Text

Amends 37 & 27 to Licenses NPF-2 & NPF-8,respectively, Allowing Operation W/Slightly Positive Moderator Temp Coefficient at Low Power Levels & Increased Hot Channel Factor Limit Below Full Power
	ML20083H023
Person / Time
Site:	Farley
Issue date:	12/30/1983
From:	Varga S; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML20083H030	List: ML20083H023; ML20083H033;
References
	TAC-52565, TAC-52566, NUDOCS 8401120597
	Download: ML20083H023 (18)
	v • d • e

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UNITED STATES E "f

, j *g NUCLEAR REGULATORY COMMISSION C

WASHINGTON, D. C. 20555 7.h

/j/

o, ALABAMA POWER COMPANY DOCKET NO. 50-348 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT N0. 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 37 License,No. NPF-2 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Alabama Power Company (the licensee) dated October 13, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission; C.

There is reasonable assurance:.(1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Conaission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-2 is hereby amended to read as follows:

%Olk$b O 00 p P

2-(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 37, are hereby incorporated in the 1.icense. The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

However, the changes to the Technical Specifications will be effective prior to entry into Mode 2 following the fifth refueling outage scheduled to start January 10, 1984.

FOR THE NUCLEAR REGULATORY COMMISSION

/^

a-

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y even

. Varga, ef Operating Reactors ranch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December 30, 1983 6a O

ATTACHMENT TO LICENSE AMENDMEllT NO. 37 FACILITY OPERATING LICENSE NO. NPF-2 DOCKET NO. 50-348 Revised Appendix A as follows:

Remove Insert 2-2 2-2 2-8 2-8 2-9 2-9 B2-1 B2-1

.3/4 1-4 3/4 1-4 3/4 2-8 3/4 2-8 9

9 4

.m_

l 665<

Unacceptable 660 Operation 655 400 psia 656.,

645<-

2250 psia 640 655-638 M psia

,23, g 629 5 psia 5615<

610 605 689.

595 Acceptable Operation 59e <-

585 589 575 578 5

6.

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1.1 1.2 555 POWER Ifr act.1on or nomine1 L

Figure 2.1-1 Reactor Core Safety Limit Thrae Loops iti Operation s

Applicability:- < 5% Steam Generator Tube k

Plugging s

FARLEY UNIT 1 Amendment No. 37 l

TABLE 2.2-1 (Centinued) nE, REACTOR TRIP SYSTEM INSTRUENTATION TRIP SETPOINTS W

F

?

NOTATION I+*1S (T-T')+K (P-P') -fg (AI)]

NOTE 1:

Overtemperature aT f AT, [K -K2 i

3 1+T S 2

~~

where:

AT, = Indicated AT at RATED THERMAL POWER T = Average temperature

F l

at RATED THERMAL POWER)

T' f 577.2'F (Maximum Reference Tavg P = Pressurizer pressure, psig P' = 2235 psig (Nominal,RCS operating pressure)

[

'l

= The function generated by the lead-lag controller for Tavg dynamic compensation 1+T S 2

avg T1 = 30 secs, y&T2 = Time constants utilized in the lead-lag controller for T T

S = Laplace transform operator, sec I.

y Operation with 3 Loops Operation with 2 Loops Ki = (values blank pending K1 1.22

=

K2 0.0154 K2 = NRC approval of o

=

K3 0.000635 K3 = 2 loop operation)

=

and f1 (61) is a function of the indicated difference between top and hattom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests :uch that:

l

.A

^

TABLE 2.2-1 (Continued)

S#

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS rn NOTATION continued i

E (i) for qt - 9b between -35 percent and +9 percent, f1 (al) = 0 (where qt ano qb are percent RATED THERMAL POWER in the top and bottom

~

halves of the core respecti~vely, and qt = 4b is total THERMAL PCWER 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 - Ab) exceeds +9 percent, the AT trip setpoint shall be automatically reduced by 1.60 percent of its value at RATED THERMAL POWER.

S Note 2:

Overpower AT 1 AT,

[Ky -K

'-K6 (T-T") -f2 IAI)3 5

1+'3b where:

AT, = Indicated AT at RATED THERMAL POWER Average temperature,*F T

=

T"=

Reference T at RATED THERMAL POWER (Calibration temperature for avg AT instrumentation, i 577.2"F)

K4=

1.08 KS= 0.02/"F for increasing average temperature and 0 for decreasing average temperature g

E K6= 0.00109/"F for T > T"; K6 = 0 for T 1 T"

=

S

'3 The function generated by the rate lag controller for T vg dynamic compensation a

=

Ef 1+1 3 3

ti

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. DH3 is not a directly measurable parameter during operation and therefore THERMAL f0WER 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 he'at 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"$110wance is

, of 1.55 and a 3

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

H 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 witndrawn 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 l

REAChlVITYCONTROLSYSTEMS MODERATOR TEMPERATURE COEFFICIENT LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

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

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 POWER.

b.

