ML20154C875

Proposed Tech Specs,Correcting Problem Areas Identified as Result of Evaluation of Gradual Decrease in Indicated delta-T Incident at Unit 2 & Correcting Unrelated Action Statement Wording Error

ML20154C875
Person / Time
Site:	McGuire, Mcguire
Issue date:	05/09/1988
From:	DUKE POWER CO.
To:
Shared Package
ML20154C874	List: ML20154C871 ML20154C875
References
NUDOCS 8805180351
Download: ML20154C875 (47)

Text

_ _ _ _ _ _ _

ATTACHMENT 1 Proposed McGuire Unit 1 and 2 Technical Specification Changes 1

1 1

8805180351 880509 PDR ADOCK 05000369 P

DCD

Ho cnA m e s en-s a s a m,

SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS 2.2 LIMITING SAFETY SYSTEM SETTINGS REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS 2.2.1 The Reactor Trip System Instrumentation and Interlocks Setpoints shall be set consistent with the Trip Setpoint values shown in Table 2.2-1.

APPLICABILITY:

As shown for each channel in Table 3.3-1.

ACTION:

With a Reactor Trip System Instrumentation or Interlock Setpoint less conservative than the value shown in the Allowable Values column of Table 2.2-1, declare the channel inoperable and apply the applicable ACTION statement requirement of Specification 3.3.1 until the channel is restored to OPERA 8LE status with its Trip Setpoint adjusted consistent with the Trip Setpoint value.

o McGUIRE - UNITS 1 and 2 2-4

TABLE 2.2-1 REACTOR TRIP SYSTEM INSTRUMENTATIO TRIP SETPOINTS

.s A

FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES

.h.

1. Manual Reactor Trip N.A.

N.A.

d

2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low setpoint - < 26% of RATED THERMAL POWER THERMAL POWER

~

k High Setpoint 1 109% of RATED High Setpoint - < 110% of RATED THERMAL POWER THERMAL POWER m

3. Power Range, Neutron Flux, 5 5% of RATED THERMAL POWER with

< 5.5% of RATED THERMAL POWER High Positive Rate a time constant 1 2 seconds with a time constant 1 2 seconds

4. Power Range, Neutron Flux,

< 5% of RATED THERMAL POWER with

~< 5.5% of RATED THERMAL POWER High Negative Rate a time constant 1 2 seconds ith a time constant 1 2 seconds

'?

5. Intermediate Range, Neutron s 25% of RATED THERMAL POWER

< 30% of RATED THERMAL POWER Flux 5

6. Source Range, Neutron Flux

$ 105 counts per second i 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 3 g

9. Pressurizer Pressure--Low 1 1945 psig 1 1935 psig f f

10. Pressurizer Pressure--High 1 2385 psig i 2395 psig

==

>v

??

~< 92% of instrument span

~< 93% of instrument span

11. Pressurizer Water Level--High

.E R

Q

12. Low Reactor Coolant Flow 1 90% of design flow per loop

• 1 89% of design flow per loop *

• 7

• Design flow is 97,220 gpm per loop.

m vv i

1

TABLE 2.2-1 (Continued) f REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

5 TRIP SETPOINT ALLOWABLE VALUES FUNCTIONAL UNIT

13. Steam Generator Water 3 12% of span from 0 to 30% of

> 11% of span from 0 to 30% of g

Level--Low-Low RATED THERMAL POWER, increasing RATED THERMAL POWER, increasing linearly to > 40% of span at to 39.0% of span at 100% of 100% of RATED THERMAL POWER RATED THERMAL POWER.

y E

[

14. Undervoltage-Reactor

> 5082 volts-each bus

> 5016 volts each bus Coolant Pumps

> 55.9 Hz - each bus

15. Underfrequency-Reactor 1 56.4 Hz - each bus Coolant Pumps

16. Turbine Trip 7

a.

Low Trip System Pressure 1 45 psig

> 42 psig b.

Turbine Stop Valve Closure 1 1% open

> 1% open

17. Safety Injection Input N.A.

N.A.

from ESF 18.

Reactor Trip System Interlocks Intermediate Range Neutron Flux, P-6,

> 1 x 10 80 amps

> 6 x 10 " amps g

Enable Block Source Range Reactor Trip a.

b.

Low Power Reactor Trips Block, P-7 y 3 10% of RATED

> 9%, < 11% of RATED y[

ea

, gg 1)

P-10 Input THERMAL POWER THERMAL POWER

-<n,ff "O

< 10% RTP Turbine

< 11% RTP Turbine 2)

P-13 Input Tapulse Pressure Tapulse Pressure t;

Equivalent '.

Equivalent oy vv i

TABLE 2.2-1 (Continued) n E}

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E

m FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E

c.

Power Range Neutron Flux, P-8,

$ 48% of RATED

$ 49% of RATED Low Reactor Coolant Loop Flow, THERMAL POWER THERMAL POWER i

and Reactor Coolant Pump Breaker g

Position a

^*

d.

Low Setpoint Power Range Neutron 10% of RATED

> 9%, $ 1EE of RATED Flux, P-10. Enable Block of THERMAL POWER THERMAL POWER Source Intermediate and Power Range Reactor Trips e.

Turbine Impulse Chamber Pressure, P-13. Input to Low Power Reactor

$ 10% RTP Turbine

$ 11% RTP Turbine Trips Block P-7 Impulse Pressure Impulse Pressure i'

Equivalent Equivalent u

19.

Reactor Trip Breakers M.A.

N.A.

20.

Automatic Trip and Interlock Logic N.A.

N.A.

• o kO i

i J

i i

TABLE 2.2-1 (Continued)

N REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS o

C k

MOTATION b

NOTE 1: OVERTEMPERATURE AT 1+t S 1

g 3)-T'] + K ( M ) - f (OI) l g

(F/M.)

(1 y s) (1 + 13s) i fE

-K2 (1 +

S)[T(3 3

1 1

a Measured AT by RID Manifold Instrumentation, Where:

AT

=

y{*3 lead-lag compensator on measured AT,

=

/

= Time constants utilized in the lead-lag controller for ti, t.

> 8 sec.. T2 1 3 sec.,

AT, 13 I

Lag compensator on measured AT, a',

7,

=

Time constants utilized in the lag cosnensator fur AT, r3 $

j.

=

13

\\

6 sec.

\\

AT, Indicated AT at RATED THERMAL POWER,)

)

=

K 5 1.200, j [

y kk 0.0222 K

=

2 oo II 1+15 l

4 The function generated by the lead-lag controller for T,,, dynamic compensation, 33 y,

g

=

NN Time constants utilized in the lead-lag controller for T,,g,

=

T4 15 K-

> 28 sec, is 5 4 sec..

14 mm EE Average temperature

• F, T

=

mw i

Lag compensator on measured T

=

g, 3

TABLE 2.2-1 (Continued) x E

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5.

5 NOTATION (Continued)

KOTE 1:

(Continued) d is Time constant utilized in the measured T,yg lag compensator, T. $

l

=

6 sec g

k T'

5 588.2*F Reference T,yg at RATED THERMAL POWER,

=

m 0.0010 %,

K

=

3 i

Pressurizer pressure, psig, P

=

2235 psig (Nominal RCS operating pressure),

P'

=

Laplace transform operator, sec 3, S

=

and fg(AI) is a 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 such that:

g are percent RATED l

l"""

~

1. 9'

1. (AI) = 0, W re q a (1) for qt 1

t THERMAL POWER fr 'he top and bottom halves of the core respectively, and q +o t

b is total THERMAL POWER in percent of RATED THERMAL POWER; ff (11) for each percent that the magnitude of q g exc M s - M, the AT Trip Setpoint l

t shall be automatically reduced by 3.151% of ita value at RATED THERMAL POWER; and

,+,+

gg (iii) for each percent that the magnitude of q g exceeds +9.0%, the AT Trip Setpoint t

M shall be automatica11,y reduced by 1.50% of its value at RATED THERMAL POWER.

I 22 i[;f 4

e 55 cc 8

TABLE 2.2-1 (Continued) a REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS c1 NOTATIGN (Continued) 5 S

NOTE 2:

OVERPOWER AT

-K3 (1

,3) (t

,q3) T -K [T(7 tgg)-T"]-fIAIIk l

(ar/a r.) N (1

)(1[,3)$ % h4 6

2 e

As defined in Note 1, Where:

AT

=

I

= As defined in Note 1 l

= As defined in Note 1 It. T2 1

As defined in Note 1,

=

y, 3

j AsdefinedinNotelh

' tai,

=

m 4

y o

K

1. M O, 4

0.02M for incnasing average twrature and 0 for decnasing average K

=

g temperature, r5 7

The function generated by the rate-lag controller for T,,g dynamic

=

y, 3

compensation, k

Time constant utilfred in

• ite-lag controller for T,,g, r, > 5 sec,

=

ry 1

As defined in Note 1, 3, E y

,e3

=

^

er As defined in Note 1 EE r,

i

=

!N 0.00169/*F for T > T" and K = 0 for T $ T",

K

=

6 j gg 6

I

==

h

TABLE 2.2-1 (Continued) 1 E

REACTOR TRIP SYSTEl' INSTRUMENTATION TRIP SETPOINTS 5

E5 NOT1TIO4 (Continued)

As defined in iote 1, T

=

w

< 588.2*F Reference T at RATED THERMAL POWER,

~

T"

=

avg E

S

=

As defined in Note 1, and

~

0 for all AI.

f (AI)

=

2 Note 3:

The channel's maximum Trip Setpoint shall not exceed its computed Trip Setpoint by more than 2%.

t 4

m b

^

EE

!= a 5?

ee

.zz

P P yg

>vh

'CC 32 l22

) :L:1 t

o re

.s 6

l M22.K, fa i~ e, o.

