Text

Amends 92 & 85 to Licenses NPF-2 & NPF-8,changing TS to Allow Use of Vantage-5 Fuel Assemblies in Both Units & Removal of RTD Bypass Manifold Sys in Unit 2
	ML20094L364
Person / Time
Site:	Farley
Issue date:	03/11/1992
From:	Adensam E; Office of Nuclear Reactor Regulation
To:	;
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ML20094L368	List: ML20094L364; ML20094L380;
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'

. (e.%jo UNITED STATES

g

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NUCLE AR REGULATORY COMMISSION 3m g

E WASMNGTON, D. C. 20555 o

d' SOUTHERN NUCLEAR OPERATING COMPANY. INC.

DOCKET NO. 50-348 AOSEPH M. FARLEY NUCl[AR PLANT. UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 92 License No. NPF-2 1.

The Nuclear Regula, tory Comission (the Comission) has found that:

i A. The application for amendment by the licensee, dated July 15, 1991, as supplemented September 10, 1991, and January 10, 1992, 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, the provisions of the Act, and the rules and regulations of the Comission; C. There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health an 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 Comission'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:

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' \\

\\

P

.. (2)

Technical Specificallani The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 92, 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.

FOR THE NUCLEAR REGULATORY COMMISSION Jbj/%(k(Is I/ Elinor G. Adensam, Director Project Directorate 11-1 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation

Attachment:

Changes to1the Technical Specifications Date.of Issuance: March 11, 1992 4

l

AUhCHMENT TO LICENSE AMENDMENT NO. 92 TO FACILITY OPERATING LICENSE NO. NP,[-2 DOCKET NO. 50-348 Replace the following pages of the Appendix A Technical Specifications with the enclosed pa9es.

The revised areas are indicated b3 marginal lines.

Remove Paaes Insert Paaes 2-2 2-2 2-5 2-5 2-8 2-8 2-9 2-9 2-10 2-10 B 2-1 B 2-1 B 2-2 8 2-2 B 2-3 8 2-3 8 2-6 B 2-6 3/4 1-4 3/4 1-4 3/4 1-19 3/4 1-19 3/4 2-4 3/4 2-4 3/4 2-7 3/4 2-7 3/4-2-8 3/4 2-8 3/4 2-14 3/4 2-14 3/4 2-15 3/4 2-15 3/4 3-10 3/4 3-10 3/4'3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-30 3/4 3-30 3/4 4-2 3/4 4-2 8 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 B 3/4 2-4 B 3/4 2-5 8 3/4 2-5 B 3/4 4-1 B 3/4 4-1 6-19 6-19 L

l

I u'

670<

660 N UNACCEPTABLE-OPERAT10N 2440 psia 650'

640, 2250 pala L

630<

- [

2000 pela p 620" a-1875 psia

-610<

1840 pela 600' 590 ACCEPTABLE OPERATION 580' 570 0.

1

.2 3.

.4 -

~.5

.6 -

- 7

.8

.9 1,

1.1 1.2 POWER (FRACTION OF RATED THERMAL POWER)

Figure-2.1-1 Reactor Core Safety Limits Three Loops in'0peration FARLEY - UNIT 1

'2-2 AMENDHENT NO. 37, 72, 57, 92 g-y y-,,, --

..m.

~

F

v1 TABLE 2.2-1 5

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS e

E-Q FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES 1.

Manual Reactor Trip Not Applicable Not Applicable 2.

Power Range, Neutron Flux Low Setpcint - f 25% of MTED Lov Setpoint - f 26% of RATED THERMAL POVER THERMAL POVER High Setpoint - f 109% of RATED High Setpoint - f 110% of RATED THERMAL POVER THERMAL POVER 3.

Power Range, Neutron Flux, f 5% of RATED THERMAL POVER vith-f 5.5% of RATED THERMAL POVER High Positive Rate a time constant 2 2 seconds with a time constant i 2 seconds 4.

Power Range, Neutron Flux, f 5% of RATED THERMAL POVER vith f 5.5% of RATED THERMAL POWER High Negative Rate a time constant 2 2 seconds with a time constant'2 2 seconds v

5.

Intermediate Range, Neutron 5 25% of RATED THERMAL POVER f 30%. of RATED THEPJfAL POVER Flux 6.

Source Range, Neutron Flux 5 10' counts per second 5 1.3 % 10' counts per.second 7.

Overtemperature AT See Note 1 See Note 3 8.

Overpower AT See Note 2 See Note 6 9.

Pressurizer Pressure-Lov 2 1865 psig 2 1855 psig 10.

Pressurizer Pressure--High f 2385 psig i 2395 r,sig Eg 11.

Pressurizer Vater f 92% of instrument span f 93% of instrument span Level-High x

524 12.

Loss of Flov 2 90% of minimum measured flov 2 88.5% of minimum measured flov l

,5 per loop

per loop

em N.*

Hinimum measured flow is'89,290 gpm per loop.

l D

TABLE 2.2-1 (Continued) g

REACTOR TRIP SYSTEM INSTRUMFETATI C TRIP SETPOINTS' n

.E 3 NOTATION a

[

Note 1: Overtemperature M 5E y

E -(1 + T,s) f AT, [K - K (1 +T s)'

(T (

1 ) - T') + K (P - P') - f (SI)]

3 2

i 3

3 (1 + T s)

(1.+T,s) 1 + T, s 3

vhere:-

M = Measured M by RTD instrumentation; E, = Indicated E at RATED THERMAL POVER;-

T = Average temperature, 'F; T' f 577.2*F (Maximum Reference T,,, at RATED THEPJML POVER);

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

Y f++Ts = The function generated by the lead-lag controller for T,,, dynamic compensation; m

gs A

y T & T,

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

T = 30 sec, T, = 4 see; 3

3

Ts'

