ML20246F623

Amend 113 to License DPR-36,modifying Tech Specs to Reflect Operating Limits for Cycle 11 Operation

ML20246F623
Person / Time
Site:	Maine Yankee
Issue date:	07/10/1989
From:	Murley T Office of Nuclear Reactor Regulation
To:
Shared Package
ML20246F621	List: ML20246F617 ML20246F623 ML20246F628
References
NUDOCS 8907130327
Download: ML20246F623 (12)

Text

_- __

j s444, o, UNITED STATES y,

l y } 3 ,e ( ' ,g NUCLEAR REGULATORY COMMISSION sg y WASHINGTON, D. C. 20555 V;3./

MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.113 License No. DPR-36

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

A. The application for amendment by Maine Yankee Atomic Power Company (the licensee), dated December 28, 1988, and as clarified May 30, 1989, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission'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 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 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 Commission's regulations and all applicable requirements have been satisfied.

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2. Accordingly, the license is amended by a change to 2.B.6(a) Maximum Power Level and changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.B(6)(b) of Facility Operating License No. DPR-36 is hereby amended to read as follows:

(a) Maximum power Level The' licensee is authorized to operate the facility at steady-state reactor core power levels not in excess of 2700 megawatts thermal.

(b) Technical Specifications The Technical Specifications contained in Appendix A, as revised through Amendment No.113, are hereby incorporated in the license. The licensee shall operate the facility in accordance with the Technical Specifications.

j 3. This license amendment is effective immediately.

FOR THE NUCLEAR REGULATORY COMMISSION Ads %e.41- A

/

Thomas E. Murley, Director Office of Nuclear Reactor Regulation l

Attachment:

Changes to the Technical Specifications Date of Issuance:

July 10,1989 l

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ATTACHMENT TO LICENSE AMENDMENT NO.113 FACILITY OPERATING LICENSE NO. DPR-36 DOCKET NO. 50-309 Revise Appendix A as follows:

Remove Pages Insert Pages 2 2 2.1-1 2.1-1 2.1-4 2.1-4 2.1-5 2.1-5 2.2-1 2.2-1 3.10-9 3.10-9 3.10-13 3.10-13 3.10-14 3.10-14 -

3.10-19 3.10-19 e

i

J Co'id Shutdown Boron Concentration

. The barcn concentration shall be sufficient to maintain the reactor at least 5% delta k/k subtritical.

Hot Shutdown Boron Concentration The boron concentration shall be sufficient to maintain the reactor at least 5% delta k/k subtritical.

- Reactor Critical The reactor is considered critical for purposes of administrative control when the neutron flux logarithmic range channel instrumentation indicates greater than 10-4% of rated power. The reactor is considered subcritical when it is not critical.

Shutdown Marcin Shutdown margin shall be the sum of:

(1) the reactivity by which the reactor is sutscritical in its present condition, and (2) the reactivity associated with the withdrawn trippable CEAs less the reactivity associated with the highest worth withdrawn trippable CEA.

Low Power Physics Testina Testing performed under approved written procedures to determine control rod worths and other core nuclear properties. Reactor ~ power during these tests shall.not exceed 2% of rated power, not including decay heat, and primary system temperature and pressure shall be in the range of 260'F to 550*F and 415 psia to 2300 psia, respectively. Certain deviations from normal operating practice which are necessary to enable performing some of these tests are permitted in accordance with the specific provisions of these Technical Specifications.

Power Ranae Physics Testina Tests performed under approved written procedures to verify core nuclear de5ign properties at power and plant response characteristics. Reactor power may be greater than 2% during these measurements. Primary system average temperature and pressure shall be in the range of 500*F to 580*F and between 1700 psia tc 2300 psia, respectively. Certain deviations from normal operating practices which are necessary to enable the performance of some of these tests are permitted in accordance with specific provisions of these Technical Specifications. l Rated Power A steady-state reactor core output of 2700 MWt. ,

duadrant Power Tilt The difference between nuclear power in any core quadrant and the average in all quadrants.

TILT = [ Power in any cuad ] -1 p

( Avg. power of all quad 3

\*

Amendnent No. ES, EE, 63, 113 ,

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2.1 LIMITING SAFETY SYSTEM SETTING - REACTOR PROTECTION SYSTEM Agglieability Appli ' to reactor trip settings and bypasses for the instrument channelt monitoring the process variables which influence the safe operation of the

. plant.