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR LIMITING CONDITION FOR OPERATION 3.2.3 The combination of indicated Reactor Coolant System (RCS) total flow rate and R shall be maintained within the region of allowable operation shown l

on Figure 3.2-3 for four loop operatien:

Where:

N p

AH R = 1.49 [1.0 + 0.3 (1.0 - M '

a.

THERMAL POWER b.

P = RATED THERMAL POWER Fh=MeasuredvaluesofFhobtainedbyusingthemovableincere c.

detectors to obtain a power distribution map..The measured valuesofFhshallbeusedtocalculateRsinceFigure3.2-3 includes penalties for undetected feedwater venturi fouling o,f 0.1% and for measurement uncertainties of 1.7% for flow and 4%

forincoremeasurementofFh, I

APPLICABILITY: MODE 1.

ACTION:

With the combination of RCS total flow rate and R outside the region of l

acceptable operation shown on Figure 3.2-3:

a.

Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> either:

1.

Restore the combination of RCS total flow rate and R to within the above limits, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoint to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

McGUIRE - UNITS 1 and 2 3/4 2-14 Amendment No.42(Unit 1)

Amendment No.23(Unit 2)

l l

l POWER DISTRIBUTION LIMITS LIMITING CONDITIO'N FOR OPERATION ACTION:

(Continued) b.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of initially being outside the above limits, verify through incore flux mapping and RCS total flow rate comparison that the combinatior: of R and RCS total flow rate are restored to within

-T' the above limits, or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />, c.

Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION h.2 and/or b. above; subsequent POWER OPERATION may proceed provided that the combination of R and indicated RCS

}

total flow rate are demonstrated, through incore flux mapping and RCS total flow rate comparison, to be within the region of acceptable,$.

operation shown on Figure 3.2-3 prior to exceeding the following THERMAL POWER levels:

1.

A nominal 50% of RATED THERMAL POWER, 2.

A nominal 75% of RATED THERHAL POWER, and

{

3.

Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of attaining greater than or equal to 95% of RATED THERMAL POWER.

SURVEILLANCE REQUIREMENTS 4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 The combination of indicated RCS total flow rate determined by process computer readings or digital voltmeter measurement and R shall be

{

within the region of acceptable operation of Figure 3.2.3:

a.

Drior to operation above 75% of RATED THERMAL POWER after each fuel ioading, and b.

At least once per 31 Effective Full Power Days.

4.2.3.3 The indicated RCS total flow rate shall be verifie'd to be within the region of acceptable operation of Figure 3.2-3 at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the most recently obtained value of R obtained per Specification 4.2.3.2, is T

assumed to exist.

4.2.3.4 The RCS total flow rate indicators shall be subjected to a CHANNEL CALIBRATION at least once per 18 months.

4.2.3.5 The RCS total flow rate shall be determined by precision heat balance measurement at h::t =c pr 10 unth:. M h-~-

an sw w.

)

McGUIRE - UNITS 1 and 2 3/4 2-15 Amendment No.

Unit 1)

I Amendment No.

Unit 2)

po c m

<. -.. _g 4

1 i

f P

i This page deleted.

I l

4 l

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i McGUIRE - UNITS 1 and 2 3/4 2-16 Amendment No.43 (Unit 1)

Amendment No.24 (Unit 2)

l

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i McGUIRE - UNITS 1 and 2 3/4 2-18 Amendment No.42(Unit 1)

Amendment No.23(Unit 2)

we c Aa~seh fas rara c~.r j s

3/4.3 INSTRUMENT 4 TION 3/4.3.1 REACTOR TRIP SYSTEM INSTRUMENTATION LIMITING CONDITION FOR OPERATION As a minimum, the Reactor Trip System Instrumentatf ore channels and 3.3.1 interlocks of Table 3.3-1 shall be OPERABLE with RESPONSE TIMES as shown in Table 3.3-2.

APPLICABILITY:

As shown in Table 3.3-1.

ACTION:

As shown in Table 3.3-1.

SURVEILLANCE REQUIREMENTS 4.3.1.1 Each Reactor Trip System Instrumentation channel and interlock shall' be demonstrated OPERABLE by the performance of the Reactor Trip System Instrumentation Surveillance Requirements specified in Table 4.3-1.

4.3.1.2 The REACTOR TRIP SYSTEM RESPONSE TIME of each Reactor trip function shall be demonstrated to be within its limit at least once per 18 months.

Each test shall include at least one train such that both trains are tested at least once per 36 months and one channel per function such that all channels are tested at least once every N times 18 months where N is the total number of redundant channels in a specific Reactor trip function as shown in the "Total No. of Channels" column of Table 3.3-1.

WrGUTRE - UNITS 1 and 2 3/4 3-1

TABLE 3.3-1 REACTOR TRIP SYSTEM INSTRimENTATION

/

5

/

E MINIMUM f

[4 TOTAL NO.

CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERA 8LE MODES ACTION 3

1.

Manual Reactor Trip 2

1 2

1, 2 1

g 2

1 2

3*, 4*, 5*

10 2.

Power Range, Neutron Flux - High 4

2

-3 1, 2 2

[

Setpoint

\\

Low 4

2 3

1,,2 2,

2 Setpoint 3.

Power Range, Neutron Flux 4

2 3

1, 2 2

High Positive Rate 4.

Power Range, Neutron Flux, 4

2 3

1, 2 2

{

High Negative Rate 5.

Intermediate Range, Neutron Flux 2

1 2

1

,2 3

6.

Source Range, Neutron Flux y

a.

Startup 2

1 2

2 4

b.

Shutdown 2

1 2

3*, 4*, 5*

10 5

c.

Shutdown 2

0 1

3, 4, and 5 I

7.

Overtemperature AT Four Loop Operation 4

2 3

1, 2 6

Three Loop Operation

(**)

(**)

(**)

(**)

(**)

u II

>0 i

)

i 1

l

?.:

c

l TABLE 3.3-1 (Continued) d@

RfACIOR 1 RIP SYSTEM INSTRUMENTAI10N

~

o MINIMUM TOTAL NO.

CllANNELS CHANNELS APPLICABLE l

E FUNCTIONAL UNIT Of CHANNELS 10 1 RIP OPERABLE MODES ACTION

~w 8.

Overpower AT w

~

g four Loop Operation 4

2 3

1, 2 6

Three Loop Operation

(**)

(**)

(**)

(**)

(**)

n 9.

Pressurizer Pressure Low 4

2 3

1 6

(***)

l 10.

Pressurizer Pressure--liigh 4

2 3

1, 2 6

(***)

l 11.

Pressurizer Water Level--liigh.

3 2

2 1

6 l

w J,

12.

Low Reactor Coolant flow 6,

l Single Loop (Above P-8) 3/ loop 2/ loop in 2/ loop in 1

a.

dny oper-eaCh oper-ating loop ating loop l

b.

Two Loops (Above P-7 and 3/ loop 2/ loop in 2/ loop 1

6 below P-8) two oper-each oper-dting loops ating IoOp El 6,

3" 13.

Steam Generator Water 4/stm. gen.

2/sta. gen.

3/sta. gen.

1, 2 kk Level--Low-Low in any oper-each oper-(***)l ating sim.

ating sta.

33 ee gen.

gen.

zz U 0.

j N

i tn 50 oe 1

~

O)

TABLE 3.3-1 (Continued)

I RE AC10R IRIP SYSTEM INS 1 RUM [MIAll0N c)%

MINIMUM M

10iAL NO.

CilANNELS CHANNELS APPLICA8LE OF CHANNELS 10 1 RIP OPERABLE M00E5 ACTION g

FUNCTIONAL UNIT 14.

Undervoltage-Reactor Coolant vi 4-l/ bus 2

3 3

6 Pumps (above P-7) w

%a 15.

Underfrequency-Reactor Coolant Pumps (above P-7) 4-l/ bus 2

3 1

6 a.

Low Fluid Oil Pressure 3

2 2

1 6

l 16.

Turbine Trip b.

Turbine Stop Valve Closure 4

4 1

1 Ils 1

17 Safety Injection Input 2

1 2

1, 2 9

from ESF w

l

}

w 18.

Reactor Irip System Interlocks a.

Intermediate Range Neutron Flux, P-6 2

1 2

2 8

b.

iow Power Reactor i

Trips Block, P-7 P-10 Input 4

2 3

1 8

or P-13 loput 2

1 2

1 8

[{

i oa c.

Power Range Neutron i

I@

4 2

3 1

8 3O flux, P-8 i

e n 5E d.

IIwSetpointPower 4

2 3

1, 2 8

l Range Neutron Flux, P-10 2

7

1. o1 w u i

Turbine Impulse Chamber 22 Pressure, P-13

~

2' 1

2 1

8 in' e.

E EE

,,l 4

n n

\\e<-

_/

TABLE 3.3-1 (Continued)

N REACIOR TRIP SYSTEN INSTRUNENTATION o5 MININUM E

101AL N0.