= The function generated by the lead-lag controller for E dynamic competsation; 4

1+Ts 3 T, &T

= Time constants utilized in the lead-lag controller for M, T,

=T

= 0 see; 3

3 1

= Lag compensator on measured T,,,;

y 1 + T, s mg T,

= Time constant utilized in the meuured T,,, lag compensator, T,

= 0.sec; r

j

- Laplace transform operator, sec *;

s Operation with 3 loops Operation with 2 loops 1.14; K

K (values blank pending l

=

3 3

ru pq K, = 0.0250; K, = NRC approval of l

0.001275; K

K 2 loop op ration) l

=

3 3

TABLE 2.2-1 (Continued)

REACTOR TRIP STSTEM INSTRUMENTATION TRIP SETPOINTS n

E NOTATION (Continued) t l

and f (AI) is a function of the indicated difference between top and bottom detectors of the pover-range y

nuclear lon chambers; with gains to be selected based on measured instrument response during plant startup 1

tests such that:

,.o for q

-q between -39 percent and +13 percent, f, (AI) = 0 (where q and q, are percent RATED l

(i)

THERM 1L PODER in the top and bottom halves of the core respectively, a,nd q, + q, is total THERMAL i

POWER in percent of RATED THERMAL POVER);

I (ii) for each percent that the magnitude of (q, - q,) exceeds -39 percent, the aT trip setpoint shall be automatically reduced by 1.92 percent of its value at RATED THERMAL POVER; and (iii) for each percent that the magnitude of (q, - q,) exceeds +13 percent, the AT trip satpoint shall be automatically reduced by 2.17 percent of its value at RATED THERMAL POVER.

Note 2: Overpower AT at (1 + t,s) f AT, (K,- K ( Ts )

(

1

) T - K, (T (

1

) - T") - f,(AI)]

w 3

3 (1 + T s) 1

+T s 1 +

T, s 1 + T, s 3

3 where:

AT = Measured af by RTD instrumentation; ST,= Indicated aT at RATED THERMAL POVER; T = Average temperature, 'F; T" = Reference T,,, at RATED THERMAL POVER (Calibration temperature for AT instrumentation,

{

f 577.2*F);

)

l g

K, = 1.07; K = 0.02/*F for increasing average temperature and 0 for decreasing average temperature; 3

l k

K, = 0.00165/'F for T > T*, K, = 0 for T f T";

5

= The function generated by the rate lag controller for T,,, dynamic compensation; Ts2 1+T s g

3 i

,3 D

TABLE 2.2-1 (Continued)-

E

, REACTOR' TRIP SYSTEM' INSTRUMENTATION TRIP SETPOINTS

+

NOTATION'(Continued)

U

- ka 1; = Time' constant utilized in the rate lag controller for T,,,,

19 = 10.sec; I*Ts

- The function generated by the lead-lag controller. for ar dynamic compensation; i

1+T35 T, & T; - Time constants utilized in the' lead-lag' controller for cr, 1q = 1; - 0'see; 1

= Lag compensator on measured T,,,;

} +T8 6

T, = TimeLeonstant utilized in the measured T,,, lag compensator, T, = 0 see;

[

s - LaplaceLtransform operator, sec

~*

o i

.f2(oI) = 0 for all AI.

Note 3: The channel's maximum trip point shall'not exceed its computed trip point by more than 1.8 percent.

. l Note 4: Pressure value to be determined during initial startup testing.

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

Note 5:

Pressure value to be determined during initial startup testing.

Note 6: 1Nue channel's maximum trip point shall not exceed its computed trip point by more than 2.3 percent.

l

!2 E

4

,s 5

L e-

2.1 SAFETY LIMITS s

BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and possible cladding perforation which would result in the release of fission products to the reactor coolant.

Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient in 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 POVER and Reactor Coolant Temperature and Pressure have been related to DNB through correlations which have been developed to predict the l

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 vould cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB thermal design criterion is that the probability of DNB not occurring on the most limiting rod is at least 95 percent (at a 95 percent confidence level) for any Condition I or Il event.

In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered.

As described in the FSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty.

Design limit DNBR values have been determined that satisfy the DNB design criterion.

i Additional DNBR margin is maintained by performing the safety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNBR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

The curves of Figures 2.1-1 and 2.1-2 show the reactor core safety limits for a range of THERMAL POVER, Reactor Coolant System pressure and average temperature which satisfy the following criteria:

a.

The average enthalpy at the vessel exit is less than the enthalpy of saturated liquid (far left line segment in each curve).

b.

The minimum DNBR satisfies the DNB design criterion (all the other line segments in each curve). Each curve reflects the most limiting result using either lov-parasitic (LOPAR) fuel or VANTAGE 5 fuel.

The VANTAGE 5 fuel is analyzed using the VRB-2 correlation with design limit DNBR values of 1.24 and 1.23 for the typical and thimble cells, respectively. The LOPAR fuel is analyzed using the VRB-1 correlation with design limit DNBR values of 1.25 and 1.24 for the typical and thimble cells, respectively.

c.

The hot channel exit quality is not greater than the upper limit of the quality range (including the effect of uncertaintles) of the DNB correlations. This is not a limiting critorion for this plant.

FARLEY - UNIT 1 B 2-1 AMENDMENT NO. 37, 92 l

-~

SAFETY LIMITS l

BASES I

j The curves of Figures 2.1-1 and_2.1-2 are based on the most limiting result uging an enthalpy hot channel factor, F,,, of 1.65 for VANTAGE 5 fuel and an r,, of 1.55 for.LOPAR fuel and a reference cosine vifh a pgak of 1.55 for axial J

power shape. An allowance is included for an increase in F,, at reduced pover based on the expression:

F",, =_l.65 [1 + 0.3 (1-P)] for VANTAGE 5 fuel and l

F",, = 1.55 [1 4 0.3 (1-P)] for LOPAR fuel l

vbere P is the fraction of RATED THERMAL POVER.

