ML20137Y842

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Amend 85 to License DPR-36,modifying Tech Specs to Reflect Cycle 9 Power Distributions,Insertion Limits & Peaking Factors & Describe Max Reactor Inlet Temp Used in Modified Safety Analyses
ML20137Y842
Person / Time
Site: Maine Yankee
Issue date: 09/30/1985
From: Butcher E
Office of Nuclear Reactor Regulation
To:
Shared Package
ML20137Y823 List:
References
NUDOCS 8510080102
Download: ML20137Y842 (20)
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i, UNITED STATES NUCLEAR REGULATORY COMMISSION

j WASHINGTON, D. C. 20555 s <

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MAINE YANKEE ATOMIC POWER COMPANY DOCKET NO. 50-309 MAINE YANKEE ATOMIC POWER STATION AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 85 License No. DPR-36

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

A. The application for amendment by Maine Yankee Atomic Power Company, (the Ifcensee) dated June 14, 1985 as supplemented August 7,1985, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Comission's rules and regulations set forth in 10 CFR Chapter I;

8. 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; D. The issuance of this amendment will not be inimical to the commen 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 Comission's regulations and all applicable reouirements 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.B(6)(b) of Facility Operating License No. DPR-36 is hereby amended to read as follows:

(b) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.85 , are hereby

- incoroorated 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 REGULATO. ', COMMIS510N w

Edward J. Butcher, Acting Chief Operating Reactors Branch #3 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 30, 1985 4

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1 ATTACHMENT TO LICENSE AMENDMENT N0.85

, FACILITY OPERATING LICENSE NO. DPR-36

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! DOCKET NO. 50-309 i

l Revise Appendix A as follows:

! Remove Paces Insert Pages  !

1.3-1 1.3-1 2.1-1 2.1-1 ,

2.1-4 2.1-4  !'

2.1-5 2.1-5 2.1-6 2.1-6 2.2-1 2.2-1

~; 3.10-8 3.10-8 .

3.10-9 3.10-9 i

! 3.10-10 3.10-10 i I 3.10-11 3.10-11 3.10-12 3.10-12 i 3.10-13 3.10-13 l 3.10-14 3.10-14  ;

3.10-16 3.10-16  !

3.10-17 3.10-17 l 3.10-18 3.10-18 L 3.10-19 3.10-19  !

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f 1.3 REACTOR Applicability l Applies to the reactor vessel, vessel core and internals, as well as the

Reactor Coolant System and components, including associated Emergency Core Cooling Systems.

Objectives To define those design criteria essential in providing for safe system operation which are not covered in Sections 2 and 3.

Specification A. Reactor Core The reactor core shall contain 217 fuel assemblies with each assembly containing 176 rods. Each fuel rod clad with Ziracioy-4 shall have a i

nominal active fuel length of 136.7 inches. The fuel shall have a maximum nominal enrichment of 3.30 weight percent U-235.

The core excess reactivity shall be controlled by a combination of boric acid chemical shim, Control Element Assemblies (CEAs) and mechanically fixed non-fuel rods when required. The non-fuel rods may be fixed alumina-boron carbide, solid metal or open tubes.

There are a total of eighty-one (81) full-length, full-strength CEAs l provided. Forty (40) of these are paired to form twenty (20) dual CEAs.

Seventy-seven (77) CEAs, including all dual CEAs, are trippable. Four (4)oftheCEAsarenontrippable.

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Amendment flo. 23,/0,f 8,63,85 1.3-1 l

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2.1 LIMITING SAFETY SYSTEM SETTING REACTOR PROTECTION SYSTEM Acolicability Applies to reactor trip settings and bypasses for the instrument channels monitoring the process variables which influence the safe operation of the plant.

Objective 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:

4 2.1.1 Core Protection j 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 is larger, j b) Thermal llargin/ Low Pressure:

Greater than or equal to: AQ0NS + BT c +C,or1835psig(whicheveris larger).

j Where T

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= cold leg temperature, OF A = 2025.

B = 17.9 l C = -10053.0 00N3 = Aj x QR)

Aj and QR) are given in Figures 2.1-la and 2.1-lb, respectively.

] This trip may be bypassed below 10% of rated power.

l c) The symmetric offset trip function shall not exceed the limits shown in Figure 2.1-2 for three loop ooeration. This trip may he bypassed below 15% of rated power.

' Amendment No. 40, Ad 32, 64, 2.1-1 74,73.g5

WHERE: Q -

DNS =A1

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AND I' PVAR = 2025.0 QDNS+ 17.9TC 10053.0 i

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Excore Symmetric Offset Y; = A=((U-L)/(U+L))+9 i

i MAINE YANKEE Thermal Margin / Low Pressure Trip Setpoint Figure i

Technlect Port 1 2.1-1c l Speelficatien (Ag versus Y,)

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! Amendment No. 23,36,70, AZ ,56,64,1A ,/d #3

WHERE:

  • QDNB " ^1 1 TRIP AND Pg = 2025.0 gNB+ * - 10053.0 C

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0,0 0,2 0.4 0.6 0,8 1. 0 1.2 Frection of Roted Thermal Power MAINE, YANKEE Thermal Margin / Low Pressure Figure '

Technicci Trip Setpoint Part 2 2.1-1b Speelfication (QR3 versus Fraction of Roted Thermal Power) 2.% 5 Amendment tio. 29,23.40,43,34,62 /4,/d, 65

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-0.6-0.5 -0. 4 -0.3 -0.2 -0.1 0.0 0.1 0.2 0.3 0,t 0.5 0.0 Excere Symmetrle Offset Y; = A*((U-L)/(U+L))+9 MA!NE YANKEE Symmetric Offset Trip Functlen Fi7Jr0 Technicc! Three Pump Operction 2.1-2 Spe:1fiectien 2.1-6 Amendment flo. 29,32,42,13,85

1 . .