CHANNELS CHANNELS APPLICABLE Of CHANNELS 10 1 RIP OPERABLE N00ES ACTION e

FUNCIl0NAL UNIT c

2 1

2 1, 2 9, 12

[

3U

19. Reactor Trip Breakers 2

1 2

3", 4*, S*

10

~

20.

Automatic Trip and Interlock 2

1 2

1, 2 9

2 1

2 3*, 4*, S*

10 Logic m

t Y.

u

..Et I

aa A ? }'

EE hh Rn 22 J;

11

$5) ee er J

i i

)

po C os A ~ s g s Pop. i m a u r

/'

TABLE 3.3-1 (Continued)

TABLE NOTATION With the Reactor Trip System breakers in the closed position, the Control Rod Drive System capable of rod withdrawal.

Values left blank pending NRC approval of three loop operation.

namComply with the provisions of Specification 3.3.2 for any portion of the channel required to be OPERABLE by Specification 3.3.2.

1. The provisions of Specification 3.0.4 are not applicable.

"Below the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

1. elow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

8 ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STAN08Y within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 2 - With the number of OPERABLE ' channels one less than the Total,

Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripped condition a.

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, l

b.

The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l

for surveillance testing of other channels per Specification 4.3.1.1, and Either, THERMAL POWER is restricted to less than or equal c.

to 75% of RATED THERMAL POWER and the Power Range Neutron Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.

l MCGUIRE - UNITS 1 & 2 3/4 3-6 Amendment NoS4(Unit 1)

Amendment No05(Unit 2)

TA8LE 3.3-1 (Continued _)

ACTION STATEMENTS (Continued)

ACTION 3 - With the number of channels OPERA 8LE one less than the Minimum Channels OPERA 8LE requirement and with the THERMAL POWER level:

Below the P-6 (Intermediate Range Neutron Flux Interlock) a.

Setpoint, restore the inoperable channel to OPERABLE status prior to increasing THERMAL POWER above the P-6 Setpoint, and b.

Above the P-6 (Intermediate Range Neutron Flux Interlock)

Setpoint but below 10% of RATED THERMAL POWER, restore the inoperable channel to OPERA 8LE status prior to increasing THERMAL POWER above 10% of RATED THERMAL POWER.

ACTION 4 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement suspend all operations involving positive reactivity changes.

ACTION 5 - With the number of OPERABLE channels one less than the Minimus Channels OPERABLE requirement, verify compliance with the SHUT 00WN MARGIN requirements of Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> and at least once per' 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> thereafter.

ACTION 6 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

The inoperable channel is placed in the tripped condition l

a.

within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, and b.

The Minimum Channels OPERABLE requirement is met; towever, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> l

for surveillance testing of other channels per Specification 4.3.1.1 and Specification 4.3.2.1.

b ACTION 7-Dele'ted ACTION 8 - With less than the Minimum Number of Channels OPERABLE, within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> determine by observation of the associated permissive annunciator window (s) that the interlock is in its required state for the existing plant condition, or apply Specification 3.0.3.

MCGUIRE - UNITS 1 & 2 3/4 3-7 Amendment No Unit 1)

Amendment No Unit 2)

[d c m fog wto ops TABLE 3.3-1 (Continued)

ACTION STATEMENTS (Continued)

ACTION 9 - With the number of OPERABLE channels one less than the Minimum Channels OPERA 8LE requirement, be in at Itast HOT STAN08Y within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />; however, one channel may l'e bypassed for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for surveillance testing per Specification 4.3.1.1, provided the other channel is OPERABLE.

ACTION 10 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or open the Reactor trip breakers within the next hour.

ACTION 11 - With the number of OPERABLE channels less than the Total Number of Channels, operation may continue provided the inoperable channels are placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

ACTION 12 - With one of the diverse trip features (Undervoltage or shunt trip attachment) inoperable, restore it to OPERA 8LE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or declare the breaker inoperable and apply ACTION 9.

The breaker shall not be bypassed while one of the diverse trip features is inoperable except for the time required for performing maintenance to restore the breaker to OPERABLE status.

t I

l i

l McGUIRE - UNITS 1 and 2 3/4 3-8 Amendment No.74(Unit 1)

Amendment No. 55(Unit 2)

. - a

N E

TABLE 3.3-2 y

m REACTOR TRIP SYSTEM INSTRtJMENTATION RESPONSE TIMES E

b FUNCTIONAL UNIT RESPONSE TIME 1.

Manual Reactor Trip N.A.

E u

2.

Power Range, Neutron Flux 1 0.5 second*

]

3.

Power Range, Neutron Flux, N.A.

High Positive Rate

^

4.

Power Range, Neutron Flux, i

High Negative Rate 1 0.5 second*

5.

Inter 1 mediate Range, Neutron Flux N.A.

wk 6.

Source Range, Neutron Flux N.A.

w e

7.

Overtemperature AT i 8.0 seconds

• 8.

Overpower AT i 8.0 seconds

• 9.

Pressurizer Pressure--Low 1 2.0 seconds k

10.

Pressurizer Pressure--High 1 2.0 seconds sir 11.

Pressurizer Water Level--High N.A.

aa

)

• oo Neutron detectors are exempt from response time testing.

Response time of the neutron flux signal port.fon o' t 22 of the channei shall be messured from detector output or input of first electronic component in channei.

' a) 11

' i {n se c>

ce 1:

TABLE 3.3-2 (Continued) g REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES c>S M

RESPONSE TIME FUNCTIONAL UNIT b

12.

Low Reactor Coolant Flow

< 1.0 second Single Lo 4 (Above P-8) i1.0second s.

.a b.

Two Loops (Above P-7 and below P-8) u

< 3.5 seconds 13.

Steam Generator Water Level--Low-Low

< 1.5 seconds 14.

Undervoltage-Reactor Coolant Pumps j

< 0.6 seconc 15.

Underfrequency-Reactor Coolant Pumps 16.

Turbine Trip R

N.A.

a.

Low Fluid Oil Pressure M.A.

i Y

b.

Turbine Stop Valve Closure 5

N.A.

17.

Safety Injection Input from ESF N.A.

18. Reactor Trip System Interlocks M.A.

19.

Reactor Trip Breakers M.A.

20. Automatic Trip and Interlock Logic 11

=

?rI ao i

in m

1

• X

'l

??

3; nn vv

)

k

TABLE 4.3-1 N

REAC10R 1 RIP SYSIEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS o.'

TRIP m"

ANALOG ACTUATING N00ES FOR CHANNEL DEVICE wtICH c

E CilANNEL CilANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLAMCE N

FUNCTIONAL UNIT Cil[CK CALIBRATION TEST TEST LOGIC TEST IS REQUIRED 1.

Manual Reactor Trip H.A.

N.A.

M.A.

R (11)

N.A.

1, 2, 3", 4", Sa l a

w 2.

Power Range, Neutron Flux High Setpoint 5

0(2, 4),

M N.A.

N.A.

1, 2 M(3, 4),

Q(4, 6),

R(4, 5) 8##

Low Setpoint S

R(4)

M N.A.

N.A.

1

,2

{

3.

Power Range, Neutron Flux, M.A.

R(4)

M N.A.

N.A.

1, 2 High Positive Rate q

U 4.

Power Range, Neutron Flux, N.A.

R(4)

M N.A.

N.A.

1, 2 High Negative Rate S.

Intermediate Range, S

R(4, 5)

S/U(1),M N.A.

N.A.

1

,2 j

Neutron Flux l

6.

Source Range, Neutron Flux 5

R(4, 5)

S/U(1),M(9)

N.A.

N.A.

2

, 3, 4, 5

$=

7.

Overtemperature af 5

R(si)

M N.A.

N.A.

1, 2 l

k l

S R (80 M

N.A.

N.A.

1, 2 l

tf 8.

Overpower AT gg IE 9.

Pressurizer Pressure--Low S

R M

N.A.

M.A.

I 10.

Pressurizer Pressure--liigh 5

R H

N.A.

M..

1, 2 gg 33 11.

Pressurizer Water Level -liigh 5

R M

N.A.

N.A.

I oo 12.

Low Reactor Coolant flow 5

R M

N.A.

N.A.

I 1

mr

~

E TABLE 4.3-1 (Continued) 8 REACTOR TRIP SYSTEM INSTRUNENTATION SURVEILLANCE REQUIREMENTS m

TRIP ANALOG ACTUATING N00ES FOR E

CHANNEL DEVICE WICN

-Q CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST TEST LOGIC TEST IS.REdulRED 13.

Steam Generator Water Level--

S R

M N.A.

N.A.

1, 2 Low-Low l

14.

Undervoltage - Reactor Coolant N.A.

R N.A.

M N.A.

1 Pumps I

'w 1S.

Underfrequency - Reactor N.A.

R N.A.

M M.A.

1 Coolant Pumps

-w 16.

Turbine Trip

]g I

N.A.

S/U(1, 10)

N.A.

1 i

a.

Low Fluid Oil Pressure N.A.

R b.

Turbine Stop Valve Closure N.A.

R N.A.

S/U(1, 10)

N.A.

1 i

i

~

17.

Safety Injection Input from N.A.

N.A.

N.A.

R N.A.

1, 2 r

ESF 18.

Reactor Trip System Interlocks a.

Intermediate Range p

Neutron Flux, P-6 N.A.

R(4)

M N.A.

