ML100250703
| ML100250703 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/04/1980 |
| From: | Early P Power Authority of the State of New York |
| To: | Schwencer N Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML100250704 | List: |
| References | |
| IPN-80-1, NUDOCS 8001140426 | |
| Download: ML100250703 (9) | |
Text
V.
POWER AUTHORITY OF THE STATE OF NEW YORK 10 COLUMBUS CIRCLE NEW YORK, N. Y. 10019 (212) 397.6200 GEORGE T. BERRY PRESIDENT
& CHIEF TRUSTEES OPERATING OFFICER JOHN S. DYSON JOHN W. BOSTON EXECUTIVE VICE CHAIRMAN PRESIDENT & DIRECTOR OF POWER OPERATIONS GEORGE L. INGALLS VICE CHAIRMAN JOSEPH R. SCHMIEDER EXECUTIVE VICE RICHAD M.
LYNNPRESIDENT
& CHIEF ROBERT 1. MILLONZI January 4, 1980 LEROY W. SINCLAIR SENIOR VICE PRESIDENT FREDERICK R. CLARK I PN-80-1
& CHIEF FI NANCIAL THOMAS R. FREY Director, Office of Nuclear Reactor Regulation SENIOR VICE PRESIDENT U. S. Nuclear Regulatory Commission.
EEA ONE Washington, D. C.20555 Attention:
Mr. Albert Schwencer, Chief Operating Reactors Branch No. 1 Division of Operating Reactors
Subject:
Indian Point 3 Nuclear Power Plant Docket No. 50-286 Auxiliary Feedwater:System
Dear Sir,
Enclosed for filing are three (3) signed originals and nine teen.(19) copies of a document entitled, "Application for Amendment to Operating License", and attachments thereto..
This application seeks to amend Section 3.4 of Appendix A of the Operating License to require that all three auxiliary feedwater pumps be operable during-power operations.
This application is pro vided in accordance with our December 21, 1979 letter (Response to recommeidAtioft>,_ GS-l) to the Commission.
An outage time limit of seven (7) days is~ proposed herein rather than the recommended seventy two (72) hours.
The justification for the seven (7) day limit is that reliability is maintained with only two (2) pumps operable, since 200%
required capacity is provided by two (2) motor-driven pumps and 300%
required capacity is provided by one (1) motor-driven pump plus one (1) turbine-driven pump.
In addition, a seven (7) day outage time limit is not inconsistent with some Engineered Safety Features limitations currently a part of the IP3NPP Technical Specifications.
In accordance with 10 CFR §170.22, the Authority has classified this application as Class III since the acceptability of this single safety issue is clearly identified in Mr. Eisenhut's letter of November 7, 1979.
Therefore, as required by the regulation, enclosed is a check for $4,000.00 for the filing fee.
This payment is being made under protest and the Authority reserves the right to challenge the validity of the imposition of these fees and the basis for the determination of the amount of the fees.
OX A do I
O 1 14OV-0P/1z
-2 In support of this application we are enclosing forty (40) copies of the Safety Evaluation corresponding to the proposed Technical Specification changes.
Very truly yours, Paul-arly Ass* tant Ch ef Engineer-Projects cc:
Hon. George V. Begany White Plains Library
0 9
BEFORE THE UNITED STATES NUCLEAR REGULATORY COMMISSION In the Matter of POWER AUTHORITY OF THE STATE OF NEW YORK Indian Point 3 Nuclear Power Plant
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Docket No. 50-286
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APPLICATION FOR AMENDMENT TO
.OPERATING LICENSE Pursuant to Section 50.90 of the regulations of the Nuclear Regulatory Commission (NRC), the Power Authority of the State of New York, as holder of Facility Operating License No. DPR-64, hereby applies for an Amendment to the Technical Specifications contained in Appendix A of this license.
The proposed change to the Indian Point 3 Technical Specifi cations seeks to require that all three auxiliary feedwater pumps be operable during power operations.
The proposed change to the Technical Specification is presented in Attachment I to this Application.
The Safety Evaluation. is in Attachment II.
POWER AUTHORITY OF THE STATE OF NEW YORK Assis/tant Ch ef Engineer-Projects Subscribed nd Sworn to before me this day of9ety l99'Y.
No1 &ry Public Yu RUTH.G. ZAPFevYr Notary Riblic. Zt
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ATTACHMENT I PROPOSED TECHNICAL SPECIFICATIONS CHANGE RELATED TO AUXILIARY FEEDWATER SYSTEM (APPENDIX A)
POWER AUTHORITY OF THE STATE OF NEW YORK INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO. 50-286 JANUARY 4, 1980
3.4 STEAM AND POWER CONVERSION SYSTEM Applicab ilit y Applies to the operating status of the Steam and Power Conversion System.
Objective To define conditions of the turbine cycle steam-relieving capacity.
Auxiliary Feedwater System operation is necessary to ensure the capability to remove decay heat from the core.
Specification A.
The reactor shall not be heated above 350OF unless the following conditions are met:
(1) A minimum ASME Code approved steam-relieving capability of twenty (20) main steam valves shall be operable (except for testing).