These limiting heat flux conditions are higher than those calculated for the range of all control rods fully'vithdrawn to the maximum allovable control rod insertion assuming the axial power imbalance is within the limits of the f 3

(delta I) function of the Overtemperature trip. Vhen the axial power imbalance is not within the toler:.nce, the axial power imbalance eftect on the overtemperature delta T trips vill 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 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.

FARLEY - UNIT 1 B 2-2 AMENDMENT NO.

26, 92

2.2 LlH! TING SAFETY SYSTEH SETTINGS BASES l

w w

2.2.1 REACTOR TRIP SYSTEM INSTRlfr(ENTATION SETPOINTS i

The Reactor Trip Setpoint Limite specified in Table 2.2-1 are the values at a

vhich the Reactor Trips are set for each functional unit.

The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and.

del;gn basis anticipated operational occurrentes and to assist the Engineered l

Safety Features Actuation System in mitigating the consequences of accidents.

Opetation with a trip set less conservative than its Trip Setpoint but within its specified Allovable Value is acceptable on the basis that the difference between each Trip Setpoint and the A11ovable Value is equal to or less than t e h

drift a11avance assumed for each trip in the safety analysis.

Ha.vaal Reactor Trip

'8hc Harual Reactor Trip is a redundant channel to the automatic protective l

insttusentation channels and provides manual reactor trip capability.

Poter Range, Neutron Flux The Power Range Neutron Flux channel high setpoint provides reactor cote prote:titn against reactivity excursions which are too rapid to be protected by

_tempatatuto and pressute protective circuitry. The lov setpoint provides l

' redundant protection in the power range for a pover excursion beginning from lov povet. The trip associated vith the lov setpoint may be manually bypassed when P-10 is active (two of the four power range channels indicate a pover level of I

eSove opprcximately 10 percent of RATED THEttHAL POVER) and is automatically reinstated when F 10 becomes inactive (three of the four channels indicate a l

L power level belov approximately 8 percent of RATED TilERHAL POVER).

L t

Power Rarte d eutron Flux, High Rates Tio Paer Range Positive Rate trip provides protection against rapid flux increases which are cnaracteristic of rod ejection events from any pn er level.

SpNLfically, this trip complements the Pover Range Neutron Flux Higi and iav t:1pu to ensute that the criteria are met for rod ejection from pad tal p, et,

j The Pover F.ange Negative Rate trip provides protection to en m ? ina.t Pe l

D% design criterion is met for control rod drop accidents. At high power a tultiple aod drop accid 9nt could cause local flux peaking'vhich, wher. %

r.onjunction with nucleaf povor being maintained equivalent to turbine pover by action of c.e automatic rod control system, could cause an unconservative local DNB4 to exist. The P6.ier Range Negative Rate trip vill prevent this from-occun try t;y tripping the reactor for multiple dropped rods. No credit was taken for operation of this trip in the accident analyses; however, its functiony.1 cepability at the specified trip setting is required by this E

specificatfori to enhance the overall reliability of the Reactor Protection System, i

j

-FAR15Y a UNIT 1 B 2-3 AMENDHENT NO. 25, 92

LlHITING_SATETY_ SYSTEM SETTINGS BASES latter trip vill ensure that the DNB design criterion is met during normel l

l operational transients and anticip.ated transients when 2 Toeps are in operatics and the overtemperature delta T trip setpoint is adjusted to the value specified for all loops in operation. Vith the Overtemperature delta T trip setpoint adjusted to the value specified for 2 loop operation, the P-8 trip at 66%

RATED TilERHAL POVER vill ensure that the DNB design criterion is met during l

normal operational transients and anticipated transients with 2 loops in operation.

Steam Generator Vater Level The Steam Generatot Vater Level Low-Lov trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allovance that there vill be suffleient vater inventory in the steam generators at the time of trip to allov for starting delays of the auxiliary feedvater system.

Steam /reedvater Flov Hismatch and Lov Steam Generator Vater Level The Steam /reedvater Flov Hismatch in coincidence with a Steam Generator Lov Vater Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor protection System.

This trip is redundant to the Steam Generator Vater Level Lov-Lov trip.

The Steam /Feedvater Flow Hismatch portion of this trip is activated when,the steam flov exceeds the feedvater flov by greater then on equal to 1.55 x 10 Its/ hour.

The Steam Generator Lov Vater Level portion of the trip is activated when the water level drops belov 25 percent, as indicated by the narrov range instrument.

These trip values include sufficient allovance in excess of normal operating values to preclude sputious trips but vill initiate a reactor trip before the steam generators are dry. Therefore, the required capac4.ty and starting tiha requirements of the auxiliary feedvater pumps ate reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltage snd Underf requency - Reactor Coolant pump Busses The Undervoltage and Underfrequency Reactor Coolant pump bus trips provide reactor core protection against DNB as a result of loss of voltage or under-frequency to more than one reactor coolant pump.

The specified setpoints assure a reactor trip signal is generated before the lov flov trip setpoint l'

FARLEY - UNIT 1 B 2-6 AMENDHEtiT No. N, 92

REACTIVITY CONTROL SYSTEMS H0DERATOR TEMPERATURE COEFFICIENT L1HITING CONDITION FOR OPERATION 3.1.1.3 The moderator temperature coefficient (HTC) shall het Less than or equal to 0.7 x 19~' delta k/k/'r for the all rods u.

withdravn, beginning of cycle life (BOL), condition for power levels up to 70% THERHAL POVER vith a linear ramp to O delta k/k/'r at 100%

THERHAL POVER.

b. Less negative than -4.3 x 10 delta k/k/*r for the all rods withdravn, end of cycle life (EOL), RATED THERHAL POVER condition.

