ML20117H472

Amends 121 & 113 to Licenses NPF-2 & NPF-8,respectively, Revising TS Contained in Apps a & B Re Reactor Core Safety Limits:Setpoints Notation,Power Distribution Limits, Surveillance Requirements & TS Bases

ML20117H472
Person / Time
Site:	Farley
Issue date:	09/03/1996
From:	Berkow H NRC (Affiliation Not Assigned)
To:
Shared Package
ML20117H474	List: ML20117H472 ML20117H486
References
NUDOCS 9609090289
Download: ML20117H472 (48)

Text

-

j u truq a

j

[

k UNITED STATES j

NUCLEAR REGULATORY COMMISSION 2

WASHINGTON, D.C. 20066 4001

%*****/

SOUTHERN NUCLEAR OPERATING COMPANY. INC.

ALABAMA POWER COMPANY DOCKET NO. 50-348 l

JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 121 License No. NPF-2 1.

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

]

A.

The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated June 12, 1996, 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, the provisions of the Act, and the rules and regulations of the Comission; 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 comon 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 Comission's regulations and all applicable requirements have been satisfied.

j 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:

I 9609090289 960903 PDR ADOCK 05000348 P

PDR A

x s!

j

,!' (2)

Technical Specifications The Technical Specifications contained in Appendices A and B, as I

revised through Amendment No.121, are hereby incorporated in the license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

FOR THE NUCLEAR REGULATORY COMMISSION l

H rbert N. Berk w, Director Project Directorate 11-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 3, 1996

)

L

1 j

ATTACHMENT TO LICENSE AMENDMENT NO.121 1

TO FACILITY OPERATING LICENSE NO. NPF-2 l

DOCKET NO. 50-348

)

Replace the following pages of the Appendix A Technical Specifications with l

the enclosed pages. The revised areas are indicated by marginal lines.

Remove Paaes Insert Paaes XIX XIX 2-2 2-2 i

2-8 2-8 2-9 2-9 j

2-10 2-10 B 2-2 8 2-2 B 2-5 8 2-5 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 j

3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 4

3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-8 3/4 2-8 a

3/4 7-2 3/4 7-2 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 e

l 8 3/4 2-3 8 3/4 2-3 B 3/4 2-4 B 3/4 2-4 t

B 3/4 2-5 B 3/4 2-5 6-19 6-19 1

1 t

1 i

1 i

i 1

l 1

0

INDEX ADMINISTRATIVE CONTROLS SECTION PApag Review...................................................

6-10 Audits...................................................

6-11 i

Authority..............'..................................

6-12 Records..................................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 Records..................................................

6-13 6.6 REPORTABLE EVENT ACTION.....................................

6-14 6.7 SAFETY LIMIT VIOLATION..................................... 6-14 6.8 PROCEDURES AND PROGRAMS.....................................

6-14 6.9 REPORTING REQUIREMENTS o

6.9.1 ROUTINE REPORTS Startup Report

.......................................... 6-15a Annual Report............................................

6-16 Annual Radiological Environmental Operating Report.......

6-17 Annual Radioactive Effluent Release Report...............

6-17 Monthly Operating Report.................................

6-19 Peaking Factor Limit Report..............................

6-19 l

Annual Diesel Generator Reliability Data Report..........

6-19 Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Sealed Source Leakage Report......................

6-20 6.9.2 SPECIAL REPORTS...........................................

6-20 6.10 RE CO RD RE TE NT ION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM................................

6-21a 6.12 HIGH RADIATION AREA......................................... 6-22 FARLEY-UNIT 1 XIX AMENDMENT NO.

121

t 680

~

UNACCEPTABLE 660 2440 psia OPERATION 2250 psia 640

\\

2000 psia LL en620

~

1840 psia 600 ACCEPTABLE OPERATION 580 l

l 560 0.0 0.2 0.4 0.6 0.8 1.0 1.2 l

POWER (FRACTION OF RATED THERMAL POWER)

Figure 2.1-1 Reactor Core Safety Limits I

Three Loops in Operation 1

FARLEY-UNIT 1 2-2 AMENDMENT NO.121 l

i o

TABLE 2-2-1 fContinued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS M

NOTATION N

Note 1: Overtemperature AT s

~

w r

3

~

r + r*s 3

r + r's 1

1 1

s T

- T' + K (P-P')-f;(AI)

H AT 5; AT, K, - K 2

3 (1 + r s; sl + r s; (1 + r s>

s 2

where AT Measured AT by RTIL instrumentation;

=

ATO Indicated AT at RATED THERMAL POWER and reference Tavg;

=

T Average temperature, *F;

=

T' Reference T at RATED THERMAL POWER (s 577.2*F);

=

ayg Pressurizer pressure, psig; P =

P' 2235 peig (nominal RCS operating pressure);

=

1+ta1 y,,s The function generated by the lead-lag controller for T,yg dynamic compensation; w

=

a t1 &T2 Time constants utilized in the lead-lag controller for T,yg, 23 = 30 sec,

=

12 = 4 sec; 1+ts 1+ts The function generated by the lead-lag controller for AT dynamic compensation;

=

T4 015 Time constants utilized in the lead-lag controller for AT, t4 = 0 sec, 25 's 6 sec;

=

1+78

• 9 6

T6 Time constant utilized in the measured T lag compensator, t6 s 6 sec; l

=

ayg Laplace transform operator, sec-1; s

=

Operation with 3 loops operation with 2 loops Ki = 1.17; Ki = (values blank pending l

=0 K2 = 0.017; K2 = NRC approval of l

N K3 = 0.000825; K3 = 2 loop operation) l

m m..

.m

...___...._..._.._.m

. _ _ ~~.

.m.

TABLE 2.2-1 (Continued)

BEACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

• n NOTATION fContinued)

E:

and ft (AI) f.a a function of the inuicated difference between top and bottom detectors of the power-range d

nuclear ion chembers; with gains to be selected based on measured instrument response during plant startup H

tests such thatt e

(1) for qt - Ob between -23 percent and +15 percent, ft (AI) = 0 (where gt and qb are percent RATED l

THERMAL MNER in the top and bottom halves of the core respectively, and gt + 9b is total THERMAL POWER 'n percent of RATED THERMAL POWER);

(ii) for each percent that the magnitude of (qt 9b) exceeds -23 percent, the AT trip setpoint shall be automatically reduced by 2.48 percent of its value at RATED THERMAL POWER; and I

(iii) for each percent that the magnitude of (qt - 9b) exceeds +15 percent, the AT trip setpoint shall I

be automatically reduced by 2.05 percent of its value at RATED THERMAL POWER.