(2)
Mhime out of three auxiliary f eedwater pumps must be operable.
(3) A minimum of 360,000 gallons of water in the condensate storage tank.
(4) System piping and valves directly associated with the above components operable.
(5) The main steam stop valves are operable and capable of closing in five seconds or less.
(6) Two steam generators capable of performing their heat transfer function.
(7) City water system piping and valves directly associated with providing backup supply to the auxiliary feedwater pumps are operable.
Amendment No.
3.4-1
B. If during power operations any of the conditions of 3.4.A above, except Item (2), cannot be met within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, the operator shall start to shut down and cool the reactor below 350OF using normal operating procedures.
For Item (2) the operator shall shutdown (Keff <0.99) and cool the reactor below 350OF within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if:
(a) only 2 out of 3 pumps are operable and the third pump is not restored to operable status within 7 days, or (b) only 1 out of 3 pumps is operable and a second pump is not restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />.
C. The gross turbine-generator electrical output at all1 times shall be within the limitations of Figure 3.4-1 or Figure 3.4-2 for the applicable condi tions of turbine overspeed setpoint, number of operable low pressure steam dump lines, and condenser backpressure as noted thereon.
Basis A reactor shutdown from power requires removal of core decay heat.
Immediate decay heat removal requirements are normally satisfied by the steam bypass to the condensers.
Thereafter, core decay heat can be continuously dissipated via the steam bypass to the condenser as feedwater in the steam generator is converted to steam by heat absorption.
Normally, the capabilit y to feed the steam generators is provided by operation of the turbine cycle feedwater system.
The twenty main steam safety valves have a total combined rated capability of 15,108,000 lbs/hr.
The total full power steam flow is 12,974,500 lbs/hr, therefore twenty (20) main steam safety valves will be able to relieve the total steam flow if necessary.
Amendment No.342 3.4-2
In the unlikely event of complete loss of electrical power to the station, decay heat removal would continue to be assured by the availability of either the steam-driven auxiliary feedwater pump or one of the two motor driven auxiliary steam generator feedwater pumps,. and steam discharge to the atmosphere via the main steam safety valves and atmospheric relief valves.
One motor-driven auxiliary feedwater pump can supply sufficient feedwater for removal of decay heat from the plant.
The minimum amount of water in the condensate storage tank is the amount needed for 24 hlours at hot shutdown.
When the condensate storage supply is exhausted, city water will be used.
Two steam generators capable of performing their heat transfer function will provide sufficient heat removal capability to remove core decay heat after a reactor shutdown.
The limitations placed on turbine-generator electrical Output due to conditions of turbine overspeed setpoint, number of operable steam dump lines, and condenser back pressure are established to assure that turbine overspeed (during conditions of loss of plant load) will be within the design overspeed value considered in the turbine missile analsis.[21In the p reparation of Figures 3.4-1 and 342 h
specified number of operable L.P. steam dump lines is shown as one (1) greater than the minimum number required to act during a plant trip.
The limitations on electrical output, as indicated in Figures 3.4-1 and"3.4-2, thus consider the required performance of the L.P. Steam Dump System in tine event of a single failure for any given number of operable dump lines.
Amendment No.3.-
3.4-3
ATTACHMENT II SAFETY EVALUATION RELATED TO AUXILIARY FEEDWATER SYSTEM POWER AUTHORITY OF THE STATE OF NEW YORK INDIAN POINT 3 NUCLEAR POWER PLANT DOCKET NO. 50-286 JANUARY 4, 1980
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(a) 1P3 FSAR (b) 1P3 SER Section I -
Descripon of Modification The operating limitation for the Steam and Power Conversion System is being changed to require that all three auxiliary feedwater (AFW) pumps be operable during power operations and that the operator shall shutdown (Keff <0.99) and cool the reactor below 350OF within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> if:
(a) only 2 out of 3 pumps are operable and the third pump is not restored to operable status within 7 daysor (b) only 1 out'of 3 pumps is operable and a second pump is not restored to operable status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />. (Presently only two out of three AFW pumps are required to be operable during power operations and if this condition is not met, 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> are allowed before shutdown to 350OF is started).
Section II -
Purpose of Modification The purpose of this modification is to increase the predicted AFWS reliability by eliminating the possibility that one train will be out of service indefinitely.
Section III -
Impact of the Change The requirement of having all three AEWS pumps operable during power operations will not alter the conclusion reached in the FSAR and SER accident analysis.
Section IV -Implementation of the Modification The modification as proposed will not impact the ALARA or Fire Protection Program at 1P3.
Section V -
Conclusion The incorporation of this modification:
a) will not increase the probability nor the consequences of an accident or mal function of equipment important to safety as previously evalu ated in the Safety Analysis Report; b) will not increase the.
possibility for an accident or malfunction of a different type than any evaluated previously in the Safety Analysis Report; and c) will not reduce the margin of safety as defined in the basis for any Technical Specification.'
Section VI -
References