ML20082T522

Core Operating Limits Rept

ML20082T522
Person / Time
Site:	Beaver Valley
Issue date:	04/24/1995
From:	DUQUESNE LIGHT CO.
To:
Shared Package
ML20082S843	List: ML20082S840 ML20082T522
References
NUDOCS 9505040065
Download: ML20082T522 (7)

Text

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INPF-73 BEAVER VALLEY UNIT 2 CYCLE 6 l CORE OPERATING LIMITS REPORT I

This Core Operating Limits Report provides the cycle specific ,

parameter limits developed in accordance with the NRC approved I methodologies specified in Technical Specification Administrative l Control 6.9.1.14. I l

Soecification 3.1.3.5 Shutdown Rod Insertion Limits l The shutdown rods shall be withdrawn to at least 225 steps.

Specification 3.1.3.6 Control Rod Insertion Limits .,

Control Banks A and B shall be withdrawn to at least 225 steps.  !

Control Banks C and D shall be limited in physical insertion as shown I in Figure 1.

Specification 3.2.1 Axial Flux Difference l NOTE: The target band is 17% about the target flux from 0% to 100% RATED THERMAL POWER.

The indicated Axial Flux Difference: j

a. Above 90% RATED THERMAL POWER shall be maintained within the 7% target band about the target flux difference.

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b. Between 50% and 90% RATED THERMAL POWER is within the limits shown on Figure 2.

c. Below 50% RATED THERMAL POWER may deviate outside the target band.

Specification 3.2.2 Fg(Z) and Fxy Limits Fg (Z) $ CEQ

• K (Z'

, for P > 0.5 P

FQ(Z) $ CEQ

• K(Z) for P $ 0.5  ;

0.5 l Where: CFQ = 2.40 P= THERMAL POWER l RATED THERMAL POWER K(Z) = the function obtained from Figure 3.

BEAVER VALLEY - UNIT 2 1 OF 6 COLR 6 i 9505040065 950424 PDR ADDCK 05000412 i p PDR  !

2  ; !NPF-73 ,

l I

The Fxy limits [Fxy(L)] for RATED THERMAL POWER within specific core planes shall be:

Fxy(L) =

Fxy(RTP) (1 + PFXY * (1-P))

Where: For all core planes containing D-BANK: ,

l Fxy(RTP) $ 1.71 For unrodded core planes:

1 Fxy(RTP) 5 1.75 from 1.8 ft. elevation to 5.7 ft. elevation i Fxy(RTP) $ 1.80 from 5.7 ft. elevation to 8.7 ft. elevation i

Fxy(RTP) 5 1,77 from 8.7 ft. elevation to 9.7 ft. elevation l

Fxy(RTP) 5 1.72 from 9.7 ft. elevation to 10.2 ft. elevation l PFXY = 0.2 P= THERMAL POWER RATED THERMAL POWER ]

l Figure 4 provides the maximum total peaking factor times relative power (Fg T *prel) as a function of axial core height during normal core operation.

Specification 3.2.3 FNDH i FNDH $ CFDH *(1 + PFDH *(1-P)) {

l Where: CFDH = 1.62 l PFDH = 0.3 P= THERMAL POWER RATED THERMAL POWER 1

l I

BEAVER VALLEY - UNIT 2 2 OF 6 COLR 6 l J

I

l  :

NPF-73 220 l l(54.53, 225)j 200 ,

/ l(100,187 3: 180 f BANK C /

7 j

160 7

/

{140 ,

E z 120 /

/ ,/

/

!(0, 114)l BANK D /

h g 100 j

7 2 /

00 60 #

/

0 '

CC 40 20 '

/

f[(8,0)l 0

0 10 20 30 40 50 60 70 80 90 100 RELATIVE POWER (Percent)

FIGURE 1 l

CONTROL ROD INSERTION LIMITS AS A FUNCTION OF POWER LEVEL BEAVER VALLEY - UNIT 2 3 of 6 COLR 6 l 1

NPF-73 100

@ gg _ UNACCEPTABLE l'II* *0) III* *0)

UNACCEPTABLE _

3 OPERATION / \ OPERATION o

/ \

m 70 '

/ \ \

60 0 f ACCEPTABLE OPERATION

\

& 50 ' '

{O 40 ~'(-31, 50) (31, 50)

$ 30 5

0- 20 10 0 l

-50 -40 -30 -20 -10 0 10 20 30 40 50 l

1 FLUX DIFFERENCE ( AI) %

l i

FIGURE 2 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER l

BEAVER VALLEY - UNIT 2 4 of 6 COLR 6 l

t a

NPF-73 1.2 (0,1.00) (6,1.00) 1.0 ~ (10.8, .94) ~

.s0 \

\ '