Note 2:

Overpower AT to a

e l + r*s' r's 1

l AT s AT. K-K T

- Kg T

- T"

- f (AI) 3 2

s 1 + r s; (1 + r s; (1 + r s; (1 + r s; s

3 where: AT = Measured AT by RTD instrumentation; ATO Indicated AT at RATED THERMAL POWER and reference Tavg; l

=

T Average temperature, "F;

=

Reference T at RATED THERMAL POWER ($ 577.2*F);

f T"

=

ayg g

K4

= 1.10; l

=l K5 0.02/*F for increasing average temperature and 0 for decreasing average

=

2 temperature; 8

K6 0.00109/*F for T > T*,

K6 = 0 for T s 'l";

l

=

8

• 3 The function generated by the rate 1a, control 1er for T dynamic compensation,

=

1.,,,

avg w

. r.

,w 3-.

,y..

,c

TABLE 2.2-1 (Continued)

l REACTOR TRIP SYSTEM INSTRUMENTATION T;IP SETPOINTS y

NOTATION fContinued1 E

Time constant utilized in the rate lag controller for T,yg, t3 = 10 see; T3

=

1 + r4s The function generated by the lead-lag controller for AT dynamic compensation; I + r$3 T4 ET5

= Time constants utilized in the lead-lag controller for AT, T4 = 0 sec, t5 s 6 see; l

= Lag compensator on measured T,yg;

= Time constant utilized in the measured T,yg lag compensator, is s 6 see; l

t6

-1; u

a = Laplace transform operator, sec d

f2(AI)

= 0 for all AI.

Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 0.4 percent AT span.

Note 44 Pressure value to be determined during initial startup testing.

Pressure value of 5 55 psia to be used prior to determination of revised value.

Note 5: Pressure value to be determined during initial startup testing.

Note 6: The channel's maximum trip point shall not exceed its computed trip point by more than 0.4 percent AT span, a

R 5

=

s O

~

1 SAFETY LIMITS BASES I

The curves of Figures 2.1-1 and 2.1-2 are based on the most limiting result using en enthalpy hot channel factor, F$H,of1.70for VANTAGE 5fuelandanF$Hof1.30forLOPARfuelandareferencecosinewitha l peak of 1.55 for axial power shape. An allowance is included for an increase inF$Hatreducedpowerbasedontheexpression:

F$H=1.70[1+0.3(1-P)) for VANTAGE 5 fuel and F$H=1.30[1+0.3 (1 - P)] for LOPAR fuel l

where P is the fraction of RATED THERMAL POWER.

These limiting heat flux conditions are higher than those calculated for l

the range of all control rods fully withdrawn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of the f 1

(delta I) function of the overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power imbalance effect on the overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

2.1.2 REACTOR COOLANT SYSTEM PRESSURE t

The restriction of this Safety Limit protects the integrity of the Reactor Coolant System from overpressurization and thereby prevents the release of radionuclides contained in the reactor coolant from reaching the containment' i

atmosphere.

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

The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

i i

l 1

l l

l

~

FARLEY-UNIT 1 B 2-2 AMENDMENT NO. 121 l

m

LIMITING SAFETY SYSTEM SETTINGS BASES

-Overnower AT The overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting) under all possible overpower conditions, limits the required range for overtemperature delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermcwell, and RTD response time delays from the core to RTD output indication.

l Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a joss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

Loss of F12W The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

This FARLEY-UNIT 1 B 2-5 AMENDMENT No.121 n

i 4

i 3/4.2 POWER DISTRIBUTION LIMITS i

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 1

3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the lLaits specified in Figure 3.2-1.*

APPLICABILITY:

MODE 1 above 50% of RATED THERMAL POWER **

l 1

i ACTION:

l With the indicated AXIAL FLUX DIFFERENCE outside of the limits a.

specified in Figure 3.2-1:

j 1.

Either restore the indicated AFD to within the limits within 1

15 minutes, or i

j 2.

Reduce. THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.

4 k

o

)

SURVEILLANCE REQUIREMENTS l

i 4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits by:

l 4

a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and i

2.

AtleastonceperhourwiththeAFDMonitorAlarminoperable.l l

i i

4 i

• The indicated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside its limits.

• • See Special Test Exception 3.10.2.

FARLEY-UNIT 1 3/4 2-1 AMENDMENT -No. 121

Tl 1

l l

t 120 a

l 100 1

l

(-12,100)

- (+9,100) i l

Unacceptable Unacceptable Operation Operation 80 l

K hO Acceptable Q-Operaton 8

60 hx W

i

.f

(-30, 50)

(+24, 50)

O*

40 l

i i

u.

O i

20 l

0 l

-50

-40

-30

-20

-10 0

10 20 30 40 50 Axial Flux Difference (Delta l)%

en.3 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION j

OF RATED THERMAL POWER FOR RAOC l

FARLEY-UNIT 1 3/4 2-2 AMENDMENT No. 121 i

t

I, '

4 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F g LIMITING CONDITION FOR OPERATION 3.2.2 Fg(Z) shall be limited by the following relationships:

l Fg(Z) 5 [L,31) [K(Z)) for P > 0.5 for VANTAGE 5 fuel P

Fg(Z) s [4.9) [K(Z)) for P s 0.5 for VANTAGE 5 fuel and Fg(Z) s [L.J.2) [K(Z)) for P > 0.5 for LOPAR fuel P

Fg(Z) 5 [4.64) [K(Z)) for P s 0.5 for LOPAR fuel where P =

THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given '

core height location.

APPLICABILITY:

MODE 1 ACTION:

With Fg(Z) exceeding its limits Reduce THERMAL POWER at least 1% for each 1% Fg(Z) exceeds the a.

limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced at least 1% for each 1% Fg(Z) exceedh the limit.

l b.