APPLICABILITY:

Specification 3.1.1.3.a - H0 DES 1 and 2* only#

Specification 3.1.1.3.b - H0 DES 1, 2 and 3 only#

ACTION:

a. Vith the HTC more positive than the limit of ',1.1.3.a above, operation in H0 DES 1 and 2 may proceed ptovided:

1. Control rod withdraval limits are established and maintained sufficient to restore the HTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in 110T STANDBY vithin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months . These withdraval limits shall be in addition to the insertion limits of Specification 3.1.3.6.

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

3. A Special Report is prepared and submitted to the Commission pursuant to Specification 6.9.2 vithin 10 days, describing the value of the measured HTC, the interim control rod withdraval limits and the predicted average core burnup necessary for restoring the positive HTC to vithin its limit for the.all rods withdrawn condition.

b. Vith the HTC more negative than the limit of 3.1.1.3.b above, be in 110T SHUTDOVN vithin 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Vith K,,, greater than or equal to 1.0 1 See Special Test Exception 3.10.3 l

TARLEY.- UNIT 1 3/4 1 4 AMENDHENT NO. 57, 86 92 I

~

,..,.. _ _-._.._ _ _ _.~ _.

R ACTIlQY CONTROL SYSTEMS J

ROD DROP TIME L!HITING CONDITION FOR OPERATION i

3.1.3.4 The individual full length (shutdown and control) rod drop time from the fully withotavn position (225 to 231 steps, inclusive)* shall be less than or equal to 2.7 seconds from beginning of decay of stationary gripper l

coil voltage to dashpot entry with:

T,,, greater than or equal to $41*F, and f

a.

b.

All reactor coolant pumps operating.

l APPLICABILITY:

H0 DES 1 and 2.

ACTION:

a.

Vic. trL drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to proceeding to MODE 1 or 2.

b.

Vith the rod drop times within limits but determined with 2 reactor coolant pumps operating. opetation may proceed provided

?

THERHAL POVER is restricted to less than or equal to 66% of RATED THERMAL POVER.

SURVEILLANCE REQUIREMENTS _

i 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

a.

For all rods following each removal of the reactor vessel head, b.-

For specifically affected individual rods following any maintenance on or modification to the control rod drive system which could affect the drop time of those speellic rods, and F

c.

At least once per 18 months.

The fully withdrawn position used for determining rod drop time shall be greater than or equal to the fully withdrawn position used during subsequent plant operation.

FARLEY - UNIT 1 3/4 1-19 AMENDMENT NO. pf, 92

POVT.R DISTRIBUTION LIMITS 3/4.2.2 HEAT f TUX HOT CHANNEL FACTOR d,{Z),

LIMITING CONDITION POR OPERATION 3.2.2 r,(Z) shall be limited by the following relationships:

F,(Z) f [2.45) lK(Z)] for P > 0.$ for VANTAGE 5 fuel l

P T,(Z) f l4.9) [K(Z)] for P f 0.$ for VANTAGE $ ivel and l

F,(Z) f l2.32) [K(Z)) for P ) 0.$ for LOPAR fuel l

P F,(Z) f 14.64) [K(Z)] for P $ 0.$ for LOPAR fuel l

vhere P THERHAL POVER KKT F TlIERHAL POVfR and K(Z) is the function obtained from Figure (3.2-2) for a given core height location.

APPLICABILITY:

H0DE 1

. ACTIONS, i

Vith r,(Z) exceeding its limits Reduce TilERHAL POVER at least 1% for each 1% F within15minutesandsimilarlyreducethePov$r(Z)exceedsthelimit a.

Range Neutron Flux-High Trip Setpoints-vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months POWER OPERATION may proceed for up to a total of 72 hour8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months s: subsequent POVER OPERATION may

- proceed provided the Overpower delta T Trip Setpoints-have been reduced-TripSetpointreductions8a(Z)exceedsthelimit.

at least 1% for each 1% F The Overpower delta T t

ll be performed with the reactor in at least 110T STANDBY.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERHAL POVER above the reduced limit required by a, above; THERHAL POVER may then be increased provided F,(Z) is demonstrated through incore mapping to be within its limit FARLEY - UNIT.1 3/4 2-4 AMENDHENT NO. 26, 7J 92 e

.,-.w.--

r+r...

,,-.yp.r.

nw-w-e,,,

---.,.w.,

,,e

,a.y rsr me-e+

4

--N

'--'*e-r-T-bet e-Wm erw'-

-e'-

w'

"'-m'-=

...- _.-.. -. - - - - -._-.- - -. -.-.~...-.-

4, 4

4 1.2-i I

00'10 s.0,1.0 j

j 12.0,0.833 m

0.8 g

40.6 E

t 80.4 M

0.2 i

0 0

2 4

6 8

10 12 CORE HEIGHT (FEET)

Figure 3.2-2 K(Z) Normalized F (Z) as a Function of Core Height n

f.

FARLEY - UNIT 1 3/4 2'7 AMENDHENT NO.

75, 92 1

t i.

l u

1

' ~ - - ~ -

......_..,______,._.,,..,,w,-+-.---,

l POVERDIS,TRIBllT10_NJIH7TS 3/4.2.3 NUCLEAR ENTilALPY 110T CllANNEL FACTOR - F",,

LIMITING CONDITION FOR OPERATION 3.2.3 F",, shall be limited by the following relationships F",, f 1.65 l1 4 0.3 (1-P)] for VANTAGE 5 fuel and l

F",, f 1.55 [1 4 0.3 (1-P)) for L9 PAR fuel l

vhere P = TilERHAL POVER RATIli THETOTA N ER pFPLICABILI,T) MODE 1 ACTION:

Vith F",, um&g its IN t:

a.

