ML20087K569

Corrected Proposed Tech Specs,Incorporating Moderator Temp Coefficient at Low Power Levels & Increased Enthalpy Hot Channel Factor Limit Below Full Power

ML20087K569
Person / Time
Site:	Farley
Issue date:	03/22/1984
From:	ALABAMA POWER CO.
To:
Shared Package
ML20087K546	List: ML20087K544 ML20087K569
References
TAC-52556, TAC-52565, TAC-52566, NUDOCS 8403260234
Download: ML20087K569 (10)

Text

'

665<

• Unacceptable 660 Operation 655-400 psia 650 <,

645<

2250 psia 640 655<

P650-2000 psia

,625<

$620- ps a 615, cn6104 -

o

" SOS- I 688-Acceptable 595 .

Operation 59e .

585 .

588 <

~

575 <-

570 .

565 .7 .9 :9 1. 1.! 1.2

9. .1 .2 .5 4 .5 .6 FRACTION OF RATED THERMAL POWER Figure 2.1-1 Reactor Core Safety Limit l Three Loops in Operation l

Applicability: 5 57 Steam Generator Tube Plugging l

FARLEY UNIT 1 Amendment No. 37 8403260234 840322 2-2 Corrected Page PDR ADOCK 05000348 P._ _ _. . . . . - - _ _ _ -

PDR _ - , _ - _ _ _

9 m '

TABLE 2.2-1 (Centinund) j- -REACTOR TRIP SYSTEM ~ INSTRUMENTATION TRIP SETPOINTS .

' NOTATION ' continued p .-(i) for qt - 9b between -35 percent and +9 percent ft (AI) = 0 (where qt and qb are percent .l RATED-THERMAL POWER in the top and bottom halves of the core respectively, and qt~ +.9b i5 .

total THERMAL' POWER in percent of RATED THERMAL POWER).

(ii) for shall each be 'percent automatically that.the magnitude reduced of (qt by 1.37 - Ab)f oits value at RATED THERMAL POWER. exceeds.l -35 perce percent (iii) for shall each percent that the be automatically magnitude reduced of.(qt by'1.60 - 9b)foits value at RATED THERMAL POWER. exceeds +9 perce percent Note 2: Overpower AT 1 ATo [K4 -K5 ]3S T-K6 (T-T") -f2 (AI)3 1+T36 where: ATo = Indicated AT at RATED THERMAL POWER T = Average temperature, *F T" = Reference T vg a at RATED THERMAL POWER (Calibration temperature for AT instrumentation,1577.2*F)

K4 = 1.08 l K5 = 0.02/*F for increasing average temperature and 0 for decreasing average temperature K6 = 0.00109/*F for T > T"; K6 = 0 for T 1 T" l pg j3s s The function generated by the rate lag controller for T vga dynamic compensation n:

8 it te

a. e g.

O

= . a REACTIVITY CONTROL = SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The. moderator temperature coefficient (MTC) shall be:

a. ,Less than or equal to'0.5 x 10 4 delta k/k/*F for the all rods

- withdrawn, beginning of cycle life (BOL), below 70 % THERMAL POWER condition. Less than or equal to 0 delta k/k/ F at or above 70% THERMAL POWER.

i

b. Less negative than -3.9 x 10-4 delta k/k/ F for the all rods withdrawn, end of cycle life (E0L), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3.a - MODES 1 and 2* only#

Specification 3.1.1.3.b - MODES 1, 2 and 3 only#

ACTION:

aE With the MTC more positive than the limit of 3.1.1.3.a above, operation ,

in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and maintained I sufficient to restore the MTC to within its limit l within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. 'These withdrawal limits shall be in addition to the insertion limits of

, Specification 3.1.3.~6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

l 3. -In lieu of any other report required by Specification 6.9.1, a Special Report is prepared and submitted to the Commission pursuant j to Specification 6.9.2 within 10 days, describing the value of the

measured MTC, the interim control rod withdrawal limits and the

predicted average core burnup necessary for restoring the positive MTC'to within its limit for the all rods withdrawn condition.

