ML20045B666

Proposed Tech Specs 15.2.1.G.3,reflecting Reduction in RCS Raw Measured Total Flow Rate for Unit 2 by 2,600 Gpm & 15.2.3 Re Limiting Safety Sys Settings,Protective Instrumentation

ML20045B666
Person / Time
Site:	Point Beach
Issue date:	06/11/1993
From:	WISCONSIN ELECTRIC POWER CO.
To:
Shared Package
ML20045B657	List: ML20045B656 ML20045B666
References
NUDOCS 9306180246
Download: ML20045B666 (10)

Text

r 15.2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 15.2.1 SAFETY LIMIT, REACTOR CORE Applicability:

Applies to the limiting combinations of thermal power, reactor coolant system pressure, and coolant temperature during operation.

Objective:

To maintain the integrity of the fuel cladding.

Specification:

1. The combination of thermal power level, coolant pressure, and coolant temperature shall not exceed the limits shown in Figure 15.2.1-1 M Un W M hd}Figuie3 512 5 2?for @ iG 2. The safety limit is exceeded if the point defined by the combination of reactor coolant system average temperature and power le':el is at any time above the appropriate pressure line.

Basis:

The restrictions of this safety limit prevent overheating of the fuel and pos-sible cladding perforation which would result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excess cladding temperature because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient.

DNB is not a directly measurable parameter during operation and therefore thermal power and Reactor Coolant temperature and pressure have been related to DNB.

Unit 1 - Amendment No. 420 15.2.1-1 May 8, M89 Unit 2 - Amendment No. 423 November 1, 1989 9306180246 930611 PT PDR ADOCK 05000266- hj P pan y

n t

This relation has been developed 'to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a -

particular core location to the local heat flux, is indicative of the margin to .

i DNB.

The DNB design basis is as follows: there must be at least a 95 percent proba-bility at a 95 percent confidence level that DNB will not occur during steady  !

state operation, normal operational transients, and anticipated transients and is an appropriate margin to DNB for all operating conditions.

The curves of Figure 15.2.1-1 a' nd"1522.l f 2 are applicable for a core of 14 x 14 0FA. The curves also apply to the reinsertion of previously-depleted 14 x 14 standard fuel assemblies into an 0FA core. The use of these assemblies is justified by a cycle-specific reload analysis. The WRB-1 correlation is used to generate these curves. Uncertainties in plant parameters and DNB correlation predictions are statistically convoluted to obtain a DNBR uncertainty factor.

This DNBR uncertainty factor establishes a value of design limit DNBR. This value of design limit DNBR is shown to be met in plant safety analyses, asing values of input parameters considered at their nominal values.

s Unit 1 - Amendment No. 420 15.2.1-2 May 8, 1989 Unit 2 - Amendment No. 423 November 1, 1989

a 1 This relation has been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB design basis is as follows: there must be at least a 95 percent proba-bility at a 95 percent confidence level that DNB will not occur during steady state operation, normal operational transients, and anticipated transients and is an appropriate margin to DNB for all operating conditions.

~

The curves of Figure 15.2.1-1 anli))2]J)) are applicable for a core of 14 x 14 0FA. The curves also apply to the reinsertion of previously-depleted 14 x 14 standard fuel assemblies into an 0FA core. The use of these assemblies is justified by a cycle-specific reload analysis. The WRB-1 correlation is used to generate these curves. Uncertainties in plant parameters and DNB correlation predictions are statistically convoluted to obtain a DNBR uncertainty factor.

This DNBR uncertainty factor establishes a value of design limit DNBR. This value of design limit DNBR is shown to be met in plant safety analyses, using values of input parameters considered at their nominal values.

Unit 1 - Amendment No. 120 15.2.1-2 May-8,-1989 {

Unit 2 - Amendment No.123 November-I r-1989

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Figure 15.2.1-1 REACTOR CORE SAFETY LIMITS ,

POINT BEACH UNIT 1 660-650-2400 PSIA 640, 2250 PSIA 650-

- 2000 PSIA o_

o 620-E F

III' ' I 610-600-590-590

8. .I .2 .5 4 .5 .6 .7 .8 .9 1. 1.1 1.2 POVER treaction or nominet1 Unit 1 - Amendment No. Ife- .May 0, 1;;;-

ju.

