ML20235L632

Amend 91 to License DPR-61,modifying Tech Spec 3.20, RCS Flow,Temp & Pressure, & Tech Spec Figure 2.2-2 to Include Revised three-loop Operation Safety Limits & RCS Flow Rate Requirements Based Upon Results of Loop Flow Measurements

ML20235L632
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	07/06/1987
From:	Thomas C Office of Nuclear Reactor Regulation
To:
Shared Package
ML20235L605	List: ML20235L601 ML20235L632 ML20235L634
References
NUDOCS 8707160767
Download: ML20235L632 (6)

Text

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~ UNITED STATES l' 3 ,., ( j' gg NUCLEAR REGULATORY COMMISSION g *'

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t WASHINGTON, D. C. 20555 s

.%[f CONNECTICUT YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-213 j HADDAM NECK PLANT AMENDMENT TO FACILITY OPERATING LICENSE ,

Amendment No. 91  :

License No. DPR-61

1. The Nuclear Regulatory Commission (the Commission) has found that:

A. The application for amendment by Connecticut Yankee Atomic Power Company (the licensee), dated December 31, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Concission's rules and regulations as 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 Commission; 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 Comission's regulations; I

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D. The issuance of this amendment will not be inimital to the common l 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 Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment and Paragraph 2.C.(2) of Facility Operating License No. DPR-61 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No. 91, are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3. This license amendment is effective as of the date of its issuance.

FOR THE NUCLEAR REGULATORY COMMISSION WA(). & msnam-Cecil 0. Thomas, Director l Integrated Safety Assessment 4 Project Directorate Division of Reactor Projects Ill/IV/V and Special Projects

Attachment:

Changes to the Technical Specifications Date of Issuance: July 6,1987 l

l ATTACHMENT TO LICENSE AMENDMENT N0. 91 FACILITY OPERATING LICENSE ND. DPR-61 DOCKET N0. 50-213

kL Revise Appendix A Technical Specifications by removing the pages identified below and inserting the enclosed pages. The revised pages are identified by the captioned amendment number and contain marginal lines indicating the area of change. '

REMOVE INSERT Figure 2.2-2 Figure 2.2-2 Page 3-39 Page 3-39 Page 3-40 Page 3-40 I

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3.20 Reactor Coolant System Flow, Temperature and Pressure l l

Applicability: Applies to MODE 1 operation.

f Objective: To set limiting conditions for operation for minimum i nominal reactor coolant flow and pressure and maximum '

inlet temperature.

Specification: A. Reactor Coolant Flow Rate 1

1) RCS Flow Rate 1257,000 gpm (four loop) j l

2) RCS Flow Rate 2197,200 gpm (three loop)  !

l B. Reactor Coolant Temperature l

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1) T inlet 6540.60F '

C. Reactor Coolant Pressure

1) Pressurizer Pressuret2000 psig* i D. 1) The RCS flow rate (four loop) shall be determined by a heat balance within 7 EFPD of achieving 100 % RATED THERMAL POWER af ter refueling.

2) If during the performance of Section D.1, Specification A.1 is not met, perform the RCS flow rate measurement (three loop) and verify compliance with Specification A.2 within the original 7 EFPD of achieving 100% rated thermal power.

3) If af ter refueling, the reactor enters Mode 1 in the 3 loop configuration before the 4 loop RCS flow test is satisfactorily completed, the RCS flow test (3 loop) shall be determined by a heat balance within 7 EFPD of achieving 65% rated thermal power. The RCS flow rate (4 loop) measurement shall also be performed per Section D.L E. Following the completion of Section D above, the above parameters shall be verified to be within the limits at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. If any of the above parameters exceeds its specified limits, l restore the parameter to within its above specified limits within two hours or reduce THERMAL l POWER to less than 5% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

This limit is not applicable during either a THERMAL POWER ramp in excess of 5% RATED THERMAL POWER per minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER. 3 l

Amendmen t No . )( y , H , 91 t_ w l

Basis:

The limiting conditions for operation have been expanded to include limits on flow, inlet temperature and pressure. {

The flow rate requirements are based on a steam generator plugging / sleeving level consistent with a maximum of 1000 total equivalent plugged tubes. The core flow rate is based on a bypass flow fraction that was conservatively shown to be 4.5% The three loop flow rate will be deemed acceptable if the four loop j measurement results are within the above limits.

Previous flow measurement results have shown that the three loop flow rate is 78% of the four loop flow rate with measurement uncertainties included. The above limits {

j conservatively assume that the three loop flow rate is 77% of the four loop flow rate.

The core inlet temperature of $40.60F includes a 440F instrument error and deadband which would allow a maximum core inlet temperature of $44.60F at i

100% power. The maximum inlet temperature of 544.60F is used in all current safety analyses, with the exception of the dropped rod analysis which used

$33.90F Tin. Sensitivity studies show that increasing the steady state temperature 10.70F from 533.90F to 544.60F will result in a small reduction in minimum DNBR. Starting the rod dropped accident from a 10.70F hotter condition will yield an increase in the end point temperature. If the dropped rod accident had been analyzed for 544.60F, the minimum DNBR would be well 4 above the fuel design limit of 1.3. l l The minimum reactor coolant pressure of 2000 psig assumes 1 30 psig for i

instrument error and deadband which would allow a minimum core pressure of j 1970 psig at 100% power.

The limiting values of the parameters in this specification are equal to, or more conservative than those assumed as the initial conditions in the accident and transient analyses; therefore, operation must be maintained within the specified limits for the accident and transient analyses to remain valid.

3-40 knendment No. M. M. 91