ML20127B077

Proposed Tech Specs Consisting of Proposed Change 117 Re Cycle 9 Operation

ML20127B077
Person / Time
Site:	Maine Yankee
Issue date:	06/14/1985
From:	Maine Yankee
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Shared Package
ML20127B075	List: ML20127B072 ML20127B077 ML20127B083
References
NUDOCS 8506210352
Download: ML20127B077 (22)

Text

,

I

,. .. . MAINE YANKEE ATOMIC POWER COMPANY i

ENCLOSURE A Summary Description of Cycle 9 Technical Specification Changes 8506210352 850614 PDR ADOCK 05000309 P PDR-5706L-HFJ

- . , . . - , . . - - - . - . - -- -.- -- . - . - - g -. ..

a...

~

i ENCLOSURE A  ;

~

. : Item- Technical

! J No. Specification Description of Change- Reason for Change' '. : -

I1 1.3 .a) Restoval of reference.to' a) Reflects replacement lof Part-Strength' CEA's .

, page 1.3-1 Part-Strength (weakened) (CEAs containing only 2 fingers with B 4C1 Control Element Assemblies (CEAs) pellets)'with Full-Strength CEAs-(Section 3.1.5 of YAEC-1479)'

f 2.. 2.1 .

a) Thermal margin / low pressure a) Reflects Cycle 9. power distributions with 1

pages 2.1-1, 2.1-4 trip coefficients modified additional-. margin for' future cycles- i 2.1-5, and 2.1-6 b) Figures 2.1-la and 2.1-lb b) Reflects Cycle 9 power. distributions with.

modified additional margin for future cycles

~

c) Figure 2.1-2 modified c) Provides margin ~for future cycles .

1 .

+

3. 2.2 a) Steady-state peak linear a) Reflects Cycle 9 Specified Acceptable Fuel I page 2.2-1 heat rates modified Design Limits (SAFDL)-for prevention of centerline melting .(Section 3.2.2_ of YAEC-1479)

4. 3.10 a) Description of maximum reactor a) Reflects 2*F increase in cold leg temperature i- pages 3.10-8 through inlet temperature assumed in assumed in Cycle 9 analysis (Section 5.1.1 of i 3.10-14, 3.10-16 safety analysis modified YAEC-1479) i through 3.10-19 b) Figure 3.10-1 modified b) Reflects Cycle 9 CEA insertion limit (Section 1-4.9.1 of YAEC-1479)'

4

c) Figure 3.10-2 and 3.10-3 c) Reflects Cycle 9 power distributions with

additional margin for future cycles i d) Figure 3.10-4 modified d) Reflects Cycle 9 radial peaking (Section 4.3

of YAEC-1479) i l e) Figure 3.10-5 modified e) Reflects Cycle 9 power distributions and RPS  ;

setpoints

} f) Figure 3.10-6 and 3.10-8 f) Reflects 2*F increase in cold leg temperature i modified assumed in Cycle 9 analysis (Section 5.1.1 of

• l' YAEC-1479) j g) Figure 3.10-9 and 3.10-10 g) Reflects Cycle 9 power distributions with -

1 modified. additional margin for future cycles '

j h) Figure 3.10-11 modified h) Reflects moderator temperature coefficient' l limits based upon a new reference LOCA analysis. ,

moderator density curve (Section .4.6 of YAEC-1479) _ _ _

s MAINE YANKEE ATOMIC POWER COMPANY h

1 d

4-ENCLOSURE B Proposed Cycle 9 Technical Specification Change Pages Maine Yankee Atomic Power Company y<

9

5706L-HFJ'

i MAINE YANKEE ATOMIC POWER COMPANY

~1;3 REACTOR Applicability-

. 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 Zircaloy-4 shall have a 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 ]

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) of the CEAs are nontrippable.

]

1.3-1 06/14/85 5706L-HFJ

,,_ , M AINE YANKEE ATOMIC POWER ' COMPANY j .

2.1 LIMITING SAFETY SYSTEM SETTING - REACTOR'PRCTECTION SYSTEM Applicability

_ 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:

2.1.1 Core Protection

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 less than orequalto10.

Where Q =' percent thermal or nuclear power, whichever is 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 = 2025. ]

B = 17.9 _

C = -10053.0

-QDNB =Al x QR1 A l and QR 1 are given in Figures 2.1-la and 2.1-lb, respectively. _

This trip may be bypassed below 105 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 15% of rated power.