Reduce THERMAL POVER to 1 css than 50% of RATED TilERHAL POVER vithin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-liigh Trip Setpoints to <"

55% of RATED TilERHAL POVER vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Demonstrate through in-core mapping that F",,THERHAL POVER to less tha b.

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 5% of RATED TilERMAL POVER vithin 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 TilERMAL POVER above the reduced limit required by g or b, abovel subsequent POVER OPERATION may proceed provided that F,b demonstrated through in-core mapping to be vithin ite limit at,a nominal 50% of RATED TilERMAL POVER prior to exceeding this THERMAL POVER, at a nominal 75% of RATED TilERHAL POVER prior to exceeding this TilERHAL POVER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter attaining 95% or greater R/.TED TilERHAL POWER.

)

l l

FARLEY - UNIT 1 3/4 2 8 AMENDMENT No. N, 37 j

ff. 92 l

4 POVER DISTRIBUTION LIMITS DNB PARAMETERS

[

L1HITING CONDITION FOR OPERATION

[

3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2 1:

n.

Reactor Coolant System T b.

Pressurizer Pressure c.

-Reactor Coolant System Total Flow Rate.

l APPLICABILITY:

H0DE 1 i

ACTION:

+

Vith any of the above parameters exceeding its limit, restore the parameter to vithin its limit vithin 2 houra or reduce THERHAL POVER to leas than 5% of RATED THERHAL POVER vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREHENTS 4.2.5.1 Each nf the parameters of Table 3.2-1 shall be verified to be within

~ their limits at least once p#r 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be vithin its limit by measurement at least once per 18 months.

4.2.5.3 The indicated RCS flow rate shall be verified to be within the acceptable limit at least once per 31 days.

t-l

- i FARLEY - UNIT 1 3/4 2-14 AHENDHENT NO.

25,- 92 9

y--w y

e.M, - -, w.r*w=-

r -mo --e t

- w we w**-~..<*=w

+-v.w--

r

~-

v-

--,wr

--eae+-*v---

---=+-e wces-uv^ev->--*mmw---

v*c

---e e--w e-+-e>--

--+wc++ew--ww-

s

~

TABLE 3.2-1 5;

DNB PARAMETERS t

'Ey LIMITS PARAME~ER 3 Loops in Operation 2 Loops in Operation Indicated Reactor Coolant System T,,,

f 580.7'F

(**)

[

Indicated Pressurizer Pressure 1 2205 psig*

(**)

[

Indicated Reactor Coolant Systes 2 267,880 gpm***

(**)

i Total Flow Rate T

n Y

G i

>x E

E Limit not applicable during either a THERMAL POVER ramp in excess of 5% of RATED Inu. MAL POVER per

{

minute or a THERMAL POVER step in excess of 10% of RATED THERMAL POWER.

5 Values blank pending NRC approval of 2 loop operation.

Value includes a 2.4% flow uncertainty (0.1% feedvater renturi fouling bias incit-ded).

l eN m.

i t

TABLE 3.3-2 E

. REACTOR TRIP SYSTEM INSTRUMEiRATION PISPONSE TIMES

. El FUNCTIONAL UNIT RESPONSE TIME c5

-1.

Manual Reactor Trip Not Applicable t

2.

Power Range, Neutron Flux a.

High f 0.5 seconds

b.

Lov Not Applicable.

3.

Pover Range, Neutron Flux.

I High Positive Rate Not Applicable F

4 Power Range, Neutron Flux, High Negative Rate Not Applicable l

l I

wg 5.

Intermediate Range, Neutron Flux Not Applicable w

6.

Source Range, Neutron Flux Not Applicable 7.

Overtemperature M f 6.0 seconds

f 8.

Overpover -E Not Applicable f

I 9.

Pressurizer Pressure-Lov

$ 2.0 seconds

10.. Pressurizer Pressure-High f 2.0 seconds i

11.

Pressurizer Vater Level-High Not Applicable E

i E

c i

4

Neutron detectors are exempt from response time testin~. Response time of the neutron flu signal portion 2

of the channel shall be measured fro: detector output or input of first electronic component in channel.

I o

NN-

,N t.

TABLE 3.3-4 (Continued)

E{

ENGINEERED SAFETT FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS I

FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES ti 2.

CONTAINMENT SPRAY a.

Maaual Initiation Not Applicable Not Applicable b.

Automatic Actuation Legic Not Applicable Not Applicable c.

Containment Pressure- -

f 27 psig

$ 28.3 psig l

High-High-High 3.

COhTAINMENT ISOIATION R

a.

Phase "A" Isolation s~

1.

Manual Not Applicable Not Applicable 2.

From Safety Injection Not Applicable Not Applicable Automatic Actuation Logic b.

Phase "B" Isolation 1.

Manual Not Applicable Not Applicable 4

2.

Automatic Actuation Logic Not Applicable Not Applicable 3.

Containment Pressure--

3 27 psig i 28.3 psig i

High-High-High E

c.

Purge and Exhaust Isolation 4

E E

1.

Manual Not Applicable Not Applicable E['

2.

Automatic Actuation Logic Not Applicable Not Applicable l

5

=

m e--

j M

TABLE 3.3-4 (Continued)

-4 5

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMEhTATION TRIP SETPOIhTS j

O e

FUNCTIONAL UNIT TRIP SETPOINT ALLO 7ABLE VALUES j

~

4.

STEAM LINE ISOLATION l

4 a.

Manual Not Applicable Not Applicable I

i b.

Automatic Actuation Not Applicable Not Applicable l

Logic t

c.

Containment Pressure-f 16.2 psig

$ 17.5 psig l

y High-High a

d.

Steam Flow in Two Steam

< A function defined as follows:

< A function defined as follcws:

Y Lines-High, Coincident A op corresponding to 40% of full I ap corresponding to 45% of full

?