THERMAL POWER may be increased provided Fg(Z) is demonstrated through incere mapping to be within its limit.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 Fg(Z) shall be evaluated to determine if it is within its limit by:

l a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

FARLEY-UNIT 1 3/4 2-3 AMENDMENT NO.121

t POWER DISTRIBUTION LIMITS SURVEILLANCE REgUIREMENTS (Continued) b.

Determining the computed heat flux hot channel factor Fg (Z),

C as follows:

Increase the measured Fg(Z) obtained from the power distribution map by 3% to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties.

Verifying that Fg (Z), obtained in Specification 4.2.2.2b above, c.

satisfies the relationship in Specification 3.2.2.

d.

Satisfying the following relationship:

F "T" Fg (2 ) s x !C (Z ) fo r P

> 0.5 P x VV (Z )

F "'" x K (Z ) fo r P F8 (2;) s O

s 0.5 0.5 x VV (2:)

RTP Where Fg (Z) is obtained in Specification 4.2.2.2b above, Fg g,

the Fg limit, K(Z) is the normalized Fg(Z) as a function of core height, P is the fraction of RATED THERdAL POWER, and W(Z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

Fg

= 2.45 (VANTAGE 5 fuel)

= 2.32 (LOPAR fuel)

K(Z) provided in Figure 3.2-2 W(Z) provided in the Radial Peaking Factor Limit Report Measuring Fg(Z) according to the following schedules e.

1.

Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which Fg(Z) was last determined *, or 2.

At least once per 31 Effective Full Power Days, whichever occurs first.

• During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or equal to 50% of RATED THERMAL POWER have been achieved and a power,

distribution map obtained.

FARLEY-UNIT 1 3/4 2-4 AMENDMENT NO. 121

1' i

r

= POWER DISTRIBUTION LIMITS i

)

SURVEILLANCE REQUIREMENTS (Continued) f.

With measurements indicating

1. F C(Z)'

q maximum K(Z)j over(Z)

(

C has increased since the previous determination of Fg (Z) either of the following actions shall be taken:

i C

C 1)

Increase Fg (Z) by the Fg (Z) penalty factor specified in the Peaking Factor Limit Report and verify that this value satisfies the relationship in Specification 4.2.2.2d, or w

C 2)

Fg (Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that i

i

' F (Z)',

o maximum is not increasing.

K(Z)j over(Z)

(

g.

With the relationships specified in Specification 4.2.2.2d above not being satisfied:

1)

Calculate the percent Fg(Z) exceeds its limits by the followint expression:

'r

-3 Fj (Z) x W (Z) 1 m axim um 1

x IH h P > M overZ p ay, j

o x K(Z)

)

L P

s "e

Fj (Z) x W (Z) m axim um 1, x IM M P s 0.5, and over Z F,7, n

(

. 0.5

.)

2)

The following action shall be taken:

Within 15 minutes, control the AFD to within new AFD limits which are determined by reducing the AFD limits specified in LCO 3.2.1, Axial Flux Difference, by 1% AFD for each percent Fg(Z) exceeds its limits as determined in Specification 4.2.2.2g.l.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits.

FARLEY-UNIT 1 3/4 2-5 AMENDMENT NO.121 i

aJ o

.-~

4 POWER DISTRIBUTION LIMITS i

SURVEILLANCE REQUIREMENTS (Continued) 1 h.

The limits specified in Specification 4.2.2.2c are applicable in all i

core plane regions, i.e.,

0 - 1004, inclusive.

t 1.

The limits specified in Specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1)

Lower core region from 0 to 15%, inclusive.

2)

Upper core region from 85 to 1004, inclusive.

a 1

1 4.2.2.3 When Fg(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2, an overall measured F (Z) shall be obtained from a power g

distribution map and increased by 3% to account for manufacturing tolerances and 7

j further increased by 5% to account for measurement uncertainty, f

4 t

4 f

I 1

1 l

l FARLEY-UNIT 1 3/4 2-6 AMENDMENT No.

121 m

i 1

POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-F$H LIMITING CONDITION FOR OPERATION F5Hshallbelimitedbythefollowingrelationship:

3.2.3 F5Hs1.70 [1 + 0.3 (1 - P)) for VANTAGE 5 fuel and F$Hs1.30(1+0.3(1-P)] for LOPAR fuel l

where P =

RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

WithF$Hexceedingitslimits a.

Within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> either:

1.

RestoreF$H to within the above limit; and demonstrate throughin-coremappingthatF$H 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 /> of exceeding the limit, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High Trip Setpoints to s 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 />, and b.

Demonstrate through in-core' mapping, if not previously performed pera.1above,thatF$H 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 Identify and correct the cause of the out of limit condition prior c.

to increasing THERMAL POWER above the reduced limit rsquired by a or b, above; subsequent POWER OPERATION may proceed provided that F$H is 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after attaining 95% or greater RATED THERMAL POWER.

i FARLEY-UNIT 1 3/4 2-8 AMENDMENT NO.121 e

1. m

i TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operatina Steam Generator (Percent of RATED THERMAL POWER) 1 60***

l 2

43 l

3 24 l

TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint, Operatino Steam Generator *

(Percent of RATED THERMAL POWER) 1 2

3

• At least two safety valves shall be OPERABLE on the non-operating steam generator.

• • These values left blank pending NRC approval of 2 loop operation.

• • • For plant operation approaching end of cycle (i.e.,

core average burnup 2 14,000 MWD /MTU), with one inoperable safety valve on any steam generator, the maximum allowable Power Range Neutron Flux setpoint may be increased from 60% to 87% RTP.

FARLEY-UNIT 1 3/4 7-2 AMENDMENT NO.121

2/1 2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) meeting the DNB design criterion during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

Fg(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

Ffg Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

I' 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper g

bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERABLE excore channels is outside the allowed AI operating space for RAOC operation specified in Figure 3.2-1 and the THERMAL POWER is greater than 50% RATED THERMAL POWER.

FARLEY-UNIT 1 B 3/4 2-1 AMENDMENT NO.121 I

r

_3 POWER DISTRIBUTION LIMITS 4

BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR. NUCLEAR ENTHALPY HOT CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance 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:

Control rods in a single group move together with no individual a.

rod insertion differing by more than i 12 steps, indicated, from the group demand position.

b.

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

The control rod insertion limits of Specifications 3.1.3.5 and c.

3.1.3.6 are maintained.