N.A.

2' b.

Low Power Reactor Trips Block, P-7 N.A.

R(4)

.M (8)

N.A.

N.A.

1

.[t!

i c.

Power Range Neutron Flux, P-8 N.A.

R(4)

M (8)

N.A.

N.A.

1 0 I

\\n

• 1

? 5)

O 1ABLE 4.3-1 (Continued)

REACTOR 1 RIP SYS1EN INSTRUMENTATION SURVEILLANCE REQUIREMENIS s

1 RIP 2

ANALOG ACIUATING M00ES'FOR CHANNEL DEVICE MtICH CHANNEL CHANNEL OPERATIONAL OPERATIONAL ACTUATION SURVEILLANCE c3d FUNCTIONAL Uhil CilECK CALIBRAT10N TEST TEST LOGIC TEST IS REQUIRED _

~

d.

Low Setpoint Power Range g

Neutron Flux, P-10 N.A.

R(4)

M (8)

N.A.

N.A.

1, 2 m

e.

Turbine Impulse Chamber Pressure, P-13 N.A.

R M (8)

N.A.

M.A.

I 19.

Reactoc Trip Breaker N.A.

N.A.

N.A.

M (7, 12)

N.A.

1,2,3",4*,Sal 20.

Automatic Trip and 1:'

Interlock Logic N A-N A-N.A.

N.A.

M (7) 1, 2, 3*,

4*, 5*

Y 21.

Reactor Trip Bypass N.A-N.A.

N.A.

M (13), R (14)

N.A.

1, 2, 3*. 4 *, 5" C

Breakers F.F na n\\

EE

=e

.-XZ E$

WW 15

a 50

&c u Q,&3 TABLE 4.3-1 (Continued)

TA8LE NOTATION With the R'eactor Trip System breakers closed and the Control Rod Orive System capable of rod withdrawal.

1 Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.

N Below P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.

H#

If not performed in previous 7 days.

(1)

Comparison of calorimetric to excore power indication above 15% of (2)

RATED THERMAL POWER.

Adjust excore channel gains consistent with calorimetric power if absolute difference is greater than 2%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

Single point comparison of incore to excore axial flux difference (3) above 15% of RATED THERMAL POWER.

Recalibrate if the absolute difference is greater than or equal to 3%.

The provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

Neutron detectors may be excluded'from CHANNEL CALIBRATION.

(4) 4 Detector plateau curves shall be obtained, evaluated, and compared (5) to manufacturer's data.

For the Intermediate Range and Power Range Neutron Flux channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

Incore - Excore Calibration, above 75% of RATED THERMAL POWER.

The (6) provisions of Specification 4.0.4 are not applicable for entry into MODE 2 or 1.

Each train shall be tested at least every 62 days on a STAGGERED (7)

TEST BASIS.

With power greater than or equal to the interlock Setpoint the (8) required operational test shall consist of verifying that the interlock is in the required state by observing the permissive l

annunciator window.

Monthly surveillance in MODES 3*, 4* and 5* shall also include (9) verification that permissives P-6 and P-10 are in their required state for existing plant conditions by observation of the permissive annunciator window.

Monthly surveillance shall include verification of the Boron Dilution Alarm Setpoint of less than or equal to five times background.

(10) -

Setpoint verification is not required.

g l

McGUIRE - UNITS 1 and 2 3/4 3-14 Amendment No.74 (Unit 1)

Amendment No.55 (Unit 2)

TABLE 4.3-1 (Continued)

/

TABLE NOTATION 1

b (11) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERASILITY of the undervoltage and shunt trip circuits for the Manual Reactor Trip Function.

(12) -

The TRIP ACTUATING DEVICE OPERATIONAL TEST shall independently verify the OPERA 81LITY of the undervoltage and shunt trip attachments of the Reactor Trip Breakers.

(13) -

Prior to placing breaker in service, a local manual shunt trip shall be performed.

(14) -

The automatic undervoltage trip capability shall be verified operable.

c, (tS} -

c<ta tr ~rio Ar sq s a rr..

, a d a r..-sn s e r r. ~r, A ~o rs., C **^ ~"I It i 4 ***'

av o s n em c.<a w~ts s n.,un en, c a s saa ro.- as rut or c a~ ~ u s os M 4 4 = 8 Af a A r r>( 34>,wwing o p g a 6 *, s w& ss wp Jw C ee.re f rtu 4,# ?><( r#f C ileaA/

(

~*arsasaas<.

,-a

<,na

.< w n.s.

n as weer as vas <> s-ass rse os rs e wo s, m a,>,,-

., u,,,,a ng,,,.,,1;s r,, n e m e a,,,ns.&,, ea a n e,e q,,

c e ~ s a ~. n.,,

, r. c o a. r,,. a,,, y, s,3 5.

i.?

5 3

i e

c.

C S

U s

2

.?

4.,

P 5

i P

McGUIRE - UNITS 1 and 2 3/4 3-14a Amendment No (Unit 1)

Amendment No (Unit 2)

}

4

iv: c mW, pn > ~ a.., ~

i 2.2 **AITING SAFETY SYSTEM SETTINGS BASES 2.2.1 REACTOR TRIP SYSTEM INSTRUMENTATION SETPOINTS The Reactur 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 core and Reactor Coolant Systes are prevented from exceeding their Safety Limits during nonnel operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents. The various Reactor trip circuits automatically open the l

Reactor trip breakers whenever a condition monitored by the Reactor Trip System reaches a preset or calculated level.

In addition to redundant chaonels and trains, the design approach provides a Reactor Trip System which monitors numerous system variables, therefore, providing Trip System functional diversity. The functional capability at the specified trip settings is required for those anticipatory or diverse Reactor trips for which no direct :

credit was assumed in the accident analysis to enhance the overall reliability of the Reactor Trip System.

The Reactor Trip System initiates a7urbine trip signal whenever Reactor trip is initiated.

This prevents the reactivity insertton that would otherwise result from excessive Reactor Coolant System cooldown and thus avoids unnecessary actuation of the Engineered Safety Features Actuation i

System.

Operation with a trip set less conservative than its Trip Setpoint but s

l within its specified Allowable Valae is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or i

less than the drift allowance for all trips including those trips assumed in the safety analyses.

Manual Reactor Trio j

The Reactor Trty System includes manual Reactor trip capability.

i l

Power Rance. Neutron Flux In each of the Power Range Neutron p ax channels there are two independent bistables, each with its own trip setting usu 'or a High and Low Range trip setting. The Low Setpoint trip provides protec. ton during suberitical and low power operations to mitigate the consequences of a power excursion beginning from low power, and the High Setpoint trip provides protection during power operations to mitigate the consequences of a reactivity excursion from all power levels.

i

)

McGUIRE - UNITS 1 and ?

8 2-3 i

Qm ca LIMITING SAFETY SYSTEM SETTINGS BASES Power Range, Neutron Flux (Continued)

The Low Setpoint trip may be manually blocked above P-10 (a power level of approximately 10% of RATED THERMAL POWER) and is automatically reinstated below the P-10 Setpoint.

Power Range, Neutron Flux, High Rates The Power Ranga Positive Rate trip provides protection against rapid flux increases which are characteristic of rod ejection events from any power level.

Specifically, this trip complements the Power Range Neutron Flux High and low trips to ensure that the criteria are met for rod ejection from partial power.

The Power Range Negative Rate trip provides protection for control rod,

drop accidents.

At high power, a rod drop accident of a single or multiple -

rods could cause local flux peaking which could cause an unconservative local DNBR to exist.

The Power Range Negative Rate trip will prevent this from occurring by tripping the reactor.

No credit is taken for operation of the Power Range Negative Rate trip for those control rod drop accidents for which g'

DNBR's will be greater than the design limit DNBR value.

i Intermediate and Source Range, Neutron Flux The Intermediate and Source Range, Neutron Flux trips provide core protection during reactor startup to mitigate the consequences of an uncon-trolled rod cluster control assembly bank withdrawal from a suberitical condition.

These trips provide redundant protection to the Low Setpoint trip of the Power Range, Neutron Flux chgnnels.

The Source Range channels will initiate a Reactor trip at about 10 5 counts per second unless manually blocked when P-6 becomes active.

The Intermediate Range channels will initiate a Reactor trip at a current level equivalent to approximately 25% of RATED THERMAL POWER unless manually blocked when P-10 becomes active.

McGUIRE - UNITS 1 and 2 B 2-4 Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

N cao m,~,. 9 LIMITING SAFETY SYSTEM SETTINGS i

BASES Overtemperature 6T The Overtemperature Delta T trip provides core protection to prevent ON8 for all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slow with respect to piping transit delays from the core to the temperature detectors (about 4 seconds),

and pressure is within the range between the Pressurizer High and Low Pressure trips. The Setpoint is automatically varied with:

(1) coolant temperature to correct for temperature induced changes in density and heat capacity of water and includes dynamic compensation for piping delays from the core to the loop temperature detectors (2) pressurizer pressure, and (3) axial power distribu-With normal axial power distribution, this Reactor trip limit is always tion.

below the core Safety Limit as shown in Figure'2.1-1. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors, the Reactor trip is automatically reduced i

according to the notations in Table 2.2-1.