b. With the MTC more negative than the limit of 3.1.1.3.b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

t

• With Keff greater than or equal to 1.0

1. See Special Test Exception 3.10.3 FARLEY-UNIT 1 3/4 1-4 Amendment No. 37 Corrected Page m- -qr y w- g y -+,- --.v- . - - - .

p, - , -w1 --,.s ,y, ,y,. ..3,-,, -,m,,--- ..,y--,v.,.r. w.,.., ,,,m.,--,..--me-- , . -,_,-% ------.,w-, ,e - - , - -* --c-.--e> + - - -

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - FfH LIMITING CONDITION FOR OPERATION

. ...=

3.2.3 F 3g shall be limited by the following relationship:

N F

3H < 1.55 [1 + 0.3 (1-P)] [1-RBP(BU)]

THERMAL POWER , and where P = RATED THERMAL POWER RPB(BU) = Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-3, where a region is defined as those asserrblies with the same loading date (reloads) or enrictment (first cores).

APPLICABILITY: MODE 1 -

ACTION:

With F H exceeding its limit:

a. Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and reduce the Power Range Neutron Flux-High Trip Setpoints 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 />, N

b. Demonstrate through in-core mapping that FAH is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit 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 />, and

c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by above;subsequentPOWEROPERATIONmayproceedprovidedthatFgorb, is demonstrated through in-core mapping to be within its limit at 3H a nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL POWER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

FARLEY-UNIT 1 3/4 2-8 Amendment No. 37 Corrected Page

i i 2.1 SAFETY LIMITS B SES -

2.1.1 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 hrat 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 4

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 W-3 correlation. The W-3 DNB correlation has been d:Eveloped 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 minimum value of the DNBR during steady state operation, normal

. cperational transients, and anticipated transients is limited to 1.30. This '

I value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average tempe,rature for which the sinimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is

cqual to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel 3 , factor, of 1.55 and a F"kilowance is raference cosine with a peak of 1.55 for axial power shape. An included for an increase in F N at reduced power based on the expression:

AH F = 1.55 [1+0.3 (1-P)]

where P is the fraction of RATED THERMAL POWER 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

FARLEY-UNIT 1 B 2-1 Amendment No. 37 Corrected Page

665<

660- Unacceptable 655< Operation 400 psia 650<

645<

640 2250 psia 655-

- 650-2000 psia

!"625-f620- psia l

g615

' 610 m

g605-l 600<

595 Acceptable 590 Operation 585<

l 580<

575<

570-565 1.1 1.2

9. .I .2 .5 4 .5 .6 .7 .9 9 1.

FRACTION OF RATED THERMAL POWER Figure 2.1-1 Reactor Core Safety Limit Three Loops in Operation

Applicability: <_ 5% Steam Generator Tube Plugging FARLEY UNIT 2 Amendment No. 27 Corrected Page 2-2

lABLE 2.2-1 (Centinuid) -

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS '.

N

< NOTATION continued (i) for qt - 9b between -35 percent and +9 percent, ft (aI) = 0 (where qt and qb are percent l

$ RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt + 9b is e total' THERMAL POWER in percent of RATED THERMAL POWER).

w (ii) for each percent that the magnitude of (qt - 9b) exceeds -35 percent, the AT trip setpoint shall be automatically reduced by 1.37 percent of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (qt - 9b) exceeds +9 percent, the AT trip setpoint shall be automatically reduced by 1.60 percent of its value at RATED THERMAL POWER.