Z vith T

-Low-Lov steam flow between 0% and 20% load steam flow between 0% and 20% load and then a op increasing linearly and then a op increasing linearly l

to a op corresponding to 110% of to a op corresponding to 111.5% of

[

full steam flov at full load with full steam flow at full load with T,,, 2 543'F T,,, 1 540*F e.

Steam Line Pressure-Lov 2 585 psig 1 575 psig

(

I 5.

TURBINE TRIP AND FEED VATER I

r ISOLATION I

i a.

Steam Generator Vater 5 75% of narrev range instrument i 76% of narrow range instrument Level-High-High span each steam generator span each steam generator 4

E E

t n

I 5..

9 N

l l

I

TABt.E3.3-5(continusdj ENGINEERED SAFET) FEATURES RES*'ONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIHE IN SECONDS 3.

Pressurizer Pressure-Lov a.

Safety Injection (FCCS) f 27.0

/12.0'il 888 b.

Reactor Trip (from SI)

$ 2.0 c.

Feedvater Isolation f 32.0

d.

Containment Isolation-Phase "A" 3 17.0

e.

Containment Purge Isolation f 5.0 f.

Auxiliary Feedvater Pumps Not Applicable g.

Service Vater System 3 77.0/87.0'8' 4.

Djfferential Pressure Between Steam Lines-High a.

Safety Injection (ECCS) f12.0/22.0

b.

Reactor Trip (from SI)-

f 2.0 c.

Feedvater Isolation f 32.0

d.

Containment Isolation-Phase "A" f 17.0/27.0

e.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable E.

Service Vater System f77.0'*3 /87.0

5.

S_ team Flov in Two Steam Lines-High coincident vith 7

-- Lov-Lov a.

Steam Line Isolation-Not Applicable l

6.

Steam Line Pressure-Lov a.

Safety Injection-(ECCS)

$ 12.0/22.0

b.

Reactor Trip (from SI) f 2.0 c.

Teodvater Isolation f 32.0

d.

Containment Isolation-Phase "A" f 17.0/27.0

e.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable

- g.'

Service Vater System j 77.0/87.0

h.

Steam Line Isolation 3 7.0 FARLEY - UNIT 1 3/4 3-30 AMENDHENT NO. /S, $9, 92

~

.,. -,.~

--.=.-. - - _. -,,..., -.

Rf ACTOR CCNANT_,$jSTEM 007 STANDDY L1HITING CONDITION _FOR OPERATION f

3.4.1.2 At least two of the Reactor Coolant Loops listed belov shall be a

OLERABLE and in operation when the rod control system is opetational or at least two Reactor Coolant Loops listed belov shall be OPERABLE vith one Reactot Coolant Loop in operation when the rod conttol system is disabled by opening the Reactot t

Trip Breakers or shutting down the tod drive motor /genetator sets:*

1.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump, 2.

Reactor Coolant Loop B and its associated steam generator and Reactor Coolant 5. ump, 3.

Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump.

APPLICABILITY:

H0DE 3 ACTION:

a.

Vith less than the above required Reactor Coolant Loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or b9 in.

l HOT SilUTDOVN vithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

i i

I b.

Vith only one Reactor Calant Loop in operation and the rod control system operational, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor Trip Breakers or shut dovn the rod drive motor / generator sets.

c.

Vith no Reactor Coolant Loops in operation, suspend all operations

. involving a reduction in boron concentration of the Reactor Coolant System and immediately initiate cortective action to return the requited coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1 At least the above required Reactor Coolant pumps, if not in operation, shall be determined tu be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.1.2.2-The required Reactor Coolant Loop (s) shall be verified to be in operation snd circulating Reactor. coolant at least once por 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

- 4.4.1.2.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% of vide range indication at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

'* All Reactor Coolint pumps may be de-energized for up to I hour provided (1) no operations are permitted that_vould cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at least 10'F belov saturation temperature.

FARLEY - UNIT 1 3/4 4-2 AHENDHENT NO. 76. 65, 92

,w,.-,

e.-m.

v.m.s

.%.,__,,_.-r_._

.. m. m m.,,,m..

,__,.__,,___,-m,,.,,.,,,,-

,.,,,g

%......~ry_.

3/6.2 POVER DISTRZDUTION LIMITS

-BASES The specifications of this section provide assurance of fuel integrity i

during Condition 1 (Normal Operation) and 11 (Incidents of Moderate Frequency) events by: (a) meeting the DNB design criterien during norical operation and l

in short term transients, and (b) limiting the fission gan release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria.

In addition, limiting the peak linear pover density during condition 1 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:

F,(Z)

Heat Flux Hot Channel Factor, is detined 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.

F",H 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 tod power.

F,Y(Z)

Radial Peaking Factor, is defined as the ratio of peak power density to average power density in the horizontal plane at core elevation Z.

,3/4.2.1 AXIAL FLUX DIFFERENCE The limits on AX1AL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times th,e normalized l

axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes.

Target flux difference is determined at equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels. Tne value of the target flux difference obtained under these conditions divided by the fraction of RATED THERHAL POVER is the target flux difference at RATED THERHAL POVER for the associated core burnup conditiona. Target flux differences for other THERHAL POVER levels are obtained by multiplying the RATED THERHAL POVER value by the appropriate fractional THERHAL POVER level. The periodic updating of the

-target flux difference value is necessary to reflect core burnup considerations.

l l

l FARLEY - UNIT 1 B 3/4 2-1 AMENDHENT NO.

N, 73, 92

POVER DISTRIBUTION LIMITS BASES I

AXIAL FLUX DIFFERENCE (Continuedl Although it is intended that the plant vill be operated with the AFD vithin the +($)% target band about the target flux difference, during rapid plant TilERHXL POVER reductions, control tod motion vill cause the AFD to deviate outside of the target band at reduced T!!ERHAL POVER levels. This deviation vill not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERHAL POVER (with the AFD vithin the target band) provided the time duration of the deviation is limited. Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure (3.2-1) while at THERHAL POVER levels bitveen 50% and 90% of RATED TilERHAL POVER.