Overpower ai The Overpower Delta T trip provides assurance of fuel integrity (e.g., no fuel pellet melting and less than 1% cladding strain) under all possible overpower conditions, limits the required range for overtemperature delta T The Setpoint protection, and provides a backup to the High Neutron Flux trip.

is automatically varied wii.h:

(1) coolant temperature to correct for tempera-ture induced changes in density and heat capacity of water, (2) rate of change of temperature for dynamic compsnsation for piping delays from the core to the loop temperature ditectors, and (3) axial power distribution, to ensure that the allowable heat generation rate (kW/ft) is not exceeded.

The Overpower AT trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP 9226, "Reactor Core Response to Excessive Secondary Steam Break."

McGUIRE - UNITS 1 and 2 8 2-5

two c un-on s c.4 me. 47 LIMITINGSAFET*SYSTEMJETTINGS BASES u_

Pressurizer Pressure In each of the press,ure channels, there are two 'ndependent bistables, each with its own trip setting to provide for a High and Low Pressure trip thus limiting the pressure range in which reactor operation is persitted.

The Low Setpoint trip protects against low pressure which could lead to DN8 by tripping the reactor in the event of a loss of reactor coolant pressure.

On decreasing power the Low Setpoint trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with turbine

'spulse chamber pressure at approximately 10% of full power equivalent); and on increasing power, automatically reinstated by P-7.

The High Setpoint trip functions in conjunction with the pressurizer relief and safety valves to protect the Reactor Coolant System against systes I

overpressure.

Pressurizer Water level The Pressurizer High Water Level trip is provided to prevent water relief

/

through the pressurizer safety valves. On decreasing power the Pressurizer High Water Level trip is automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 10% of full equivalent); and on increasing power, automatically reinstated by P-7.

Low Reactor Coolant Flow The Low Reactor Coolant Flow trips provide core protection to prevent DN8 by sitigating the consequences of a loss of flow resulting from the loss of one or more reactor coolant pumps.

On increasing power above P-7 (a power level of approximately 10% of RATED THERMAL POWER or a turbine impulse chamber pressure at approximately 10%

of full power equivalent), an automatic Reactor trip will occur if the flow in Above P-8 (a more than one loop drops below 89% of nominal full loop flow.

power level of approximately 48% of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any single loop drops below 89% of nominal full loop flow.

Conversely on decreasing power between P-8 and the P-7 an automatic Reactor trip will occur on loss of flow in more than one loop and below P-7 the trip function is automatically blocked.

j l

1 l

l McGUIRE - UNITS 1 and 2 8 2-6 i

l l

YaC w

n -....,

LIMITING SAFETY SYSTEM SETTING $

BASES Steam Generator Water Level The Steas Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater flow missatch resulting from loss of normal feedwater.

The specified Setpoint provides f

allowances for starting delays of the Auxiliary Feedwater System.

l Undervoltane and Underfrecuency - Reactor Coolant Puno Susses The Un'dervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNS as a result of complete loss of forced coolant flow.

The specified Setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached.

Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious For undervoltage,,

Reactor trips from momentary electrical power transients.

the delay is set so that the time required for a signal to reach the Reactor-trip breakers following the simultaneous trip of two or more reactor coolant,

For underfrequency, pump bus circuit breakers shall not exceed 1.5 seconds.

the delay is set so that the time required for a signal to reach the Reactor l

trip breakers after the Underfrequency Trip Setpoint is reached shall not On decreasing power the Undervoltage.and Underfrequency exceed 0.6 second.

Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10% of RATED THERMAL POWER with a turbine impulse chamber pressure at approximately 105 of full power equivalent); and on increasing power, reinstated automatically by P-7.

Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power the Turbine l

trip is automatically blocked by P-8 (a power level of approximately 48% of RATED THERMAL POWER with a turbine impulse chamber at c; proximately 48% of full power equivalent); and on tr, creasing power, rein

• 4ted autoestically by l

l P-8.

Safety injection Inout from ESF If a Reactor trip has not already been generated by the Reactor Trip System Instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.

The ESF Inscrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-3.

McGUIRE - UNITS 1 and 2 8 2-7

p YC cas one an y j

LIMITING SAFETY SYSTEM SETTINGS 8ASES Reactor Trio System Interlocks The Reactor Trip System Interlocks perform the following functions:

P-6 On increasing power P-6 allows the manual block of thc Source Range Reactor trip and de-energizing of the high voltage to the detectors.

On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.

P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, Turbine trip, pressurizer low pressure and pressurizer high 1o91. On decreasing power the above listed trips are automatically blocked.

P-8 On incre& sing power P-8 automatically enables Reactor trips on low flow in one or more reactor coolant loops. On decreasing power the P-8 automatically blocks the above listed trips.

~

P-10 On increasing power. P-10 allows the manual block of the Intermediate F,

Range Reactor trip and the Flow Setpoint Power Range Reactor trip; and automatically blocks the Source Range Reactor trip and de energizes the Source Range high voltage power. On decreasing power the Inter-mediate Range Reactor trip and the Low Setpoint Power Range Reactor trip are automatically reactivated.

Provides input to P-7.

P-13 Provides input to P-7.

l McGUIRE - UNITS 1 and 2 8 2-8 m,

s.

mQ fu >~ r o :-

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR The limits-on heat flux hot channel factor, RCS flow rate, and nuclear enthalpy rise hot channel factor ensure.that:

(1) the design limits on peak local power density and minimum DNBR are not excee.ied, and (2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS accep-tance criteria limit.

Each of these is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3.

This periodic surveillance is sufficient to insure that the limits are maintained provided:

a.

Control rods in a single group move together with no individual rod insertion differing by more than + 13 steps from the group demand position; b.

Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6; c.

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained; and j

d.

The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits.

FhwillbemaintainedwithinitslimitsprovidedConditionsa.through

d. above are maintained.

As noted on Figure 3.2-3, RCS flow rate and power may be "traded off" against one another (i.e., a low measured RCS flow rate is acceptable if the power level is decreased) to ensure that the calcu-lated DNBR will not be below the design DNBR value.

TherelaxationofFhas a function of THERMAL POWER allows changes in the radial power shape for all permissible red insertion limits.

R as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts forFhlessthanorequalto1.49.

This value is used in the various accidant analyses where F influences parameters other than DNBR, e.g., peak clad tem-g perature, and thus is the maximum "as measured" value allowed.

Margin between the safety analysis limit DNBRs (1.47 and 1,49 for thimble and typical cells, respectively) and the design limit DNBRs (1.32 and 1.34 for thimble and typical cells, respectively) is maintained.

A fraction of this margin is utilized to accommodtte the transition core DNBR penalty (2%) and the appropriste fuel rod bow DNBR penalty (WCAP - 8691, Rev. 1).

When an F measurem.ent is taken, an allowance for both experimental error g

and manufacturing tolerance must be made.

An allowance of 5% is appropriate McGUIRE - UNITS 1 and 2 B 3/4 2-2a Amendment No.42 (Unit 1)

Amendment No.23 (Unit 2)

(r.a ws.MN.~ y

/Tc-This page dele ted.

McGUlce - UNITS 1 and 2 B 3/4 2-3 Amendment No. 42 (t!nic 1)

Amendment No. 23 (Unit 2)

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POWER DISTRIBUTION LIMITS BASES HEAT FLUX HOT _OHANNEL FACTOR and RCS FLOW RATE AND NUCLEAR ENTHALPY RISE HOT CHANNEL FACTOR (Continued) for a full-core map taken with the Incore Detector Flux Happing System, and -

3% allowance is appropriate for manufacturing tolerance.

When RCS flow rate and F are measured, no additional allowances art g

necessary prior to comparison with the limits of Figure 3.2-3.

Measurement errors of 1.7% for RCS total ' flow rate and 4% for Fhhave been allowed for in determination of the design DN3R value.

The measurement error for RCS total flow rate is based upon performing a precision heat balance and using the result to calibrate the RCS flow rate indicators.

Potential fouling of the feedwater venturi which might not be detected could bias the' result from the precision heat bclance in a non-conservative manner.

Therefore, a penalty of 0.1% for undetected fouling of the feedwater venturi is included in Figure 3.2-3.

Any fouling which mightt bias the RCS flow rate measurement greater than 0.1% can be detected by monitoring and trending various plant performance parameters.

If detected, action shall t,a taken before performing subsequent precision heat balance measurements, i.e., either the effe:t of the fouling shall be quantified and f

compensated for in the RCS flow rate measurement or the venturi shall be cleaned to eliminate the fouling.

The 12-hour periodic surveillance of indicated RCS flow is sufficient to detect only flow degradation which could lead to operation outside the accept-able region of operation shown on Figure 3.2-3.

The hot channel factor F (z) is measured periodically and increased by a cycle and height dependent power factor appropriate to either RA00 or Base load operation, W(z) or W(z)BL, to prcvide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the effects of normal q

operatiori transients and was determined from expected power control maneuvers over the full range of burnup conditions in the core. W(z)BL accounts for the more restrictive operating limits allowed by Base load operation which result in less severe transfer, values.

The W(z) function fer normal operation is provided in the Peaking Facter Limit Report per Specification 6.9.1.9.

l I

i 1

McGUIRE - UNITS 1 and 2 B 3/4 2-4 Amendment No. 42 (Unit 1)

Amendment No. 23 (Unit 2) l

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.ao y 3/4.3 INSTRUMENTATION BASES 3/4.3.1 and 3/4'.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION The OPERA 8ILITY of the Reactor Trip and Engineered Safety Features Actuation System instrumentation and interlocks ensure that:

(1) the associated ACTION and/or Reactor trip will be initiated when the parameter monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic and sufficient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance consistent with maintaining an appropriate level of reliability of the Reactor Protection and Engineered Safety Features Instrumentation and (3) sufficient system functions capability is available from diverse parameters.