Figure 15.2.1-2 REACTOR CORE SAFETY LIMITS POINT BEACH UNIT 2 660 650 -

. 2400 psia

~

640 - ,

2250 psic 630 - -

C h

620 - -

a 2000 psic g

610 - - -

1775 pslo

g. .

590 - -

580 , . . . . . . . . . . . .

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 1.1 1.2 Core Power (fraction of norninoO Unit 2 - Amendment No. 1993

. - . . . . ~

e p .

(3) Low pressurizer pressure - 21865 psig for operation at 2250 psia primary system pressure 21790 psig for operation at !"' sia primary system pressure (4) Overtemperature 1 AI ( In S) ,

1 1"2S SAT, (K,-K,(T( 1 n,S )-T')( 1 +r,S +K,(P-P ') -f( AI) ) ,

where AT o - indicated AT at rated power, F T = average temperature, F ,

T' s 573.9 F (tinit(1)

T -$1 570 ?0*Ff(Uni t' .2)'

P = pressurizer pressure, psig P' = 2235 psig K, s 1.30 K, = 0.0200 K3 = 0.000791 r, = 25 sec r, = 3 sec r3 - 2 sec for Rosemont or equivalent RTD

= 0 sec for Sostman or equivalent RTD

r. - 2 sec for Rosemont or equivalent RTD

= 0 sec for Sostman or equivalent RTD [

and f(AI) is an even function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests,- where q, and q, are the percent power in the top and bottom halves of the core respectively, and q, + q, is total core power in percent of rated power, such that:

(a) for q, - q, within -17, +5 percent, f(AI) = 0.

(b) for each percent that the magnitude of q, - q, exceeds +5 ,

percent, the AT trip setpoint shall be automatically reduced by an equivalent of 2.0 percent of rated power.

Unit 1 - Amendment No. M0 15.2.3-2 May4r-1989 Unit 2 - Amendment No. M 3 November 1, 1989

+ ,

~

(c) for each percent that the magnitude of q, - q, exceeds -17 percent, the AT trip setpoint shall be automatically reduced by an equivalent-of 2.0 percent of rated power.

(5) Overpower l AT ( 1 +r,S )

r,S 1 1 sat,[K,-K, ( r,S+1 ) ( 1 +r,S ) T-K,[T( 1 +r,S ) - T'))

where ATo = indicated AT at rated power, F T = average temperature, F T' s 573.9 F '(lfrjjpl)

T 'T } (570iO%(Unjfl2)'

K, s 1.089 of rated power K, = 0.0262 for increasing T

= 0.0 for decreasing T Kg = 0.00123 for T 2 T'

= 0.0 for T < T' r3 = 10 sec r3 = 2 sec.for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD r, = 2 sec for Rosemont or equivalent RTD 0 sec for Sostman or equivalent RTD (5) Undervoltage - 275 percent of normal voltage (7) Indicated reactor coolant flow per loop -

290 percent of normal indicated loop flow '

(8) Reactor coolant pump motor breaker open (a) Low frequency set point 255.0 HZ (b) Low voltage set point 275 percent of- normal voltage.

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1 Unit 1 - Amendment No. 4 M 15.2.3-3 July 31, 1989 Unit 2 - Amendment No. M6 November 1, 1989 i

With normal axial power distribution, the reactor trip limit, with allowance for errors (", is always below the core safety limit as shown on Figure 15.2.1-1 {dh Un i t51].a rid T F i g u rif l 572 ? li2ffpujii t ?2. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the reactor trip limit is automatically reduced ("(U.

The overpower, overtemperature and pressurizer pressure system setpoints include the effect of reduced system pressure operation (including the effects of fuel densification). The setpoints will not exceed the core safety limits as shown in Figure 15.2.1-1 foffUniQandiFigursil5[2?lf 2[for*Uhifi2.  ;

The overpower limit criteria is that core pnwer be prevented from reaching a value at which fuel pellet centerline melting would occur. The reactor is prevented from reaching the overpower limit condition by action of the nuclear P

overpower and overpower AT trips.

s The high and low pressure reactor trips limit the pressure range in which reactor  ;

operation is permitted. The high pressurizer pressure reactor trip setting is '

lower than the set pressure for the safety valves (2485 psig) such that the reactor is tripped before the safety valves actuate. The low pressurizer pres-sure reactor trip trips the reactor in the unlikely event of a loss-of-coolant accident (".