(12,.647) g .60 g

.40

.20 I

0.00 O 2 4 6 8 10 12 CORE HEIGHT (Feet)

FIGURE 3 Fa' NORMAllZED OPERATING ENVELOPE, K(2)

BEAVER VALLEY - UNIT 2 5 of 6 COLR 6 I

I /

NPF-73 I

2.6 I i l l '

l H 0.0, 2 . 4 0 }- '6.0, 2.40] l l 44 ,,  !, AgadbAAAA4&Ai &d6A, '

l & "A&ay j *Angggg,,gagggAAA &d AAll 3

34 l  ! I A k '

i

^

4 2.0 1  !

' S i i I i\

1.8 \

m I I i l l i \ 1 g &

l l 2. 0, 1.553{

b w

I I i j t-1.4 l 1.2 H

1.0 BASIS FXY 0.8 -

1.75 FROM 1.8 FT. UP TO 5.7 FT.

0.6 1.80 FROM 5.7 FT. UP TO 8.7 FT.

. 1,77 FROM 8.7 FT. UP TO 9.7 FT.

0.4 1.72 FROM 9.7 FT. UP TO 10.2 FT.

n ,,

0.2 0.0 s 0 2 4 6 8 10 12 l CORE HEIGHT (Feet)

FIGURE 4 MAXIMUM (FJ* Pm) VS. AXIAL CORE HEIGHT DURING NORMAL CORE OPERATION l l

i BEAVER VALLEY- UNIT 2 6 of 6 COLR 6 l l 1

l i

l

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4 ATTACHMENT A l

Beaver Valley Power Station, Unit No. 2 Cycle 6 Reload and Core Operating Limits Report Technical Specification Bases Change

- . .\

This change modifies Bases 2.1.1, Reactor Core, to incorporate the Improved Standard Technical Specification wording from NUREG-1431 to address the change in peaking factors provided in the Cycle 6 COLR.

Remove Insert l 1

B 2-1 B 2-1 B 2-2 B 2-2

+ l l

l l

l l

l l

r

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NPF-73 2.1 SAFETY LIMITS BASES L1.1 REACTOR CORE The restrictions of this safety limit prevent overheating of the fuel and possible 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 excessive cladding temperatures 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 through the WRB-1 correlation. The WRB-1 DNB correlation 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 probability that the minimum DNBR of the limiting fuel rod during Condition I and II events is greater than or equal to the DNBR limit of the DNB correlation being used (the WRB-1 correlation in this application). The correlation DNBR limit is based on the entire applicable experimental data set such that there is a 95 percent probability with 95 percent confidence that DNB will not occur when the minimum DNBR is at the DNBR limit (1.17 for the WRB-1 correlation).

Incorporating the peaking factor uncertainties in the correlation limit results in a DNBR design limit value of 1.21.

This DNBR value must be met in plant safety analyses using nominal values of the input parameters that were included in the DNBR uncertainty evaluation. In addition, margin has been maintained in the design by meeting a safety analysis DNBR limit of 1.33 in performing safety analyses.

The curve of Figure 2.1-1 shows the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

The curves are based on enthalpy hot channel factor limits prov ded in the Core Operating Limits Report (COLR).

i BEAVER VALLEY - UNIT 2 B 2-1 Revised by NRC letter dated

1 o y NPF-73 SAFETY LIMITS BASES _

2.1.1 REACTOR CORE, continued These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f(AI) function of the overtemperature AT trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature AT trip will reduce the setpoint to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE The restriction of this safety limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment atmosphere.

The reactor pressure vessel, pressurizer, and the RCS piping, valves and fittings are designed to Section III of the ASME Code for Nuclear Power Plants which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with the design criteria and associated code requirenents.

The entire Reactor Coolant System is hydrotested at 3107 psig to demonstrate integrity prior to initial operation.

2.2.1 REACTOR TRIP INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1 are the nominal values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and Reactor Coolant System are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The setpoint for a reactor trip system or interlock function is considered to be adjusted consistent with the nominal value when the "as measured" setpoint is within the band allowed for calibration accuracy.

To accommodate the instrument drift assumed to occur between operational tests and the accuracy to which setpoints can be measured and calibrated, Allovable Values for the reactor trip setpoints have been specified in Table 2.2-1. Operation with setpoints less conservative chan the Trip Setpoint but within the Allowable Value is BEAVER VALLEY - UNIT 2 B 2-2 Revised by NRC letter dated 1 i