The OPERABILITY of these systems is required to provide the overall reliability, redundancy, and diversity assumed available in the facility design for the protection and mitigation of accident and' transient conditions.

The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses.

The Surveillance Requirements specified for these systems ensure that the overall system functional capability is maintained comparable to the original design standards.

The>

periodic surveillance tests performed at the minimum frequencies are sufficient to demonstrate this capabi.lity, i

Specified rurveillance intervals and surveillance and maintenance outage tir.es have been determined in accordance with WCAP-10271, "Evaluation of Sur-veillance Frequencies and Out of Service Times for the Reactor Protection Instrumentation System," and supplements to that report.

Surveillance inter-vals and out of service times were determined based on maintaining an appro-priate l'evel of reliability of the Reactor Protection System and Engineered Safety Features instrumentation.

(Implenientation of quarterly testing of RTS is being postponed until after approval of a similar testing interval for ESFAS.) The NRC Safety Evaluation Report for WCAP-10271 wcs provided in a letter dated February 21, 1985 from C. O. Thomas (NRC) to J. J. Sheppard (WOG-CP&L).

The measurement of response time at the specified frequencies provides assurance that the Reactor trip and the Engineered Safety Feature actuation i

associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicateu as not applicable.

Response time may be demonstrated by any series of sequential, overlapping, or total channel test measurements provided that such tests demonstrate the total channel response time as defined.

Sensor response time verification may be demonstrated by either:

(1) in place, onsite, or offsite test measurements, or (2) utilizing j

replacement sensors with certified response times.

The Engineered Safety Features Actuation System senses selected plant parameters and determinos whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various acciden'.s, events, and transients.

Once the required logic combination is completed, the system sends actuation signals to those Engineered Safety Features components whose aggregate function best serves the requirements of the condition.

As an example, the McGUIRE - UNITS 1 and 2 8 3/4 3-1 Amendment NoS4(Unit 1)

Amendment No35(Unit 2) j i

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fn > ~ c, wy INSTRUMENTATION BASES 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATION (Continued) following actions may be initiated by the Engineered Safety Features Actuation System to mitigate the consequences of a stease line break or loss-of-coolant (1) Safety Injection pumps start and autcoatic valves position, accident:

(2) Reactor trip, (3) feedwater isolation, (4) startup of the emergency diesel generators, (5) containment spray pumps start and automatic valves position, (6) containment isolation, (7) steam line isolation, (8) Turbine trip, (9) auxiliary feedwater pumps start and automatic valves position, and (10) nuclear service water pumps start and automatic valves position.

Technical Specifications for the Reactor Trip Breakers and the Reactor Trip Bypass Breakers are based upon NRC Generic letter 85-09 "Technical Specifica-tions for Generic Letter 83-28, Item 4.3," dated May 23, 1965 i

McGUIRE - UNITS 1 and 2 8 3/4 3-la Amendment Ho74(Unit 1)

Amendment No)5(Unit 2)

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INSTRUMENTATION JASES REACTOR PROTECT' -

7,'W AND ENGINEERED SAFETY FEATURES ACTUATION SYSTEM INSTRUMENTATP d.oni

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,eered Safety atures Actuation System interlocks perform the following r. -r.iore P-4 Ret

- Actuates Turbine trip, closes main feedwater valves on T,y

'ow Setpoint, prevents the opening of the main feedwater valve which were closed by a Safety Injection or High s

Steam Generator Water Level signal, allows Safety Injection block so that components can be reset or tripped.

Reactor not tripped prevents manual block of Safety Injection.

j P-11 Defeats the manual block of Safety Injection actuation on low pressurizer pressure and low steamline pressure and defeats steam-line isolation on negative steamline pressure rate.

Defeats the manual block of the moter-driven auxiliary feedwater pumps on trip of main feedwater pumps and low-low steam generator water level.

P-12 On increasing reactor coolant loop temperature, P-12 automatically provides an arming signal to the steam dump system.

On decreasing reactor coolant loop temperature, P-12 automatically removes the arming signal from the steam dump system.

"g e-14 On increasing steam generator level, P-14 automatically trips all feed-water isolation valves and inhibits feedwater control valve modulation.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING FOR PLANT OPERATIONS The OPERABILITY of the radiation monitoring instrumentation for plant operations ensures that:

(1) the associated action will be initiated when the radiation level monitored by each channel or combination thereof reaches its Setpoint, (2) the specified coincidence logic is maintained, and (3) suffi-cient redundancy is maintained to permit a channel to be out-of-service for testing or maintenance. Th.e radiation monitors for plant operations senses radiation levels in selected plant systems and locations and determines whether or not predetermined limits are being exceeded.

If they are, the signals are combined into logic matrices sensitive to combinations indicative of various accidents and abnormal conditions.

Once the required logic combination is completed, the system sends actuation signals to intiate alarms or automatic isolation action and actuation of Emergency Exhaust or Ventilation Systems.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minir's complement of equipment ensures that the measurements obtained from

'e o? this system accurately represent the spatial neutron flux distribution McGUIRE - UNITS 1 and 2 B 3/4 3-2

ATTACHMENT 2 Justification and Safety Analysis Following the startup of McGuire Unit 2/ Cycle 2 in May 1985, a gradual de-crease in the indicated value of the full-power Delta-T across the core was identified.

The reactor core temperature difference (Delta-T) is the dif-ference between the reactor inlet water temperature (cold leg) and the outlet temperature (hot leg).

This Delta-T is used as a measure of reactor power for both the overtemperature and overpower Delta-T reactor trip setpoints (the Overtemperature Delta-T and Overpower Delta-T trip setpoints are detailed in Technical Specification Table 2.2-1), and is calibrated (scaled) to read 100%

's determined by precision secondary system calorimetric measurements.

In the latter part of June, approximately six weeks after power escalation commenced for Cycle 2, a precision heat balance was performed to verify Reactor Coolant System (RCS) flow (in accordance with T.S.

4.2.3. 5).

During this six week time frame, the indicated Delta-T had decreased linearly in each loop by ap-proximately 1 degree-F.

The precision heat balance indicated an increase in RCS flow from Cycle 1 and a corresponding decrease in the measured full-power Delta-Ts for the four loops.

Because the Delta-T channels for the overtem-perature Delta-T and overpower Delta-T setpoints were scaled to the full-power Delta-Ts obtained during the Cycle 1 precision heat balance, the Delta-T chan-nels were underpredicting core power by as much as 5% rated thermal power.

In view of the apparent non-conservative indication of power in the Delta-T chan-nels, the Delta-T channels were rescaled to the more conservative, lower val-ues of Delta-T obtained from the more recent precision heat balance.

This event is more fully discussed in Licensee Event Report 370/85-24 (Attach-ment 2A).

Several Technical Specification problem areas surfaced as a result of the evaluation associated with the Delta-T incident at PkGuire which should be revised in order to avoid future interpretation problems.

Some of these Tech.

Spec. problem areas and proposed resolutions are delineated below (other prob-lem areas are being addressed by previous submittals or administrative 1y with-in Duke l'ower):

I.

Technical Specification Table 2.2-1 Notes 1 and 2 of Technical Specification Table 2.2-1 represent the ~;ertem-perature Delta-T and overpower Delta-T trip setpoints in units of degrees Fahrenheit.

However, at McGuire the Delta-T channels and the overtemperature Delta-T and overpower Delta-T setpoints provide an indication in units of percent full-power Delta-T.

In order to complement the manner in which the instrumentation is calibrated at the station, Notes 1 and 2 should be rewrit-ten as follows:

Note 1:

Overtemperature LT (aT/LTo) (

• ) (

)<K

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1+T s 1+

s 2

3 Note 2:

Overpower AT 1

(6T/ ATo) (

s) (

)<g 1+T s 1+t s -

2 3

l Representing the setpoint equations in this manner shows that rescaling Delta-T (indicated Delta-T at rated thermal power) has no effect on the set-point and simply adjusts the value of power indicated by the Delta-T channels.

This proposed amendment is only a rearrangement of the setpoint equations and does not involve any technical changes to the equatfons themselves.

In addi-tion, the definition of Delta-T in the tables is relocated to maintain cor-respondence with the order of the term in the equations.

These changes are administrative in nature and involve no safety concerns.

II.

Technical Specification 4.2.3.5 and Table 4.3-1 Technical Specification Table 4.3-1 requires, on an 18 month frequency, a channel calibration of the overpower Delta-T and overtemperature Delta-T chan-nels.

This is interpreted as requiring a verification of calibration for the cir-cuitry that derives the Overpower Setpoint, Overtemperature Setpoint, and the Tavg channel constants as stated in Technical Specification Table 2.2.1 notes 1 and 2.

This "Channel Calibration" is performed at the required 18 montn frequency.

The Delta-T portion of overpower Delta-T and overtemperature Delta-T should be rescaled and recalibrated to a conservative value at the beginning of each fuel cycle prior to power escalation.

The conservative values of Delta-T are necessary based on operational exper-1ence which has shown that actual values of Delta-T at 100% may dif fer from one fuel cycle to the next.

