Text

Proposed Tech Spec Changes to Allow Operation W/Slightly Positive Moderator Temp Coefficient at Low Power Levels & Increased Enthalpy Hot Channel Factor (F Delta H) Limit Below Full Power
	ML20085L055
Person / Time
Site:	Farley
Issue date:	10/13/1983
From:	; ALABAMA POWER CO.
To:	;
Shared Package
ML20085L052	List: ML20085L049; ML20085L055;
References
	NUDOCS 8310210261
	Download: ML20085L055 (13)
	v • d • e

~

l ATTACIDENT 2 8310210261 831013 PDR ADOCK 05000348 P PDR

665-660-Unacceptable 655 Operation 400 psia 650 645 2250 psia 640 635 650 2000 psia 6 5<

h620

- 1875 psia a 615

610 605 600 595 Acceptable \

590 Operation 585 580 575 570- -

565 .8 1. 1.1 1.2

9. .I .2 .5 4 .5 .6 .7 ;9 POWER (fraction of nominell Figure 2.1-1 Reactor Core Safety Limit Three Loops in Operation Applicability: < 5% Steam Generator Tube Plugging l

FARLEY UNIT I 2-2 i

TABLE 2.2-1 (Centinu d) h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS -

A

? NOTATION E

NOTE 1: Overtemperature AT I AT, [K i-K2 +IS (T-T')g(P-P')-fi (AI)]

1+T S 2

where: AT, = Indicated AT at RATED THERMAL POWER T = Average temperature, F T' 1 577.2 F (Maximum Reference Tavg at RATED THERMAL POWER)

P = Pressurizer pressure, psig P' = 2235 psig (Nominal RCS operating pressure)

[ 'l = The function generated by the lead-lag controller for T avg dynamic com,nsation 1+T 3 2

T1 &T2 = Time constants utilized in the lead-lag controller for T vg a T1 = 30 secs, S = Laplace transform operator, sec 1 Operation with 3 Loops Operation with 2 Loops

, K1

= 1.22 K1 = (values blank pending i

j K2

= 0.0154 K2 = NRC approval of K3

= 0.000635 K3 = 2 loop operation)

! and f1 (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear i

ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

m TABLE 2.2-1 (Continued)

D -

REACTOR TRIP SYSTEh INSTRUMENTATION TRIP SETPOINTS Q

, NOTATION continued E

(i) for qt - 9b between -35 percent and +9 percent, ft (AI) = 0 (where qt and qb are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and qt = 9b is total THERMAL POWER in percent of RATED THERMAL POWER).

(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: [Ky -K T3 S Overpower AT 1 AT, 5 -K6 (T-T") -f2 (OI)3 1+T33 where: AT, = Indicated AT at RATED THERMAL POWER T = Average temperature, F T"= Reference Tavg at RATED THERMAL POWER (Calibration temperature for AT instrumentation, f 577.2 F)

K4= 1.08 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 f T" T 3 3

1+T 3

=

The function generated by the rate lag controller for T vga dynamic compensation 3

2.1 SAFETY LIMITS BASES

=

---=== u =_ =_- -

2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which wold result in the release of fission products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB through the 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 cperational 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.

The curves of Figuras 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average tempercture 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.

These curves are based on an enthalpy hot 3 , factor, channel of 1.55 and a F"$110wance i reference 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 ~

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 1 -

B 2-1

F 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 O delta k/k/ F at or above 70% THERMAL 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#

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

ACTION:

b. 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 to less positive than 0 delta k/k/ F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Keff greater than or equal to 1.0

See Special Test Exception 3.10.3 FARLEY-UNIT 1 3/4 1-4

m POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR H0T CHANNEL FACTOR - FAH LIMITING CONDITION FOR OPERATION

==_

3.2.3 F shall be limited by the following relationship:

H N

F 1.55 [1 + 0.3 (1-P)] [1-RBP(BU)]

THERMAL POWER , and where P = RATED THERMAL POWER RPB(BU) = Rod Bow Penalty at 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).