For THERHAL POVER levels between 15% and 50% of RATED THERHAL POVER, deviations of the AFD outside of the target band are less significant.

The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Honitor Alarm. The computer determines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm messaga immediately if the AFD for 2 or more OPERABLE excore channels are outside the target band and the THERHAL POVER is greater than 90% of RATED TilERHAL POWER.

During operation at TilERHAL POVER levels between $0% and 90% and between 15% and $0% RATED TilERHAL POVER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 !! EAT FLUX !!0T CHANNEL FACTOR, NUCLEAR ENTHALPY il0T

~

CfilfiNEL FACT 0h 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 vill not exceed the 2200'F ECCS acceptance criteria limit.

Each of these is measurable but vill normally only be determined

. periodically as specified in Specificationr 4.2.2 and 4.2.3.

This periodic

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

a.

Control rods in a sinFle group move together with no individual tod insertion differing by more than + 12 steps, indicated, from the group demand position.

b.

Control rod banks ate sequenced with overlapping groups as described L.

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.

1 d,

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

i FARLEY - UNIT 1 B 3/4 2-2 AMENDHENT NO. 76, 92 L

1.

e POVT.R DIFTk1Btff!0N 1! HITS

[

BA$tS 7

F" H vill be maintained within its limits provided conditions a.

through,d. above are maintained. The relaxation of F" THERHAL POVER allows changes in the radial power shape,Il as a function of l

for all permissible rod insertion limits.

Vhen an F measurement is taken, an allovance for both experimental errorandmanuiacturingtolerancemustbemade.

An allovance of $% is appropriate for a full core map taken with the incore detector flux mapping system and a 3% allovance is appropriate for manufacturing tolerance.

Vhen F" H is measured, experimental error must be allowed for and 4% is the appropriate allnvance for a full core map taken with the incore detection system. The specified lin.it for F H contains an 8% allowance for

+

uncertainties. The8%allovanceisbasedonthefollowingconsiderations Abnormal perturbations",in the radial power shape, such as from tod a.

misalignment, affect F H more directly than F,.

b.

Although tod mov ment has a direct influence upon limiting F to vjthinitslimit,suchcontrol1snotreadilyavailabletolimit F,H and-c.

Errors in prediction for co'itt ol povet shape detected during startup r

physics tests can be compensated for ig F,is less readily available.

by restricting axial flux distribution. This compensation for F 113 E

l 1

l FARLEY - UNIT 1 B 3/4 2-4 AMENDMENT NO. ?$, 64, 92 r

L

9100ER DIKTRIBUTTON LIMITS

. BASES The radial peaking factor T,E(Z), is measured periodically te provide additional assurance that the hot chann l factor F Z. remains within its limit.

The F" limitforRATEDTHERHAL1"0VER(pKTP)a,sh(rov)idedintheRadialFeakingFactot Ilmit repot t per Specification 63.1.11 vaa determined f rom expected power control maneuvers over the full range of burnup conditions in the core.

3/4.2.4 OUADRANT POVER TILT RATIO 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 power operation.

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 tvo 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 corrtet the tilt, the margin for uncertainty on F is reinstated by reducing the maximum alloved pover by 3 percent for each perce,nt of tilt in excess of 1.0.

For purposes of monitoring QUADRANT F0VER TILT RATIO vhen one excore detector is inoperable, the movable incore detectors are used to confirm that the normallred symmetric power distribution is consistent with the QUADRANT POVER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of 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-II, H-3, H-13, L-5, L-11, and H-8.

3/4.2.5 DNB PARAHETERS 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 value of $80.7'F-is based on the average of two control board readings anE ln indication uncertainty of 2.5'F.

The indicated pressure value of 2205 psig is based o the average of two control board readings and an indication uncertainty of 20 psi. The indicated total RCS flow rate is based on one elbov tap measurement from each loop and an uncertainty of 2.4% flov (0.1% flov is included for feedvater venturi fouling).

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance of Tavg and pressurizer pressure through the control board readings are sufficient to ensure that the parameters are restored within their limits following-load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate is a precision measurement that verifies the RCS flov requirement at the beginning of each fuel cycle and ensures correlation of the flow indication channels with the measured loop flows. The monthly surveillance of the total RCS flow rate is a reverification of the RCS flow requirement using loop elbov tap measurements that are correlated to the precision RCS flow measurenent at the beginning of the fuel cycle. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months RCS flov surveillance is a qualitative verification of significant flov degradation using the control board indicators and the loop E

elbov tap measurements that are correlated to the precision RCS flow measurement at the beginning of each fuel cycle.

FARLEY - UNIT 1 B 3/4 2-5 AMENDHENT NO. 61, 92

___- - -..-__ --- -~

t 3/4'.4_ RF. ACTOR C00LANT SYSTEH DASLG 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION P

The plant is designed to operate with all Reactor Coolant Loops in operation, and meet the DNB design criterion during all normal operations and l

anticipated transients.

In H0 DES 1 and 2 vith one Reactor Coolant Loop not in operation this specification requires that the plant be in at least HOT STANDBY vithin I hour.

In H0DE 3, two Reactor Coolant Loops provide sufficient heat removal l

capability for removing core heat even in the event of a bank withdraval accidents hovever, a single Reactor Coolant Loop provides sufficient decay heat i

removal capacity if a bank vithdraval accident can be prevented:

1.e.,

by opening the Reactor Trip Breakers or shutting dovn the rod drive motor / generator sets.

l In MODE 4, a single reactor coolant or RHR loop provides sufficient heat removal capability for removing decay heat, but single failure considerations

-require that at least two loops be OPERABLE.

