of the power upgrade effect on reactor vessel accumulated fluence concluded that fluence would change in direct proportion to the power (1.3%) which is not expected to be significant.

D. Environmental Inpact An appraisal of potential environmental impacts of the power upgrade was performed. Water consumption, cooling tower drif t, fogging and other meteorological conditions, and aquatic-related concerns are bounded by previous environmental assessments based on the combined operation of TMI-1 and TMI-2. Since the operation of TMI-2 has been curtailed the previous environmental assessments bound the site conditions with the TMI-1 rated power stretch.

Thermal, chemical, and radiological effluent parameters will increase slightly as a result of the rated power stretch. These increases will not exceed the National Pollutant Discharge Elimination System (NPDES) and the 10CFR50, Appendix I limits.

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Changes in the population dose estimates in the TMINS Annual Radiological Environmental Monitoring Program (REMP) reports will be negligible.

IV. NO SIGNIFICANT HAZARDS CONSIDERATIONS ,

GPUN has determined that this Technical Specification Change Request (

poses no significant' hazards as defined by NRC in 10 CFR 50.92. j

1. 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 thermal-hydraulic and nuclear characteristics of the reactor '

core were initially designed for a rated power of 2568 MWt. The rated powe upgrade ioes not change the original design conditions. Nst of the FSAR Chapter 14 accident analyses are performed at the reference design core power level of 2568 MWt.

All other accident analyses were evaluated for the upgraded power level and determined to have insignificant impact on the accident resul ts. Therefore, operation in accordance with the proposed amendment does not involve a significant increase in the probability of occurrence or consequences of an accident previously evaluated.

2. Operation of the facility in accordance with the proposed amendment L 4

would not create the possibility of a new or different kind of 4

accident from any accident previously evaluated. The rated power upgrade does not change the original design conditions of the reactor core. Therefore, it is concluded that operation in accordance with the proposed amendment would not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. Operation of the facility in accordance with the proposed amendment would not involve a significant reduction in a margin of safety, i The rated power upgrade does not change the original design conditions of the reactor core. All existing reactor design and safety criteria are preserved at the upgraded power level.

Therefore, the margin of safety associated with each design basis safety analysis event or response remains unchanged. Therefore, it is concluded that operation in accordance with the proposed amendment does not involve a significant reduction in a margin of safety.

4 The Commission has provided guidelines pertaining to the application of the three standards by listing specific exanples in 48 FR 14870. The proposed amendment is considered to be in the same category as example (i) of amendments that are considered not likely to involve significant hazards consideration in that the proposed change is an administrative change to the Technical Specification and Facility Operating License to allow operation at the reference design power level of 2568 MWt.

Original safety analyses remain conservative for the actual plant response at the design power level. Thus, operation of the facility in accordance with the proposed amendment involves no significant hazards considerations.

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l V. IWLEMENTATION -

It is requested that the amendment authorizing this change become effective upon Cycle 7 startup, i VI. AMENDMENT FEE (10 CFR 170.21)  ;

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

$150.00.  !

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Table 1 Plant Systems The following plant systems were reviewed.

t Main Feedwater

Extraction Steam Heater Drains and Vents Circulating Water Nuclear Services Cooling Water Radwaste Systems Heat Sink Protection System Turbine-Generator Isolated Phase Bus Reactor Coolant System High Pressure Injection '

Low Pressure Injection ,

Core Flood Tanks Reactor Protection System.

Engineered Safety Features Actuation System Decay Heat Renoval Reactor Building Spray Reactor Building Cooling Makeup and Purification Spent Fuel Cooling Electrical Systems Instrumentation Systems '

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Table 2 Safety Analyses The following TMI-1 design basis and safety analyses documents were reviewed ,

to support a rated power of 2568 MWt,  ;

Uncompensated Operating Reactivity Changes Startup Accident Rod Withdrawal Accident at Rated Power Operation Moderator Dilution Accident Cold Water Accident Loss of Coolant Flow Stuck-Out, Stuck-In or Dropped Control Rod Accident Loss of Electric Power Steam Line Break Steam Generator Tube Failure

  • Fuel Handling Accident Rod Ejection Accident
  • Large Break Loss-of-Coolant Accident 3.211 Break Loss-of-Coolant Accident
  • Max 1,'wm Hypothetical Accident Waste Gas Tank Rupture Loss of Feedwater Accident -

Fuel Cask Drop Accident

! Emergency Feedwater System Response HPI Flow Split / Cooling Capacity Feedwater Line Break Accident Heat Sink Protection System Response Containnent Peak Pressure Response Containment Blowdown Response Intermediate Building Blowdown Response Condensate Storage Tank Capacity

  • Emergency Procedure Guidelines Steam Generator Tube Plugging Effects Intermediate Building Flood Level t Reactor Building Flood Level Environnental Qualification
  • 0riginally analyzed at 2568 MWt or higher. i 4

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i ATTACHMENT Technical Specification Changes 1

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