1 j 2.2 SAFETY LlMlTS - REACTOR CORE j Applicability i

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

f 1 Objective

, To maintain the integrity of the fuel cladding and prevent the release of

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significant amounts of fission products to the reactor coolant. 1 Specifications t

! A. The reactor and the Reactor Protection System shall be operated such that

! the Specified Acceptable Fuel Design Limit (SAFDL) on the departure from j nucleate boiling heat flux ratio (DriBR):

, DriBR = 1.20 using the YAEC-1 DNB heat flux correlation is not exceeded during normal operation and anticipated operational I occurrences. I

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8. The reactor and the Reactor Protection System shall be operated such that S the SAFDLs for prevention of fuel centerline melting.

l A steady-state peak linear heat rate equal to:  !

Fuel Type LHGR Limit, kw/ft  !

B0C E0C E 20.6 20.6 j L 21.0 19.9 M 21.9 20.9  :

il 23.0 22.3 ,

i j are not exceeded during normal operation and anticipated operational  :

occurrences. The LHGR limit decreases linearly with Core Average

Burnup(CAB). The EOC Burnup for purposes of establishing a linear relationship is 14,000 ff40/MTU CAB.

! Basis -

j To maintain the integrity of the fuel cladding, thus preventing fission product i release to the Primary System,'it is necessary to prevent overheating of the l cladding. This is accomplished by operating within the nucleate boiling regime l of heat transfer, and with a peak linear heat rate that will not cause fuel  !

j centerline melting in any fuel rod. First, by operating within the nucleate l

boiling regime of heat transfer, the heat transfer coefficient is large enough j so that the maximum clad surface temperature is only slightly greater than the l coolant saturation temperature. The upper boundary of the nucleate boiling
regime is termed " Departure from flucleate Boiling" (DilB). At this point, there i is a sharp reduction of the heat transfer coefficient, which would result in i

{ higher cladding temperature and the possibility of cladding failure.  ;

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Amendment flo. 29,f 3,74,73,M 2.2-1  !

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fixed incore detectors or in the movable incore detector traces. The axial peaking factor can be determined from the fixed incore detectors, the movable incore detector traces or the excore detectors. The requirement that the core power distribution be shown to be within the design limits after every refueling not only ensures that the reactor can be operated safely but will also provide reasonable verification that the core was properly loaded. The requirement for operability of incore instrumentation in the instance of an excore detector channel being out of service ensures that an unobserved quadrant power tilt will not occur.

The moderator temperature coefficient, coolant pressure, flow rate, and temperature specified are consistent with the values assumed in the safety analysis. The safety analysis assumes ranges in cold leg temperature corresponding to the allowable coolant conditions given in Figure 3.10-6.

The actual values assumed in the safety analysis include an uncertainty on temperature measurements of + 4*F conservatively applied to the allowable values. The exception permits testing to determine decay heat removal capabilities of the Primary System while on natural circulation, prior to operation at higher power.

Operation with the turbine in IMPIN mode could result in a core power increase during a CEA drop transient above the initial pre-drop power level due to automatic opening of the throttle valves combined with moderator reactivity effects. Thus, additional initial overpower margin is required to preclude i violation of the SAFDLs. The modified symmetric offset LCO band provides this additional margin.

6 AmendmentNo.63.f3,73,55' 3.10-8 l

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COORDINATES (EXCORE SYMMETRIC OFFSET, % RATED POWER)

(-0.40,20.) (-0.40,50.) (+0.0,60.) (+0.40,50.) (+0.40,20.)

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Restricted Above the 100% Roted Power Insertion Limit.

3.10 -10 Amendment tio. 73,55

COORDINATES (EXCORE SYMMETRIC OFFSET, % RATED POWER)

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i 3 .1 0 - 11 i

Amendment fio. 73 85 L

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MA!NE YANKEE Al:ewebte Unredded Radic! Peck Figure Technicc! Versus 3.10-4 Specificction Cycle Average Burnup 3.to-12 l

Amendment flo. /0,/3,33,f 3,73,85

NOTE: CEA's are maintained at or above 100%' power Insertion limit when cpplying 3.10.C.2.2b 110 , ,

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MAINE YANKEE Allowchie 3 Loop Steady State figure Technical Coolent Conditions 3.10 - 6 Specifica ti on 3.10 -14 Amendment tio. AD,85

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l MAINE YANKEE Reference Power Level Figure

IcaN cA:. Versus SPEC:F:cATIc:: 3.10-8 Nominc! Cold Leg Tempercture 3.10-16 Amendment fio. 73,85

120' 1

I I

110 ,

T I I

6 1 l '/ ! 'm" 100 7

. 1

_ /- 1 LCO

, / , i '. (-0.4004 20.0 )

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l Execre Symmetrie Offset

{

Y, = Ae((U-L)/(U-K.))-Hii l.

MAJNE YANKEE Excore Symmetric Offset LCO Figure l

, Technical 3.10 - 9 l Spccificcticn 3.10-f7 A.mendment No. 772,.85 Q _ _ , . - , - , , ,

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j , i 100 1 IMPIN MODE LCO

,& (-0.4004 20.0 )

90 ,

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Specificatien 3.10 -18 Amendment flo. 74,co-0 - _. - - - _ _ _ _

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,.' .' UNACCEPTABLE OPERAi10N Os ix, l

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I 0 10 20 30 40 50 60 70 80 90 100 POWER LEVEL (7. OF PATED POWER) l, MAINE YANKEE Mederster Te:perature Coefficien Upper Li=its '

Figure TECHNICAL Versus i 3.10-11 SPECIFICATICN Power level l 3.10-19 Amendment No. 73, 85

.

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