The process of rescaling Delta-T to a conservative value is performed by sub-tracting a predetermined value, typically 1 degree-F, from the previous cycle Delta-T.

This ensures that the Delta-T channels conservatively overpredict power until the cycle specific 100% values for each loop's Delta-T can be obtained.

The cycle specific 100% values for each loop Delta-T's are deter-mined during the Reactor Coolant System Flow Test which is performed as soon as possible af ter the unit reaches 100% power.

Once the 100% Delta-T values are supplied, rescaling of Delta-T is performed, procedure changes are made incorporating these new values for use over the next fuel cycle, and calibra-tion of the Delta-T circuit hardware is initiated and completed.

The new 100%

Delta-T values are also incorporated into the monthly procedures for verifica-tion of calibration under the "Analog Channel Operational Test" requirements.

Westinghouse has performed calculations which show that the preferred method to account for these differences is to rescale Delta-T to the " As-Mea sure d" 100% value for each loop and to set the T' and the T" constants to the values as stated in Table 2.2.1 notes 1 and 2.

This ensures that the assumptions in the safety analysis and operational margin are maintained.

Failure to rescale Delta-T will either restrict operational margin or remove analysis margin possibly to the point where the assumptions of the analysis are violated.

The omission of the calibration of the Delta-T channels during power esca-lation has affected both McGuire Unit 1 and Unit 2.

Although the decreasing Delta-T in Unit 2 brought the problem into view, the Unit I channels had not been calibrated during the startup of each fuel cycle (however, since the full power Delta-T at Unit I had not changed dramatically over the first three cycles, the Delta-T channel errors were not of the magnitude present at U-nit 2).

The channel calibration was being performed on an 18 month basis without regard to the cycle startup requirements.

Calibration of the Delta-T channels to the new 100% values should only be necessary at the beginning of each cycle, as it is expected that all drifts and fluctuations over the course of a cycle should remain within the allowances assumed in the Safety Analysis for that cycle.

Aside from rescaling the full power Delta-T, rescaling and recalibration of the Overpower Setpoint, Overtemperature Setpoint, and Tavg constants need only be performed if changes to Technical Specification Table 2.2.1 notes 1 and 2 are made.

These changes would be made due to new safety analysis or thermal hydraulic analysis for a specific upcoming fuel cycle and would be submitted to the NRC for review and approval.

Recalibration of the af fected setpoints would be made during the unit shutdown prior to startup.

Accordingly, the proposed amendments to Technical Specification 4.2.3.5 and Table 4.3-1 reflect this position.

(Duke has been administrative 1y imple-menting these requirements since the McGuire incident).

These items are clar-ification of the intent and do not involve relaxation of any existing require-ments, and in fact are more conservative / restrictive since the current speci-fications do not specifically require rescaling at the beginning of each fuel cycle or if more than one cycle occurs within an 18 month span.

III. Technical Specificatien Table 3.3-1 Technical Specification Table 3.3-1 Action Statement 7 is revised to read "deleted" rather than "delete" to better reflect that a previously existing action statement had been deleted (ref. McGuire License Amendments 54(Unit 1)/35(Unit 2)) and is not an action to delete something.

This change is ad-ministrative in nature and involves no safety concerns.

Note:

This change is not related to the previously discussed Delta-T event.

Based upon the preceding justification and safety analysis, Duke Power Company concludes that the proposed amendments are necessary and will not be inimical to the health and safety of the public.

No changes to the Technical Specifi-cation Bases are necessitated as a result of the proposed amendments.

ATTACHMENT 3 Analysis of Significant Hazards Consideration As required by 10CYR 50.91, this analysis is provided concerning whether the proposed amendments involve significant hazards considerations, as defined by 10CFR 50.92.

Standards for determination that a proposed amendment involves no significant hazards considerations are if operation of the facility in accordance with the proposed amendment would not:

1) involve a significant increase in the probability or consequences of an accident previously evalu-ated; or 2) create the possibility of a new or different kind of accident from any accident previously evaluated; or 3) involve a significant reduction in a margin of safety.

t I.

Technical Specification Table 2.2-1 The proposed rearrangement of the setpoint equations of Technical Specifica-tion Table 2.2-1 (along with attendant relocation of the definition of Del-l ta-T it easier to visualize the fact that rescaling Delta-T has no effeE) makes t on the setpoint side of the equation but simply adjusts the Salue of

{

power indicated by the Delta-T channels.

Since the proposed amendment does t

not involve any technical changes to the equations or hardware changes in the plant, but rather is only a reordering of the equation's terms and definitions l

as presented in the technical specifications, the changes are administrative l

in nature ard involve no significant hazards considerations.

l The Commission has provided examples of amendments likely to involve no sig-nificant hazards considerations (48FR14870). One example of this type is (i),

[

"A purely administrative change to Technical Specifications :

For example, a change to achieve consistency throughout the Technical Specifications, correc-i tion of an error, or a change in nomenclature".

This example can be applied to the proposed reordering.

j i

II.

Technical Specification 4.2.3.5 and Table 4.3-1 The proposed amendments rewording Technical Specification 4.2.3.5 and adding a

[

footnote to Technical Specification Table 4.3-1 require that the calibration of Delta-T channels (for the 18 month channel calibration of the overpower Delta-T and overtemperature Delta-T reactor trip system instrumentation chan-

[

nels) be performed at the beginning of each fuel cycle (upon completion of the precision heat balance). This ensures that assumptions in the safety analysis and operational margin are maintained since operational experience has shown that actual values of Delta-T at 100% power may differ from one fuel cycle to i

the next (failure to rescale Delta-T will either restrict operational margin or remove analysis margin possibly to the point where the assumptions of the analysis are violated).

These changes are clarification of the intent and do i

not involve relaxation of any existing requirements, and in fact are more conservative / restrictive since the current specifications do not specifically require rescaling at the beginning of each fuel cycle or if more than one i

cycle occurs within an 18 month span.

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The proposed amendments would not involve a significant increase in the prob-ability or consequences of an accident previously evaluated as the changes are j

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clarifications of intent of the current specifications and do not constitute any actual changes to plant procedures or hardware and thus can have no effect on accident causal mechanisms or consequences.

No possibility of a new or different kind of accident from any accident previously evaluated is created by the proposed changes since the changes are administrative in nature and accident causal mechanisms are not af fected as discussed above.

The changes do not involve a significant reduction in a margin of safety because they do not involve relaxation of any existing requirements, and in fact are more conservative / restrictive than the current specification's wording.

Another Commission provided example of actions not likely to involve a signi-ficant hazards consideration is (ii), "A change that constitutes an additional limitation, restriction, or control not presently included in the technical specifications:

For example, a more stringent surveillance requirement".

Since the proposed wording is more specific as to when the af fected surveil-lances are to be perfo rmed, the above cited example can be applied to these amendments.

III. Technical Specification Table 3.3-1 The proposed amendment to Technical Specification Table 3.3-1 Action State-ment 7 to read "deleted" rather than "delete" is a purely administrative change to better reflect that a previously existing action statement had been deleted, not an action to delete something.

As such, clearly no significant hazards consideration is involved, and the previously cited commission provid-ed Example (1) in Section I above can be applied to this amendment.

Based on the preceding analyses, Duke Power Company concludes that the pro-posed amendments do not involve a significant hazards consideration.

Document Control Desk November 13, 1985 Page 2 bec:

P. M. Abraham R. S. Bhatnager R. T. Bond D. R. Bradshav K. S. Canady L. M. Coggins R. C. Futrell S. A. Gewehr R. L. Gill G. W. Hallman T. P. Harrall C. L. Hartzell A. R. Hollins (SSD)

T. L. McConnell E. O. McCraw J. J. McCool M. D. McIntosh T. E. Mooney P. B. Nardoci R. P. Ruth (MNS)

N. A. Rutherford A. L. Snow J. G. Torre G. E. Vaughn R. L. Weber L. L. Williams (W)

File:

MC-801.02-MC-815.04 4

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A DUKE POWER GOMPANY P.O. DOX 33189 CHARLOTTE. N.C. 98949 HAL B. TUCKER Trturnown (704) 3 4 4838 exa P829133F9 W1CLEAS Ptoet w oe November 13, 1985 Document Control Desk U. S. Nuclear Regualtory Commission Washington, D. C. 20555

Subject:

McGuire Nuclear Station Docket No. 50-370 LER 370/85-24 Gentleme.n:

Pursuant to 10 CFR 50.73 Sections (a)(1) and (d), attached is Licensee Event Report 370/85-24 concerning a gradual decrease in the indicated full power delta-T.

This event was considered to be of no significance with respect to the health and safety of the public.

Very truly yours.

fW Hal B. Tucker JBD/hrp Attachment cc:

Dr. J. Nelson Grace, Regional Administrator M&M Nuclear Consultants U. S. Nuclear Regulatory Commission 1221 Avenue of the Americas Region II New York, New York 10020 101 Marietta Street. NV, Suite 2900 Atlanta, Georgia 30323 Mr. W. T. Orders NRC Resident Inspector American Nuclear Insurers McGuire Nuclear Station c/o Dottie Sherman, ANI Library The Exchange, Suite 245 270 Farmington Avenue Farmington, CT 06032 INPO Records Center Suite 1500 1100 Circle 75 Parkway Atlanta, Georgia 30339 Of I I 0 I n rn l-

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Following the startup of Unit 2 Cycle 2 in May 1985, a gradual decrease in the indicated value of the full power Delta-T was identified. The indicated Delta-T decreased linearly in each loop by approximately 1 degree-F. Delta-T is used as a measure of reactor power for both the overpower and overtemperature Delta-T reactor trip setpoints. The decrease in Delta-T has caused these dynamic reactor trip functions to be improperly scaled in a non-conservative direction.