I d.

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

F$gwillbemaintainedwithinitslimitsprovidedconditionsa.through

d. above are maintained. TherelaxationofF%gasafunctionofTHERMALPOWER allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowance of 5% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor Fg(Z) is measured periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(Z), to provide assurance that the limit on the heat flux hot channel factor FQ(Z) is met.

W(Z) accounts for the effects of normal j

operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

FARLEY-UNIT 1 B 3/4 2-2 AMENDMENT NO. '121

1l 1

I S

PAGE INTENTIONALLY LEFT BLANK 8

e O

4 s

FARLEY-UNIT 1 B 3/4 2-3 AMENDMENT No.121

. POWER DTSTRIBUTTON LfMITS

! l BASES When Fh is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. The specified limit for Fh contains an 8% allowance for uncertainties. The 8% allowance is based on the following considerations:

Abnormal perturbations in the radial power shape, such as from rod i

a.

misalignment, af fect Fh more directly than F,

g b.

Although rod movement has a direct influence upon limiting Fg to within its limit, suchcontrolisnotreadilyavailabletolimitFh,and Errors in prediction for control power shape detected during startup c.

physics tests can be compensated for in Fg by restricting axial flux distribution. This compensation for Fh is less readily available.

If Fh exceeds its limit, the unit will be allowed 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to restore Fh to within its limits. This restoration may, for example, involve realigning any misaligned rods or reducing power enough to bring Fh within its power dependent limit. When the Fh limit is exceeded, the DNBR limit is not likely violated in steady state operation, because events that could significantly perturb the Fh value, e.g., static control rod misalignment, age considered in the safety analyses. However, the DNBR limit may be violated if a DNB limiting event occurs while Fh is above its limit. The increased allowed action time of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> provides an acceptable time to restore Fh to within its limits without allowing the plant to remain in an unacceptable condition for an extended period of time.

Once corrective action has been taken, e.g.,

realignment of misaligned rods or reduction of power, an incore flux map must be obtained and the measured value of Fh verified not to exceed the allowed limit. Twenty additional hours are provided to perform this task above the four hours allowed by Action Statement 3/4.2.3.a.

The completion time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is acceptable because of the low probability of having a DNB limiting event within this 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period and, in the event that power is reduced, an increase in DNB margin is obtained at lower power levels. Additionally, operating experience has indicated that.

this completion time is sufficient to obtain the incore flux map, perform the required calculations, andevaluateFh.

FARLEY-UNIT 1 B 3/4 2-4 AMENDMENT NO. 121 n

i l

POWER DISTRIBUTION L7MITS 1

BASES a

3/4.2.4 OUADRANT POWER TILT RATIO 1

1 The quadrant power tilt ratio limit assures that the radial power distribution i

satisfies.the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during power operation.

i The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts.

k The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the maximum allowed power by 3 percent for each Arcent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is j

inoperable, the movable incore detectors are used to confirm that the normalized i

symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

1 The incore detector monitoring is done with a full incore flux map or two sets of i

four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-ll, H-3, H-13, L-5, L-ll, and N-8.

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are i

maintained within the normal steady state envelope of operation assumed in the i

transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to meet the DNB design criterion throughout each analyzed transient. The indicated.T value of 580.7'F is based on the average of two control board readings and kn i

indication uncertainty of 2.5'F.

The indicated pressure value of 2205 psig is i

based on the average of two control board readings and an indication j

uncertainty of 20 psi.

The indicated total RCS flow rate is based on one elbow tap measurement from each loop and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).

I j

The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T and pressurizer pressure through the control ay board readings are sufficient t$ ensure that the parameters are restored 4

j within their limits following load changes and other expected transient operation.

l l

The 18 month surveillance of the total RCS flow rate is a precision measurement that verifies the RCS flow requirement at the beginning of each 3

fuel cycle and ensures correlation of the flow indication channels with the j

measured loop flows. The nonthly surveillance of the total RCS flow rate is a reverification of the RCS flow requirement using loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of the fuel l

cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators and the loop i

elbow tap measurements that are correlated to the precision RCS flow measurement l

at the beginning of each fuel cycle.

i i

)

FARLEY-UNIT 1 B 3/4 2-5 AMENDMENT NO.121 i

2 A

1 i ADMINISTRATIVE CONTROLS MONTHLY OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Commission, pursuant to 10 CFR 50.4, no later than the 15th of each month following the calendar month covered by the report.

PEAKING FACTOR LIMIT REPORT l

c 6.9.1.11 The cycle dependent function W(Z) and the curnup dependent Fg {g)

C penalty factors, required for calculation of Fg (Z) specified in LCO 3.2.2, " Heat Flux Hot Channel Factor - Fg(Z)," shall be documented in the Peaking Factor Limit Report in accordance with the methodology in WCAP-lO216-P-A, " Relaxation of s

Constant Axial Offset Control FQ Surveillance Technical Specification," Rev. 1, February 1994 (W Proprietary).

The Peaking Factor Limit Report shall be provided to the Commission, pursuant to 10 CFR 50.4, upon issuance prior to each reload cy~cle (prior to MODE 2).

In the event that the limit would be submitted at some other time during core life, it will be submitted upon issuance, unless otherwise exempted by the commission.

ANNUAL DIESEL CENERATOR RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually.

This report shall contain the information identified in Regulatory Position C.'3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.

i FARLEY-UNIT 1 6-19 AMENDMENT NO. 121

m..

_.m

_.-_.m_

g i

UNITED STATES i

s j

NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 30806 4001

\\...../

SOUTHERN NUCLEAR OPERATING COMPANY. IN G ALABAMA POWER COMPANY i

DOCKET NO. 50-364

~

JOSEPH M. FARLEY NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE i

l Amendment No.113 License No. NPF-8' 1.

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

i A.

The application for amendment by Southern Nuclear Operating Company, Inc. (Southern Nuclear), dated. June 12, 1996, complies with the standards and requirements of the Atomic. Energy Act of j

1954,.as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I;

)

8.

The facility will operate in conformity with the application, the provisions of-the Act, and.the rules and 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 Commission's regulations and all applicable requirements have been satisfied.

2.

Ac'cordingly, 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:

i i

l

l

. (2)

Technical Soecifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.113, are hereby incorporated in the license.