2.1-1 06/14/85 5706L-HFJ

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2.1-4

- - _.W

WHERE:

QDNB " ^1

• 1 TRIP APO - 10053.0 PVAR = 2025.0 NB+ 17.9TC T = COLD LEG TEMPERATURE, F C

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l MAINE. YANKEE Thermal Margin / Low Pressure Figure '

Technical Trip Setpoint Part 2 2.1-1b Specification (QR$ versus Fraction of Rated Thermal Power) 2.1-5

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p

. _ MAINE YANKEE ATOMIC POWER COMPANY

,l' l .2.2 SAFETY LIMITS ' REACTOR CORE

~ Applicability L

- Applies to the limiting combinations of reactor power, and Reactor Coolant l

. System flow,' temperature, and' pressure during operation.

! Objective L To maintain the integrity of the fuel cladding and prevent the release of l- significant amounts of fission products to the reactor coolant.

I Specifications A.: The reactor and the Reactor Protection System shall be operated such that the Specified Acceptable Fuel Design Limit'(SAFDL) on the departure from nucleate boiling heat flux ratio (Df6R):

DPER = 1.20 using the YAEC-1 DNB heat flux correlation

-is not exceeded during' normal operation and anticipated operational occurrences.

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

the SAFDLs for prevention of fuel centerline melting.

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

Fuel Type LW R Limit, kw/ft ,

tsuc t.uu - .

E ~20.6 20.6 ]

L 21.0 19.9 ]

M -21.9 ~20.9 ]

l N 23.0 ~ 22.3 ]

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 MWD /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 greater than the coolant saturation temperature. The upper boundary of the ,

nucleate boiling regime is termed " Departure.from Nucleate Bolling" (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 failure.

2.2-1 06/14/85 5706L-HFJ

,, , MAINE YANKEE ATOMIC POWER COMPANY )

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 quandrant power tilt .will not occur.

-The moderator temperature coefficient, coolant pressure,. flow rate, and temperature specified are consistant 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 ]

tenperature measurements of + 40F 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.

<0peration with the turbine'in IWIN 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 violation of the SAFDLs. The modified symmetric. offset LCO band provides this additional margin.-

1 3.10-8 06/14/85 5706L-HFJ

m I .

je42 o z MAXIMUM OF ACTUAL OR REFERENCE POWER LEVEL (% OF RATED POWER) VS. CEA WITHDRAWAL (STEPS) 100 5

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/ CEA WITHDRAWAL BY GROUP (STEPS)

- - - - - . . ~ _ _ _ _ _ _ _ . _ _ _ _ . - . _- ___._-- . . ________ _ - - _ . _ _

COORDINATES (EXCORE SYMMETRIC OFFSET, % RATED POWER)

(-o.40,20.) (-o.40,so.) (+o.o.co.) (+o.40.5n ) (+o.40,20.)

90.0 80.0 ,

70.0 s_

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- Restricted Above the 100% Roted Power Insertion Limit.

3.10 -10

COORDINATES (EXCORE SYMMETRIC OFFSET, % RATED POWER)

(-o.40,20.) (-o.40,40.) (+o.0,50.) (+o.40,40.o) (+o.40,20.o) 80.0 70.0 60.0 m u

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.s -

NOTE: 1. This curve includes 10% calculational uncertainty

2. x 1.03 Ff = F

3. MeasuredFfshouldbeaugmentedbymeasurementuncertainty(8%)

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Allowoble Unrodded Radio! Peak Figure MAINE YANKEE Versus 3.10-4 Technical Specificotton Cycle Average Burnup  :

3.10-12

)

NOTE: CEA's are molntained at or above 100%' power insertion limit when applying 3.10.C.2.2b ,

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R MAINE YANKEE Allowable Power Level vs. Increase in Figure Technical Total Radial Peak 3.10 - 5 Speeification 3.10 -13 ,

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Maine Yankee a Cycle 9 Core Performance Analysis -

(YAEC-1479).

Page No. _ Paragraph No.* Present Text Reads: Change Text to React:3 161 Ordinate Label . . . HFP, WWRV (PEN . . . HFP WWRV Figure 5.33 W/RCP TRIP FAILURE W/EFWPS 173 Reference (56) FMV-81-162 .... December 11, 1981 USNRC Letter to MYAPCo i dated December 11,'1981 4

o unless indicated otherwise 06/14/85

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