Note 2: Overpower AT < ATo [K4 -K5 J3S T-K6 (T-T") -f2 (AI)3 1+t3S where: ATo = Indicated AT at RATED THERMAL POWER T = Average temperature *F T" = Reference T vga at RATED THERMAL POWER (Calibration temperature for AT i nstrumentati on ,1- 577.2*F)

K4 = 1.08 l K5 = 0.02/*F for increasing average temperature and 0 for decreasing average temperature K6 = 0.00109/*F for T > T"; K6 = 0 for T 1 T" l 0E . S 3[ J'3 g = The function generated by the rate lag controller for T vga dynamic compensation as ELE 28 2-"

d

~

2.1 SAFETY LIMITS BASES 2.1.1 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 ralated to DNB through the W-3 correlation. The W-3 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 minimum value of the DNBR during steady state operation, normal operational transients, and anticipated transients is limited to 1.30. This value corresponds to a 95 percent probability at a 95 percent confidence level that DNB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

l The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor C9olant System pressure and average temperature for which the minimum DNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

, of 1.55 and a These curves are based reference cosine with a peak of 1.55 for axial power shape. An on an enthalpy hot channel 3 factor, F"Nilowance is included for an increase in F N at reduced power based on the expression:

AH F (

tH "

where P is the fraction of RATED THERMAL POWER 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 FARLEY-UNIT 2 B 2-1 ge"d*Ndpah rre N

-,,-,.i-,.---,,-~.--y- + - ,,,,,,-m-- -w- -

w ry.-,.,,%--,,,,y,..-,.,a .,,,,---,y.m__,-e,,#.,m,,7--, _ , , , . - - - - - - . , - - - -,-.w,.v-... .n--wy-w,m-

' REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

a. Less than or equal to 0.5 x 10-4 delta k/k/*F for the all rods withdrawn, beginning of cycle life (B0L), below 70 % THERMAL POWER

, condition. Less than or equal to 0 delta k/k/'F at or above 70% THERMAL l POWER.

b. Less negative than -3.9 x 10-4 delta k/k/*F for the all rods withdrawn,

. end of cycle life (E0L), RATED THERMAL POWER condition.

APPLICABILITY: Specification 3.1.1.3.a - MODES 1 and 2* only#

Specification 3.1.1.3.b - MODES 1, 2 and 3 only#

ACTION:

a. With the MTC more positive than the limit of 3.1.1.3.a above, operation

! in MODES 1 and 2 may proceed provided:

1. Control rod withdrawal limits are established and. maintained sufficient to restore the MTC within its limit l within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. These

! withdrawal limits shall be in addition to the insertion limits of

! Specification 3.1.3.6.

2. The control rods are maintained within the withdrawal limits established above until a subsequent calculation verifies that the MTC has been restored to within its limit for the all rods withdrawn condition.

3. In lieu of any other report required by Specification 6.9.1, a Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days, describing the value of the measured MTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive MTC to within its limit for the all rods withdrawn condition.

! b. With the MTC more negative than the limit of 3.1.1.3.b above, be in HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

l l 'With Keff greater than or equal to 1.0

1. See Special Test Exception 3.10.3 FARLEY-UNIT 2 3/4 1-4 Amendment No. 27 Corrected Page l

l l -. -.- .".. .. _-_-

,' POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR F"H LIMITING CONDITION FOR OPERATION N

3.2.3' F 3H shall be limited by the following relationship:

N F

3H $1.55 [1 + 0.3 (1-P)] [1-RBP(BU)]

THERMAL POWER ,and where P = RATED THERMAL POWER RPB(BU) = Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-3, where a region is defined as those assemblies with the same loading date (reloads) or enrichment (first cores).

APPLICA'BILITY: MODE 1 ACTION:

With F H exceeding its limit:

a. Reduce THERMAL POWER to less than 50%'of RATED THERMAL POWER within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and -reduce the Power Range Neutron Flux-High Trip Setpoints 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 />, N

b. Demonstrate through in-core mapping that FAH is within its limit within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after exceeding the limit 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 />, and l c. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER above the reduced limit required by above;subsequentPOWEROPERATIONmayproceedprovidedthatFgorb, is demonstrated through in-core mapping to be within its limit at 3Na nominal 50% of RATED THERMAL POWER prior to exceeding this THERMAL POWER, at a nominal 75% of RATED THERMAL P0.WER prior to exceeding this THERMAL power and within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

FARLEY-UNIT 2 3/4 2-8 Amendment No. 27 Corrected Page

_ . _ _ _ _ - . . _ _ _ _ __