Southern Nuclear shall operate the facility in accordance with the Technical Specifications.

3.

This license amendment is effective as of its date of issuance and shall be implemented within 30 days of issuance.

F0 THE NUCLEAR REGULATORY COMMISSION He bert N. Berkow, Director Project Directorate II-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of Issuance:

September 3,1996 1

A

i ATTACHMENT TO LICENSE AMENDMENT NO.113 TO FACILITY OPERATING LICENSE NO. NPF-8

)

DOCKET NO. 50-364 Replace the following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Remove Paaes Insert Paaes XIX XIX 2-2 2-2 2-8 2-8 2-9 2-9 l

2-10 2-10 B 2-2 B 2-2 B 2-5 B 2-5 3/4 2-1 3/4 2-1 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-8 3/4 2-8 3/4 7-2 3/4 7-2 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-3 B 3/4 2-3 8 3/4 2-4 B 3/4 2-4 8 3/4 2-5 B 3/4 2-5 6-19 6-19 b

q INDEX ADMINISTRATIVE CONTROLS SECTIOli PAGE Review...................................................

6-10 Audits...................................................

6-11 Authority................................................

6-12 Records..................................................

6-12 6.5.3 TECHNICAL REVIEW AND CONTROL Activities...............................................

6-12 l

1 l

Records..................................................

6-13 L6 REPORTABLE EVENT ACTION.....................................

6-14 6.7 SAFETY LIMIT VIOLATION

..................................... 6-14 6.8 PROCEDURES AND PROGRAMS.....................................

6-14 6.9 REPORTING REQUIREMENTS 6.9.1 ROUTINE REPORTS i

Startup Report

.......................................... 6-15a Annual Report............................................

6-16 Annual Radiological Environmental Operating Report.......

6-17 Annual Radioactive Effluent Release Report...............

6-17 Monthly Operating Report.................................

6-19 l

Peaking Factor Limit Report..............................

6-19 Annual Diesel Generator Reliability Data Report..........

6-19 Annual Reactor Coolant System Specific Activity Report... 6-20 Annual Sealed Source Leakage Report......................

6-20 6.9.2 SPECIAL REPORTS...........................................

6-20 6.10 RECORD RETENTION............................................ 6-20 6.11 RADIATION PROTECTION PROGRAM................................

6-21a 6.12 HIGH RADIATION AREA.........................................

6-22 7ARLEY-UNIT 2 XIX AMENDMENT No. 113

680 UNACCEPTABLE 660 2440 psia OPERATION I

2250 psia 640 2000 psia 4

u.

v en620

>g 1840 psia

?

600

~

ACCEPTABLE OPERATION 580 4

560 i

0.0 0.2 0.4 0.6 0.8 1.0 1.2 POWER (FRACTION OF RATED THERMAL POWER)

Figure 2.1-1 Reactor Core Safety Limits Three Loops in Operation FARLEY-UNIT 2 2-2 AMENDMENT No. 113 n

TABLE 2.2-1 fContinued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

NOTATION N

Note 1: Overtemperature AT 1 + r* s' 1 + r's' 'T 1

- T' + K (P - P')- f (AI)

AT s AT, K, - K 2

3 i

(1 + r s; (1 + r s; s 1 + r s; w

3 2

o where:

AT Measured AT by RTD instrumentation;

=

ATO Indicated AT at RATED THERMAL POWER and reference Tavg;

=

T Average temperature, 'F;

=

T' Reference T at RATED THERMAL POWER (s 577.2*F);

=

ayg Pressurizer pressure, psig; P

=

P' 2235 peig (nominal RCS operating pressure);

=

1+Ts1 l+T, The function generated by the lead-lag controller for T,yg dynamic compensation; 2

m T1 &T2 Time constants utilized in the lead-lag controller for T,yg, t1 = 30 sec,

=

T2 = 4 sec; 1+Ts4 1+v85 Time constants utilized in the lead-lag controller for AT, t4 = 0 see, T5 s 6 see; T4 & v5

=

1+T8

• 9 6

t6 Time constant utilized in the measured T lag compensator, t6 s 6 see; l

=

ayg Laplace transform operator, sec~1; s

=

ag Operation with 3 loops Operation with 2 loops K1 = 1.17; Ki_= (values blank pending l

s

.O K2 = 0.017; K2 = NRC approval of l

[

K3 = 0.000825; K3 = 2 loop operation) l

.4 v

~

TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS N

NOTATION fContinued1 a

c:

and ft (AI) is a function of the indicated difference between top and bottom detectors of the power-range d

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

u (1) for qt - 9b between -23 percent and +15 percent, ft (AI) =0 (where qt and qb are percent RATED f

THERMAL POWER in the top and botto,m 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 -23 percent, the AT trip setpoint shall be automatically reduced by 2.48 percent of its value at RATED THERMAL POWER; and (iii) for each percent that the magnitude of (qt - 9b) exceeds +15 percent, the AT trip setpoint shall be automatically reduced by 2.05 percent of its value at RATED THERMAL POWER.

Note 2:

Overpower AT u

a r i + r,s' r,s l

l AT s AT. K. - K T

- Kc T

- T"

- f (AI) 3 2

s1+rss (1 + r ss (1 + r s; s1 + r ss 3

3 o

where AT = Measured AT by RTD instrumentation; ATO Indicated AT at RATED THERMAL POWER and reference Tavgi

=

T = Average temperature, 'F; 1

T"

= Reference T at RATED THERMAL POWER (s 577.2*F);

l ayg K4

= 1.10; l

]

=

{

K5 0.02/*F for increasing average temperature and 0 for decreasing average

=

g temperature;

]

5 Ks = 0.00109/*F for T > T", K6 = 0 for T s T";

l 1s 3

The function generated by the rate lag controller for T yg dynamic compensation; 1

=

i

TABLE 2.2-1 (Continued)

N REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Q

NOTATION fContinued1 c:

5" Time constant utilized in the rate lag controller for T,yg, t3 = 10 sec; 23

=

w 1 + r4s The function generated by the lead-lag controller for AT dynamic compensation;

=

I+

s 4 & v5

= Time constants utilized in the lead-lag controller for AT, T4 = 0 sec, 25 s 6 see; j 1

I

= Lag compensator on measured Tayg; I+r s 6

6

= Time constant utilized in the measured T lag compensator, T6 s 6 see; l-t ayg

-1; w

a

= Laplace transform operator, sec d

f2(AI) 0 for all AI.