Two potential causes for the decrease have been identified:

a change in hot leg te=perature streaming patterns which supply coolant samples to the temperature sensors; a reduction in thermal power, possibly caused by fouling of the feedwater flow venturi meters.

The channel errers have been analyzed to determine the impact on FSAR accident analyses. The results of this evaluation indicate that only a narrow range of steam line break accidents may have exceeded the design basis during this incident.

Calibration procedures will be developed to set Delta-T values for each new fuel cycle; and the possible feedwater venturi fouling will be investigated.

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..e w w.,on Following the startup of Unit 2 Cycle 2, a gradual decrease in the indicated value of the full-power reactor coolant temperature difference was identified. During the eight weeks of full-power operation following the cycle 2 refueling, the indicated Delta-T had decreased linearly in each of the four loops by approximately 1 degree-F. Delta-T is us:ed as a measure of reactor power for both the over-te=perature and overpower Velta-T reactor trip setpoints.

The decrease in Delta-T had caused these dynamic reactor trip functions to be improperly scaled in a non-conservative direction. The Delta-T channels were underpredicting full. core power by as much as 5%.

In evaluation which followed the discovery, two potential causes for the Delta-T decrease were identified:

a change in the bot leg temperature streaming patterns which supply coolant samples to the temperature sensors.

a reduction in thermal power, possibly caused by fouling of the feedwater flow venturi ceters.

The Delta-T channel errors have been analyzed to determine the impact on the Final Safety Analysis Report accidents which take credit for the overtemperature and overpower trip functions.

The results of this safety evaluation indicate that only a narrow range of steam line break accidents (0.4 to 0.9 square feet breaks) may have exceeded the design basis during this decreasing Delta-T incident.

Background

Delta-T is used as a measure of reactor power and is calibrated (scaled) to read 100% as determined by the precision secondary system calorimetric measurements.

The scaled value is used in the overtemperature and overpower Delta T core protection circuits.

The overtemperature Delta-T trip actuation circuit loops autocatically vary with:

1) coolant average te=perature, 2) pressurizer pressure, and 3) axial power distribution to protect the reactor core f rom departure f rom nucleate boiling (DNB) during certain design transiests.

The overpower Delta-T trip circuit loops automatically vary with:

1) coolant te=perature, 2) rate of change of average te=perature, and 3) axial power dis-tribution to protect the fuel from excessive heat generation rates (KW/ft).

The overteeperature and overpower Delta-T trip setpoints are detailed in Technical Specification Table 2.2-1.

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I U 8 efuCLSAA stovtatcay consasing g meC Pere 3e4A UCENSEE EVENT REPORT (LER) TEXT CONTINUATION A=ovio owe =n nio a==s a v e f acetsty seassa ns pocast savessem m Lam htaases

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..e w su.,nn Description of Event The results of the evaluation have identified the two possible causes of decreased Delta-T as:

1) changes in the hot leg streaming patterns, and 2) fouling of the feedwater flow venturi nozzles.

The hot leg streaming patterns are results of fuel assembly exit temperature difference.

The high flow rates and differences in individual fuel assembly outlet temperatures cause temperature stratification within the Reactor Coolant system piping. A special device is installed in the hot leg piping which samples the reactor coolant at different areas within the piping. These samples mix and pass by the hot leg temperature sensor to provide an average hot leg temperature.

The fuel assembly exit temperatures did change during the period when the Delta-T was decreasing. This is an indication that the temperature streaming patterns in the piping also changed.

The evaluation states that it is possible f ar a 0.5 degree-F decrease in Delta-T on Unit 2 based on the data provided. This would account for about half of the total decrease.

The venturi nozzle fouling is a result of a crud buildup on the feedvater flow sensing device. The buildup affects the flow readings obtained for the heat balance measure =ents.

The venturi is a device which develops a differential pressure when flow is passed through it.

This differential pressure can be accurately scaled to indicate flow as long as the venturi dimensions are not altered. The crud buildup inside the venturi affects the accuracy of the developed pressure across the venturi.

Venturi fouling has possibly caused a Delta-T decrease of approx 1=ately 0.5 degree-F at McGuire.

The total 1.0 degree-F Delta-T drift caused by hot leg streaming and venturi

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nozzle fouling was within the uncertainty allowances assumed in the safety analysis.

The additional errors associated with not calibrating the circuits during the cycle 2 startup caused the total Delta-T uncertainty to exceed the uncertainty allowances assuued in the safety analysis. The actual error for each loop in units of percent full power was:

Loop A Loop B Loop C Loop D 2.5% F.P.

5.3% F.P 5.5% F.P 3.4% F.P These errors have been evaluated for their impact on certain Final Safety Analysis Report accidents. The evaluation results indicace that only a narrow range of steam line break accidents (0.4 to 0.9 square feet) may have been in violation of the design basis.

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The hot leg streaming patterns have been identified to the industry as early as 1968.

The RID Bypass System, which is installed at McGuire, was developed to reduce the effects of the te=perature. gradients in the piping. Westinghouse reports that no significant proble=s have arisen in the industry pertaining to the streaming patterns although some variations of 0.5 degree-F or less have been reported.

The feedwater f1me venturi nozzles have been reported by several nuclear plants as indicating a fouling condition.

The stations affected by this condition have taken ceasures to compensate for the errors and are monitoring the plant parameters.

The omission of the calibration of the Delta-T channels during power escalation has affected both Unit 1 and Unit 2.

Although the decreasing Delta-T in Unit 2 brought the problem into view, the Unit 1 channels had not been calibrated during the startup of each fuel cycle.

This requirement was not clearly defined to the station personnel.

The channel calibration was being performed on an 18 month basis without regard to the cycle startup requirements.

CORRECTIVE ACTIONS:

I==ediate:

None Subsequent:

1.

The results of the reactor coolant flow test and how it affected the Delta-T circuits have been reviewed.

2.

A work request to recalibrate the Delta-T circuits was initiated.

3.

Westinghouse reviewed the key plant parameters to evaluate the Delta-T drift and submitted a report to Duke Power.

Planned:

1.

A calibration procedure to set the Delta-T to a conservative value prior to reactor secrtup at the beginning of each cycle will be established. Necessary controls will be implerented to ensure the work is completed prior to reactor startup.

2.

A calibration procedure to set the Delta-T values for the new fuel cycle based on final measured Delta-T values will be established. Testing will deter =ine the actual 100% power level.

3.

A program to monitor the full power Delta-T during the entire fuel cycle and to report significant temperature drifts to the Reactor Safety group for analysis will be developed. Data has j

already begun regarding this program.

4.

Measures will be taken to identify the extent of the venturi fouling and initiate actions to compensate for the apparent loss of power. Alternate tests have been initiated to determine the extent of error.

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z wc w amm u nn SAFETY ANALYSIS The evaluation stated that the approximately 1 degree-F drif t associated with the Delta-T channels is within the uncertainty allowances assumed in the safety analysis.

Therefore, the precision heat balance performed in June may be considered valid.

In addition, since the error associated with the calculated flow is within the uncertainty allevance assumed in the safety analyses, there are no flow related Tech Spec violations associated with the Delta-T drift incident. However, when the 1 degree-F drift is added to the errors associated with not calibrating the Delta-T channels during startup, the total error associated with the Delta-T channels is greater than the uncertainty allowances assumed in the safety analyses.

The Delta-T channel errors can potentially impact the FSAR accidents which take credit for a reactor trip on the overte=perature or overpower Delta-T trip functions. There is sufficient margin included in the overtemperature Delta-T setpoint calculation to account for the channel errors.

Therefore, the accident analyses which take credit for a reactor trip on overtemperature Delta-T (RCCA Withdrawal at Power, RCCA Misalignment, and Boron Dilution) remain valid.

However, for the overpower trip function, there is not suf ficient margin in the setpoint calculation to account for the channel errors.

It should be noted that credit is not explicitly taken for the overpower Delta-T trip function in any of the accidents analyzed in Chapter 15 of the McGuire FSAR.

However, WCAP-9226. "Reactor Core Response to Excessive Secondary Steam Releases",

shows that the overpower Delta-T trip function may be relied upon to provide DNB protection for so=e steam line breaks at power.

Plant-specific analyses have been performed for a spectrum of intermediate steam line breaks at power.

It was determined that, for break sizes equal to or greater than 0.9 square feet, reactor trip occurred on SI actuation on low steam line pressure. For break sizes less than.0.4 square feet, either no trip was necessary, or a trip occurred on low-low steam generator level. For breaks inside containment in the size range of interest, reactor trip occurs on SI actuation on high containment pressure.

Thus, the overpower Delta-T trip function provides the reactor trip only for breaks between 0.4 square feet and 0.9 square feet outside containment.

There were no pipe break events during the period which would have affected the health and safety of the public.

It is also possible that other trip functions such as the high flux trip, or trading off available analytical margin may have revealed acceptable results for the narrow range of affected steam line breaks.

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