Obiective To provide automatic protective action in the event that the process variables approach a safety limit.

Specification .

The Reactor Protective System trip setting limits and bypasses for the required operable instrument channels shall be as follows:

a 2.1.1 Core Protection I a) Variable Nuclear Overpower:

Less than or equal to Q + 10, or 106.5 (whichever is smaller) for Q greater than or equal to 10 and less than or equal to 100, and less than or equal to 20 for Q 1ess than or equal to 10.

Where , ,

Q = percent thermal'or nuclear power, whichever l's larger.

b) Thermal Margin / Low Pressure:

Greater than or equal to: A QDNB + BTc '+ C. or 1835 psig (whichever is larger).

Where Tc = cold leg temperature, 'F A = 2070.6 l 4 B - 17.9 C -  :-10053.0 QDNB - Aj X QR1 , j Aj and QR1 are given in Figures 2.1-la and 2.1-1b, respect'ively.

This trip may be bypassed below 101,of rated power.

c) The symmetric offset trip function shall not exceed the limits shown in ,

Figure 2.1-2 for three loop operation. This trip may be bypassed below 157. of rated power.

2.1-1 Ameridment No. 48, AB, 58, 68,  !

74, 78, 85, 113 .

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Excore Symmetric Offset Y, = A*((U-L)/(U+L))+B Thermal Margin / Low Pressure Trip Setpoint Figure MA!NE YANKEE Technical Part 1 2.1-1c Specification (A) versus Y,)

2.1 -4 _

Amendment No. 29, 38, AB, A3, 53, 53, 7A, 78, 85, 113

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0.0 0.2 0.4 0.5 0.8 2.0 1.2 Froction of Roted Thermal Power MAINE YANKEE Thermal Margin / Low Pressure Figure Technical Trip Setpoint Port 2 2.1-1b i

Specification (QR$ versus FrocUon of Roted Thermc! Power) 2.1-5 Amendment,ilo. 27, 35, pp, AB, EE, $..$, 74,78.,85,113 . . . ..

2.2 SAFETY LIMITS - REACTOR CORE Aeolicability .

Applies to the limiting combinations of reactor power, and Reactor Coolant System flow, temperature, and pressure during operation.

Obiective To miti.ntain the integrity of the fuel cladding and prevent the release of significant amounts of fission products to the reactor coolant.

Specifications A. The reactor and the Reactor Protection System shall be operated such that the following Specified Acceptable Fuel Design Limit (SAFDL) on the departure from nucleate boiling heat flux ratio (DNBR) is not exceeded during normal operation and anticipated operational occurrences. ,

DNBR - 1.20 using the YAEC-1 DNB heat flux correlation B. The reactor and the Reactor Protection System shall be operated such that the following SAFDLs for prevention of fuel centerline melting are not exceeded during normal operation and anticipated oper!.tional occurrences.

A steady-state peak linear heat generation rate (LHGR) equal to:

Fuel Tvee LHGR Limit. kw/ft E E M.L 20.8 20.1 N 21'.2 20.1 P 22.3 21.1 Q

23.2 22.2 i where the LHGR limit for each fuel type decreases linearly with Cycle Average Burnup (CAB), and the.EOC Burnup for the purposes of establishing a linear relationship is 14,500 MHD/MTU CAB.

Basis To maintain the integrity of the fuel cladding, thus preventing fission product release to the Primary System, it is necessary to prevent overheating of the cladding. This is accomplished by operating within the nucleate boiling regime of heat transfer, and with a peak linear heat rate that will not cause fuel centerline melting in any fuel rod. First, by operating within the nucleate boiling regime of heat transfer, the heat transfer coefficient is large enough so that the maximum clad surface temperature is only slightly I greater than the coolant saturation temperature. The upper boundary of the nucleate boiling regime is termed " Departure from Nucleate Boiling" (DNB). at this point, there is a sharp reduction of the heat transfer coefficient, which would result in higher cladding temperature and the possibility of cladding I failure.

2.2-1 Amendment No. 29', f#, 7#, N. M, 76, 47,113

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Amendment No. 113 3.10-9

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MAINE YANKEE Unear Heat Generation Rote (LHGR) Umits g Technical Versus 3.1 0 - 11 Specification Core Height Amendment No. 707, 113 3.10 -19

_ -___ ____-___-