=

Note 3: The channel's maximum trip point shall not exceed its computed trip point by more than 0.4 percent AT span.

Note 4: Pressure value to be determined during initial startup testing. Pressure value of 5 55 psia to be used prior to determination of revised value.

Note 5: Pressure value to be determined during initial startup tasting.

Note 6: The channel's maximum trip point shall not exceed its computed trip point by more than 0.4 percent AT span.

n E

5 x

O w

.m

SAFETY LIMITS BASES i

i The curves of Figures 2.1-1 and 2.1-2 are based on the most limitingresultusinganenthalpyhotchannelfactor,F[H,of1.65for VANTAGE 5fuelandanF$Hof1.30forLOPARfuelandareferencecosinewitha j

peak of.l.55 for axial power shape. An allowance is included for an increase inF5Hatreducedpowerbasedontheexpression F$H=1.65 [1 + 0.3 (1 - P)) for VANTAGE 5 fuel and F5H=1.30[1+0.3(1-P)] for LOPAR fuel l

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 f i-(delta I) function of the Overtemperature trip. When the axial power imbalance is not within the tolerance, the axial power Lmbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

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

The reactor pressure vessel, pressurizer and the reactor coolant system piping and fittings are designed to Section III of the ASME Code for Nuclear i

Power Plant which permits a maximum transient pressure of 110% (2735 psig) of design pressure. The Safety Limit of 2735 psig is therefore consistent with l

the design criteria and associated code requirements.

i The entire Reactor Coolant System is hydrotested at 3107 psig, 125% of design pressure, to demonstrate integrity prior to initial operation.

I f

l FARLEY-UNIT 2 8 2-2 AKENDMENT NO.

113 1

]

n.

i l

I LIMITING SAFETY SYSTEM SETTINGS BASES Overnower AT The Overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet molting) under all possible overpower conditions, limits the required range for Overtemperature delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for axial power distribution, changes in density and heat capacity of water with temperature, and dynamic compensation for transport, thermowell, and RTD response time delays from the core to RTD output indication.

l Pressurizar Pressure The Pressuriser High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 peig).

The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure.

Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

t Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 10 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

i Above 36% (P-8) of RATED THERMAL POWER, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow.

This 1

l FARLEY-UNIT 2 8 2-5 AMENDMENT NO.113 A

4 s

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD1 LIMITING COF9ITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained within the limits specified in Figure 3.2-1.*

APPLICABILITY:

MODE 1 above 50% of RATED THEF. MAL POWER **

l ACTION:

a.

With the indicated AXIAL FLUX DIFFERENCE outside of the limits specified in Figure 3.2-1:

1.

Either restore the indicated AFD to within the limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes.

SURVRILLANCE REQUIREMENTS 4.2.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits by:

l a.

Monitoring the indicated AFD for each OPERABLE excore channel:

1.

At least once per 7 days when the AFD Monitor Alarm is OPERABLE, and 2.

At least once per hour with the AFD Monitor Alarm inoperable. l l

1

• The ind!.cated AFD shall be considered outside of its limits when at least 2 OPERABLE excore channels are indicating the AFD to be outside its limits.

i

• • See Special Test Exception 3.10.2.

1 FARLEY-UNIT 2 3/4 2-1 AMENDMENT NO. 113 1

h A

= _ - - -. -.

4 1

i 120 4

I i

100

.i 3

(-12,100)

(+9,100) i Unacceptable Unacceptable Operation Operation j

80

(

1 m

)(

Acceptable 0-Operaton 8

60 y

l 2

I w

h

(-30, 50)

(+24, 50) i o

W 40 j

u_O i

i 1

-l 20 1

i i

o l

-50

-40

-30

-20

-10 0

10 20 30 40 50 l

Axial Flux Difference (Delta 1)%

. 42 3 FIGURE 3.2-1 AXIAL FLUX DIFFERENCE LIMITS AS A FUNCTION OF RATED THERMAL POWER FOR RAOC l

FARLEY-UNIT 2 3/4 2-2 AMENDMENT NO.113 i

i A

1 POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR - F g LIMITING CONDITION FOR OPERATION 3.2.2 Fg(Z) shall be limited by the following relationships:

Fg(Z) s [2 mil) [K(Z)) for P > O.5 for VANTAGE 5 *uel P

Fg(Z) s (4.9) [K(Z)) for P s 0.5 for VANTAGE 5 fuel and Fg(Z) s [2.32) [K(Z)] for P > 0.5 for LOPAR fuel P

Fg(Z) s [4.64) [K(Z)] for P s 0.5 for LOPAR fuel where P =

THERMAL POWER RATED THERMAL POWER and K(Z) is the function obtained from Figure 3.2-2 for a given>

core height location.

APPLICABILITY:

MODE 1 ACTION:

With Fg(Z) exceeding its limits a.

Reduce THERMAL POWER at least 1% for each 1% P (Z) exceeds the g

limit within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower delta T Trip setpoints have been reduced at least 1% for each 1% Fg(Z) exceeds the limit.

l b.

THERMAL POWER may be increased provided Fg(Z) is demonstrated through incore mapping to be within its limit.

SURVEILLANCE REQUIREMENTS 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

4.2.2.2 F (Z) shall be evalrated to determine if it is within its limit by:

l g

a.

Using the movable incore detectors to obtain a power distribution map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

l FARLEY-UNIT 2 3/4 2-3 AMENDMENT NO.113

'4 I

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) b.

Determining the computed heat flux hot channel factor Fg (Z), as follows:

Increase the measured Fg(Z) obtained from the power distribution map by 34 to account for manufacturing tolerances and further increase the value by 5% to account for measurement uncertainties.

c.

Verifying that Fg (Z), obtained in Specification 4.2.2.2b above, satisfies the relationship in specification 3.2.2.

d.

Satisfying the following relationships F "" x K (Z )

Fj (Z ) s fo r P > 0.5 P x W (Z )

Fj(Z) s F "" x K (Z ) fo r P s 0.5 0.5 x W (Z)

Where Fg (Z) is obtained in Specification 4.2.2.2b above, Fg is the Fg limit, K(Z) is the normalized Fg(E) as a function of core height, P is the fraction of RATED THERMAL POWER, and W(2) is the cycle dependant function that accounts for power distribution transients encountered during normal operation.