F The low flow reactor trip protects the core against DNB in the event of either a decreasing actual measured flow in the loops or a sudden loss of power to one or i both reactor coolant pumps. The setpcint specified is consistent with the value  :

used in the accident analysis (*'. The low loop flow signal is caused by a condi-tion of less than 90 percent flow as measured by the loop flow instrumentation.

The loss of power signal is caused by the reactor coolant pump breaker opening i

t Unit 1 - Amendment No. 420 15.2.3-6 May 8, 1989 Unit 2 - Amendment No. 423 November 1, 1989

i as actuated by either high current, low supply voltage or low electrical fre-quency, or by a manual control switch. The significant feature of the breaker trip is the frequency setpoint, 55.0 HZ, which assures a trip signal before the pump inertia is reduced to an unacceptable value. The high pressurizer water i level reactor trip protects the pressurizer safety valves against water relief.

The specified setpoint allows adequate operating instrument error (') and transient-  ;

overshoot in level before the reactor trips.

The low-low steam generator water level reactor trip protects against loss of feedwater flow accidents. The specified setpoint assures that there will be sufficient water inventory in the steam generators at the time of trip to allow for starting delays for the auxiliary feedwater system.")

Numerous reactor trips are blocked at low power where they are not required for protection and would otherwise interfere with normal plant operations. The prescribed setpoint above which these trips are unblocked assures their availability in the power range where needed. Specifications 15.2.3.2.A(1) and 15.2.3.2.C have il% tolerance to allow for a 2% deadband of the P10 bistable which is used to set the limit of both items. The difference between the nominal and maximum allowed value (or minimum allowed value) is to account for "as measured" rack drift effects.

Sustained operation with only one pump will not be permitted above 3.5 percent power. If a pump is lost while operating between 3.5 percent and 50 percent power, an orderly and immediate reduction in power level to below 3.5 percent is allowed. The power-to-flow ratio will be maintained equal to or less than unity, which ensures that the minimuu: DNB ratio increases at lower flow because the maximum enthalpy rise does not i1 crease above the maximum enthalpy rise which '

occurs during full power and fult flow operation.

Referent _el

") ") ")

FSAR 14.1.1 FSAR 14.3.1 FSAR 3.2.1

(*) (" ("

FSAR, Page 14-5 3 FfAR 14.1.2 FSAR 14.1.9

") ("

FSAR 14.2.6 F3AR 7.2, 7.3 ")

FSAR 14.1.11 Unit 1 - Amendment No M3 15.2.3-7 July 23, 1986

. Unit 2 - Amendment No.- M6

a G. OPERATIONAL LIMITATIONS The following DNB related parameters shall be maintained within the limits shown during Rated Power operation:

1. T,., shall be maintained below 578 F.  !

}

2. Reactor Coolant System (RCS) pressurizer pressure shall be maintained:  !

22205 psig during operation at 2250 psia, or 21955 psig during operation at 2000 psia.

3. Reactor Coolant System raw measured Total Flow Rate 2181,800 gpar(See Basis).

a. l:LUdiyl%181{800}g'p.mi.0iiifl b; _ L Unip 2::l$ 179; 200fsprrflun;ilt E 2 Basis: [

i The reactor coolant system total flow rate fo6UriitD of 181,800 gpm is based on an assumed measurement uncertainty of 2.1 percent over thermal design flow  !

(178,000 gpm). Thbfllr'slaht oEcldo1 sntjysltsistst AEflll6slfifEf6?UM t][2M@ 79120;0 gKipbAsed[on(fan 'sssitmidime a.sureriisntidnjsEt;'si 6jyJ6 f521Qe Rish{ dis @ffefniil distin7fibw((175,400 gpin)) The raw measured flow is based upon the use of normalized elbow tap differential pressure which is calibrated against a precision flow calorimeter at the beginning of each cycle. -

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Unit 1 - Amendment No. 4@ 15.3.1-19 May 8, 1989 Unit 2 - Amendment NO. 443 Novembe- 1, 1989 u.. .. _ __ _