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

APPLICABILITY:

Specification 3.1.1.3.a - MODES I 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 HOT 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 .

0With Keff greater than or equal to 1.0

See Special Test Exception 3.10.3 FARLEY-UNIT 1 3/4 1-4 Amendment No. 37 1

i POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR HOT CHANNEL FACTOR - FaH LIMITING CONDITION FOR OPERATION 3.2.3 F shall be limited by the following relationship:

H F h 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).

APPLICABILITY: MODE 1 ACTION:

With F H exceeding its limit:

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 2 a.

hours 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 is within its limit within b.

Demonstrate through in-core mapping that FaH 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 Identify and correct the cause of the out of limit condition prior to c.

above;subsequentPOWEROPERATIONmayproceedprovidedthatFgorb, increasing THERMAL POWER above the reduced limit required by is 3H demonstrated through in-core mapping to be within its limit at 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 cttaining 95% or greater RATED THERMAL ~ POWER.

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

37 I

neroq%

UNITED STATES

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NUCLEAR REGULATORY COMMISSION n

,E WASHINGTON, D. C. 20555 ALABAMA POWER COMPANY DOCKET NO. 50-364 JOSEPH M. FARLEY NUCLEAR PLANT, UNIT N0. 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 27 License No. NPF-8 1.

The Nuclear Regulatory Commission (the Commission) has found that:

A.

The application for amendment by Alabama Power Company (the licensee) dated October 13, 1983, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.

The facility'will operate in conformity with the application, as amended, the provisions of the Act, and the regulations of the Commission; C.- There is reasonable assurance: (i) that the activities authorized by this' amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this license amendment will not be inimical to the common defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requi en:ents have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical

-Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(~2) of Facility Operating License No. NPF-8 is hereby

-amended to read as.follows:

t

i 1 ;

(2) Technical Specifications The Technical Specifications contained in Appendices A ard B, as revised through Amendment No. 27., are hereby incorporated in the license.

The licensee shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance.

However, the changes to the Technical Specifications will be effective prior to entry into flode 2 following the third refueling outage scheduled to start December 4,1984.

FOR THE NUCLEAR REGULATORY COMMISSION bVM Steven arga, Ch e Operating Reactors Branch #1 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: December 30, 1983

WlI

~

-ATTACHMENT TO' LICENSE AMENDMENT NO. 27 FACILITY GPERATING LICENSE NO. NPF-8 DOCKET N0. 50-364 Revised Appendix A as follows:

Remove Insert 2-2 2-2 2-8 2 2-9 2-9 B2-1 B2-1 3/4 1-4 3/4 1-4 3/4 2-8 3/4-2-8 N

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4 665-660 Unacceptable 655-Operation 400 psia 650 645 2250 psia 640 655<

630 2000 psia

, g, h620 1875 psia 615 610 605 600 Acceptable 595 590 Operation 585-580-575<

570 565 9.

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1.

1.1 1.2 POWER treaction or nominall Figure 2.1-1 Reactor Core Safety Limit Three Loops in Operation Applicab.ility:

< 5% Steam Generator Tube Plugging FARLEY UNIT 2 Amendment No.

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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 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 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 eperating 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.

Thesecurvesarebasedonanenthalpyhotchannelfactor,F"$110wanceis

, of 1.55 and a 3

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

=1.5g[1+0.3(1-P)]

g 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 8 2-1 Amendment No. 27

l REACTIVITY CONTROL SYSTEMS MODERATOR TEMPERATURE C0 EFFICIENT -

LIMITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (MTC) shall be:

Lessuthan or equal to 0.5 x 10-4 delta k/k/ F for the all rods a.

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, 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:

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 withir, 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 2 3/4 1-4 Amendment No. 27

k POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR HOT CHANNEL FACTOR - FAH LIMITING CONDITION FOR OPERATION N

3.2.3 F shall be limited by the following relationship:

3H 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 cores).

APPLICABILITY: MODE 1 ACTION:

s With F g exceedin'g 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 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, is 3H demonstrated through in-core mapping to be within its limit at a nominal 3

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 2 3/4 2-8 Amendment No. 27

_ _ _ - - _ _ - _ _ _. _ _ _ _ _ _ _ _

v t e Letter Sequence Other
TAC:52565, (Approved, Closed) TAC:52566, (Approved, Closed)
Initiation Request, Request Acceptance... Administration Questions Answers Results Approval Other: ML20083H023, ML20087A383, ML20087A402, ML20087K569, ML20087K815
MONTHYEAR1984-03-02 [Table View]

ML20083H023

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