Fg

= 2.45 (VANTAGE 5 fuel) i

= 2.32 (LOPAR fuel)

K(Z) provided in Figure 3.2-2 W(Z) provided in the Radial Peaking Factor Limit j

Report e.

Measuring FQ(Z) according to the following schedules j

1.

Upon achieving equilibrium conditions after exceeding by 20%

or more of RATED THERMAL POWER, the THERMAL POWER at which

{

Fg(Z) was last determined *, or 2.

At least once per 31 Effective Full Power Days, whichever occurs first.

• During power escalation after each fuel loading, power level may be increased until equilibrium conditions at any power level greater than or f

equal to 50% of RATED THERMAL POWER have been achieved and a power ~

distrlhucion map obtained.

FAALEY-UNIT 2 3/4 2-4 AMENDMENT NO.113

. POWER DISTRIBUTION LIMITS i

SURVEILLANCE REgUIREMENTS (Continued) f.

With measurements indicating

'F C(Z)'

n maximum K(Z)s over(Z) s i

has increased cince du previous determination of Fg (Z) either of the following actions

.1 be taken:

i C

1)

Increase Fg (Z) by the Fg (Z) penalty factor specified in the Peaking Factor Limit Report and verify that this value satisfies the relationship in Specification 4.2.2.2d, or 2)

Fg (Z) shall be measured at least once per 7 Effective Full Power Days until two successive maps indicate that I

1. F (Z)'

o maximum is not increasing.

K(Z)s over(Z) l s

)

g.

With the relationships specified in Specification 4.2.2.2d above ryot being satisfied:

1)

Calculate the percent Fg(Z) exceeds its limits by the following expression:

'r y

C Fo (Z) x W (Z) m axim um

-1 x 100 for P > 0.5 over Z F "'"

j n

,s P

_s

'r Fo (Z) x W (Z) m axim um

-1 x 100 kr P s 0.5, and over Z F,,,

n

(

_ 0.5

_s i

2)

The following action shall be taken Within 15 minutes, control the AFD to within new AFD liraits which are determined by reducing the AFD limits specified in LCo 3.2.1, Axial Flux Difference, by 1% AFD for each percent Fg(Z) exceeds its limits as determined in Specification 4.2.2.2g.1.

Within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, reset the AFD alarm setpoints to these modified limits.

FARLEY-UNIT 2 3/4 2-5 AMENDMENT No.113

i POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) l h.

The limits specified in Specification 4.2.2.2c are applicable in all core plane regions, i.e.,

0 - 100%, inclusive.

j i.

The limits specified in specifications 4.2.2.2d, 4.2.2.2f, and 4.2.2.2g above are not applicable in the following core plane regions:

1)

Lower core region from 0 to 15%, inclusive.

2)

Upper core region from 85 to 1004, inclusive.

i 4.2.2.3 When Fg(Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2, an overall measured Fg(E) shall be obtained from a power distribution map and increased by 34 to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty.

o l

)

L U

t FARLEY-UNIT 2 3/4 2-6 AMENDMENT No. 113 i

_. _. =

.j' POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEARENTHALPYHOTCHANNELFACTOR-F5H LIMITING CONDITION FOR OPERATION F{Hshallbelimitedbythefollowingrelationship:

3.2.3 F5Hs1.65 [1 + 0.3 (1 - P)] for VANTAGE 5 fuel and F5Hs1.30[1+0.3(1-P)] for LOPAR fuel l

where P =

RATED THERMAL POWER APPLICABILITY:

MODE 1 ACTION:

WithF{Hexceedingitslimits 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 Satpoints to s 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 />, b.

Demonstratethroughin-coremappingthatF$H 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 Identify and correct the cause of the out of limit condition prior c.

to increasing THERMAL POWER above the reduced limit required by a or b, above; subsequent POWER OPERATION may proceed provided that F{H is demonstrated through in-core mapping to be within its limit at a nomiral 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 2 3/4 2-8 AMENDMENT NO.113 i

J m

1 i

TABLE 3.7-1 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 3 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint 1

Ooeratina Steam Generator (Percent of RATED THERMAL POWER) 1 60***

l 2

43 l

3 24 l

TABLE 3.7-2 MAXIMUM ALLOWABLE POWER RANGE NEUTRON FLUX HIGH SETPOINT WITH INOPERABLE STEAM LINE SAFETY VALVES DURING 2 LOOP OPERATION Maximum Number of Inoperable Maximum Allowable Power Range Safety Valves on Any Neutron Flux High Setpoint Operatina Steam Generat;,gr*,

(Percent of RATED THERMAL POWER) i i

1 i

2

.3 l

i i

• At least two safety valves shall be OPERABLE on the non-operating steam l

generator.

s

• • These values left blank pending NRC approval of 2 loop operation.

• • • For plant operation approaching end of cycle (i.e.,

core average burnup 2 14,000 MWD /MTU), with one inoperable safety valve on any steam generator, the 4

maximum allowable Power Range Neutron Flux setpoint may be increased from 60% to i

j 87% RTP.

1 i

i FARLEY-UNIT 2 3/4 7-2 AMENDMENT NO. 113 i

i L

1

,3/4.2 POWER DISTRIBUTION LIMITS i

BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by:

(a) meeting the DNB design criterion during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200*F is not exceeded.

The definitions of certain hot channel and peaking factors as used in these specifications are as follows:

Fg(Z)

Heat Flux Hot Channel Factor, is defined as the maximum local heat flux on the surface of a fuel rod at core elevation Z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods and measurement uncertainty.

Ffg Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of the integral of linear power along the rod with the highest integrated power to the average rod power.

1 3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the Fg(Z) upper

~

bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately i.! the AFD for 2 or more OPERABLE excore channels is outside the allowed AI operating space for RAOC operation specified in Figure 3.2-1 and the THERMAL POWER is greater than 50% RATED THERMAL POWER.

I FARLEY-UNIT 2 B 3/4 2-1 AMENDMENT NO.