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to <-

55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , N

b. Demonstrate through in-core mapping that 3 F g is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after attaining 95% or greater RATED THERMAL POWER.

FARLEY-UNIT 1 3/4 2-8

665-Unacceptable 660- \

655 Operation 400 psia 650 645<

640, 2250 psia 655 650-625 h620 1875 psia 615 -

610 605 600<

595 Acceptable 590 Operation 585 580 575 570- .

565

0. .I .2 .5 4 .5 .6 .7 .0 '9 1. 1.1 1.2 POWER (fraction or nominell Figure 2.1-1 Reactor Core Safety Limit Three Loops in Operation Applicab.ility: < 5% Steam Generator Tube Plugging FARLEY UNIT 2 2-2

a TABLE 2.2-1 (Continued)

S .

jE REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS O NOTATION E

U NOTE 1: Overtemperature AT I AT, [K i-K2 1+*1b (1-T')+y(P-P')-fi (AI)]

N 1+T S 2

where: AT, = Indicated AT at RATED THERMAL POWER T = Average temperature, F T' 1 577.2 F (Maximum Reference Tavg at RATED THERMAL POWER)

P = Pressurizer pressure, psig P' = 2235 psig (Nominal RCS operating pressure) 1+T S m 1 = The function generated by the lead-lag controller for Tavg dynamic compensation

& 1+T S 2

T y&T2 = Time constants utilized in the lead-lag controller for T avg T1 = 30 secs, S = Laplace transform operator, sec-1 Operation with 3 Loops Operation with 2 Loops K1

= 1.22 Ki = (values blank pending K2

= 0.0154 K2 = NRC approval of K3

= 0.000635 K3 = 2 loop operation) and f t (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:

TABLE 2.2-1 (C:ntinuid) s REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i2 Q NOTATION continued C

3

(i) for qt - Ab between -35 percent and +9 percent, fg (AI) = 0 (where qt and qb are percent RATED THERMAL POWER in the top and bottom

" halves of the core respectively, and qt

4b is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (qt - 4b) 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 - 4b) exceeds +9 percent, the AT trip setpoint shall be automatically reduced by 1.60 percent of its value at RATED THERMAL POWER.

T3 b m Note 2: Overpower AT 1 AT, [Ky -K 5 e 1+t3b -K6 (T-T") -f2 (AI)3 where: AT, = Indicated AT at RATED THERMAL POWER T = Average temperature, F T"= Reference T avg at RATED THERMAL POWER (Calibration temperature for AT instrumentation,1577.2 F)

K4= 1.08 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" T

3 = The function generated by the rate lag controller for Tavg dynamic compensation 1+r33

T

,a .

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 wold 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 nave been related to DNB through the W-3 correlation. The W-3 DNB correlation has becn 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.

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 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.

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

AH N

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 2 B 2-1 'I

f REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE COEFFICIENT

' LIMITING CONDITION FOR OPERATION

=

4 3.1.1.3 The moderator temper [ture 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. .

b. Less negative than -3.9 x 10-4 delta k/k/*F for the all rods withdrawn, g end-of cycle life (EOL), 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:

b. 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 to less positive than 0 delta k/k/ F within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in H0T STANDBY within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . 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 cubmitted 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 H0T SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

With Keff greater than or equal to 1.0

See Special Test Exception 3.10.3 FARLEY-UNIT 2 3/4 1-4 h

e .

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR HOT CHANNEL FACTOR - F AH LIMITING CONDITION FOR OPERATION f

3.2.3 F H

shall be limited by the following relationship:

Fh 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 fores).

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 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip Setpoints to I 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , N

b. Demonstrate through in-core mapping that FAH is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , 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, 3 is demonstrated through in-core' mapping to be within its limit at aH 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 nours after attaining 95% or greater RATED THERMAL POWER.

FARLEY-UNIT 2 3/4 2-8

,

ML20085L055

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