113

i

. POWER DISTRIBUTION LIMITS BASES 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, NUCLEAR ENTHALPY HOT CHANNEL FACTOR The limits on heat flux hot channel factor, and nuclear enthalpy rise hot channel factor ensure that 1) the design limit on peak local power density is not exceeded, 2) the DNB design criterion is met, and 3) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance 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:

Control rods in a single group move together with no individual a.

rod insertion differing by more than i 12 steps, indicated, from the group demand position, b.

Control rod banks 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.

d.

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

l F%g will be maintained within its limits provided conditions a. through

d. above are maintained. The relaxation of F%g as a function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits.

When an Fg measurement is taken, an allowance for both experimental error and manufacturing tolerance must be made. An allowsnee of 5% is appropriate for a full core map taken with the in: ore detector flux mapping system and a 3% allowance is appropriate for manufacturing tolerance.

The heat flux hot channel factor Fg(Z) is measurad periodically and increased by a cycle and height dependent power factor appropriate to RAOC operation, W(Z), to provide assurance that the limit on the heat flux hot channel factor Fg(Z) is met.

W(Z) accounts for the effects of normal operational transients within the AFD limits and was determined from expected power control maneuvers over the full range of burnup conditions in the core.

i FARLEY-UNIT 2 B 3/4 2-2 AMENDMENT NO.

113

l l

PAGE INTENTIONALLY LEFT BLANK l

I i

l 1

)

i 1

FARLEY-UNIT 2 B 3/4 2-3 AMENDMENT No.113 i

i

. POWER DISTRIBUTION LIMITS BASES L

WhenFfH is measured, experimental error must be allowed for and 4% is the appropriate allowance for a full core map taken with the incore detection system. ThespecifiedlimitforF$gcontainsan8%allowancefor uncertainties.

The 8% allowance is based on the following considerations:

a.

Abnormal perturbations in the radial power shape, such as from rod F[H more directly than Fg, misalignment, affect b.

Although rod movement has s direct influence upon limiting Fg to within its limit, such control is not readily available to limit F[H, and Errors in prediction for control power shape detected during startup c.

physics tests can be compensated for in Fg by restricting Exial flux distribution. ThiscompensationforF%gislessreadilyavailable.

I 1

FARLEY-UNIT 2 B 3/4 2-4 AMENDMENT NO. 113 m

t 1

POWER DISTRIBUTION LIMITS l

BASES 3/4.2.4 OUADRANT POWER TILT RATIO l

The quadrant power tilt ratio limit assures that the radial power distribution satisfies the design values used in the power capability analysis. Radial power distribution measurements are made during startup testing and periodically during i

power operation.

J The limit of 1.02, at which corrective action is required, provides DNB and linear heat generation rate protection with x-y plane power tilts.

The two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allow identification and correction of a 4

dropped or misaligned control rod.

In the event such action does not correct the tilt, the margin for uncertainty on Fo is reinstated by reducing the maximum i

allowed power by 3 percent for each Mrcent of tilt in excess of 1.0.

For purposes of monitoring QUADRANT POWER TILT RATIO when one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POWER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of 4

four symmetric thimbles. The two sets of four symmetric thimbles is a unique set of eight detector locations. These locations are C-8, E-5, E-ll, H-3, H-13, L-5, L-ll, and N-8, l

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the parameters are maintained within the normal steady state envelope of operation assumed in the transient and accident analyses. The limits are consistent with the initial FSAR assumptions and have been analytically demonstrated adequate to meet the DNB design criterion throughout each analyzed transient.

The indicated T 4

l value of 580.7'F is based on the average of two control board readings and 2n indication uncertainty of 2.5'F.

The indicated pressure value of 2205 peig is based on the average of two control board readings and an indication uncertainty of 20 psi.

The indicated total RCS flow rate is based on one elbow tap measurement from each loop and an uncertainty of 2.4% flow (0.1% flow is included for feedwater venturi fouling).

l The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> surveillance of T,y and pressurizer pressure through the control j

boardreadingsaresufficientt$ensurethattheparametersarerestored i

within their limits following load changes and other expected transient i

operation.

i 4

The 18 month surveillance of the total RCS flow rate is a precision measurement that verifics the RCS flow requirement at the beginning of each l

fuel cycle and ensures correlation of the flow indication channels with the measured loop flows. 'Ihe monthly surveillance of the total RCS flow rate is a twverification of the RCS flow requirement using loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of the fuel cycle. The 12 hour1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> RCS flow surveillance is a qualitative verification of significant flow degradation using the control board indicators and the loop elbow tap measurements that are correlated to the precision RCS flow measurement at the beginning of each fuel cycle.

FARLEY-UNIT 2 B 3/4 2-5 AMENDMENT NO. 113

_ ~ _..

4

. ADMINISTRATIVE CONTROLS I

1 a

l MONTHLY OPERATING REPORT l

6.9.1.10 Routine reports of operating statistics and shutdown experience, j

including documentation of all challenges to the PORV's or safety valves, shall be submitted on a monthly basis to the Commission, pursuant to

.10 CFR 50.4, no later than the 15th of each month following the calendar month I

covered by the report.

PEAKING FACTOR LIMIT REPORT l

6.9.1.11 The cycle dependent function, W(E),Cand the burnup dependent Fg (2) penalty factors, required for calculation of Fg (E) specified in LCO 3.2.2, " Heat Flux Hot Channel Factor - Fg(E)," shall be documented in the Peaking Factor Limit Report in accordance with the methodology in WCAP-lO216-P-A, " Relaxation of Constant Axial Offset Control FQ Surveillance Technical Specification," Rev. 1, February 1994 (H Proprietary).

The Peaking Factor Limit Report shall be provided to the Commission, pursuant to 10 CFR 50.4, upon issuance prior to each reload cycle (prior to NODE 2).

In the event that the limit would be submitted at some other time during core life, it will be submitted upon issuance, unless otherwise exempted by the Commission.

ANNUAL DIESEL GENERATog RELIABILITY DATA REPORT 6.9.1.12 The number of tests (valid or invalid) and the number of failures to start on demand for each diesel generator shall be submitted to the NRC annually.

This report shall contain the information identified in Regulatory Position C.*3.b of NRC Regulatory Guide 1.108, Revision 1, 1977.

i i

FARLEY-UNIT 2 6-19 AMENDMENT NO. 113

.