Thus,-if the Reactor Coolant Loops are not OPERABLE, this specification requires two KilR loops to be OPERABLE.

In MODE 5, single failure considerations require two RHR loops to be OPERABLE.

The operation of one Reactor Coolant fump or one RHR pump provides adequate flow to ensure mixing, prevent stratification and ptoduce gradual reactivity changes during boron concentration reductions in the Reactor coolant System.

The reactivity change rate associated with boron reduction vill, therefore, be within the capability of operator recognition and contral.

The restrictions on starting a Reactor Coolant Ptmp vith one or more Reactor Coolant System cold legs less than or equal to 310'r are provided to prevent Reactor Coolant System pressure transients, caused by energy additions from the secondary system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The Reactor Coolant System vill be protected against overpressure transients and vill not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing a volume for the primary coolant to expand into, or (2) by restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

FARLEY - UNIT-1 B 3/4 4-1 AMENDHENT NO.

76, 65, 92

ADMINISTRjlllLfjNTR0ls Type of container (e.g., LSA, Type A, Type B, large Quantity) and e.

f.

Solidification agent (e.g., cement, urea formaldehyde).

The radioactiva effluent release reports shall include unplanned releases from

-the site to unrestricted areas of radioactive materials in gaseous and liquid effluents on a quarterly basis.

The radioactive effluent release reports shall include any changes to the PROCESS CONTROL PROGRAM (PCP) made during the reporting period.

I MONTHtY 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.

4 Any changes to the OFFSITE DOSE CALCULATION MANUAL shall be submitted with the Monthly Operating Report within 90 days in which the change (s) was made effective, in addition, a report of any major changes to the radioactive waste treatment systems shall be submitted with the Monthly Operating Report for the period in which the change was implemented.

RADI AL PEAKl@ FACTOR tlMIT REPORT The-Fxy "imit for Rated Thermal Power (FgP) for all core planes 6.9.1.11 l

containing bank D control rods and all unrodded core planes shall be established and documented in the Radial Peaking f actor Limit Report before each reload cycle (prior to MODE 2) and provided to the Commission, pursuant to 10 CFR 50.4, upon issuance, 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.

AnyinformationneededtosupportThP will be by request from the NRC and need not be included in this. report.

ANNUAL '0!ESEL GENERATOR REllABitITY DATA REPORI 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.

FARLEY-UNIT 1 6-19 AMENDMENT NO. 57., 70..

P2, 92

ww.

1

/psnem'c, UNITE D sT ATis l'

'i NUCt.E AH HEGULATORY COMMISSION f,

w aswiuov ow,o e r<S %

f SOUTHERN NUCLEAR OPERAlll{G COMPANY. lHC.

DOCKET NO. 50-311

&$EPH M. FARLEY_ NUCLEAR PLANT. UNIT 2 AMENDMENT TO FACILITY OPERATING LI{I1(SI Amendment No. 85 License No. NPT-8 1.

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

A. The application for amendment by the licensee, dated July 15, 1991, as supplemented September 10, 1991, and January 10, 1992, 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, the provisions of the Act, and the rules and 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 Commission's regulations; D. The issuance of this license ameridment 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.

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:

(2) 1Rchnical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 85, 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.

FOR lilE NUCLEAR REGULATORY COMMISSION

!*y Y/!fb 4IvElinor G. Adensam, Director t

Project Directorate 11-1 Division of Reactor Projects - !/11 Office of Nuclear Reactor Regulation

Attachment:

Changes to the Technical Specifications Date of issuance:

March 11, 1992 4

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AlleDibHT,10 (1(fMLE!L!&ltU._h0 85 10_16CILITY OPERATINQ_Li[MS[_NO. NPf-8 DOCKET NO. 50-364 Replace the following sages of the Appendix A Technical Specifications with the enclosed pages.

11e revised areas are indicated by marginal lines.

Remove Paati insert Paacs 2-2 2-2 2-5 2-5 2-8 2-B 2-9 2-9 2-10 2-10 B 2-1 B 2-1 B 2-2 B 2-2 B 2-3 B 2-3 B 2-4 B 2-4 8 2-5 B 2-5 B 2-6 B 2-6 3/4 1-4 3/4 1-4 3/4 1-19 3/4 1-19 3/4 2-4 3/4 2-4 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-14 3/4 2-14 3/4 2-15 3/4 2-15 3/4 3-10 3/4 3-10 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-28 3/4 3-28 3/4 3-30 3/4 3-30 3/4 4-2 3/4 4-2 B 3/4 2-1 B 3/4 2-1 B 3/4 2-2 B 3/4 2-2 B 3/4 2-4 8 3/4 2-4 B 3/4 2-5 B 3/4 2-5 B 3/4 4-1 B 3/4 4-1 6-19 6-19

(

l

i 670" Uf4 ACCEPTABLE 660 OPERATlON 2440 psia 650'

640, 2250 pela 630 b

2000 pala p 620' 1875 psia 610<

1840 psla 600' 590' ACCEPTABLE OPERA 110N

\\

580" 570 O.

.1

.2

.3

.4

.5

.6

.7

.8

.9 1.

1.1 1'.2 POWER (FRACT!ON OF RATED THERMAL POWER) i Figure 2.1-1 Reactor Core Safety Limits Three Loops in Operati n e

' A 2

2-2 AMENDMENT NO. 77, H, 79, 85

-~ _

. ~ -. _. _. _, - _ -,

TABLE 2.2-1 s

E REACTOR TRIP STSTEM INSTRUMENTATION TRIP SETPOIhTS N

FUNCTICNAL UNIT TRIP SETPOIhT ALLOWABLE VALUES c.5

]

1.

Manual Reactor Trip Not Applicable Not Applicable 2.

Power Range, Neutron Flux Lov Setpoint - $ 25% of RATED Lov Setpoint - $ 26% of PATED THERMAL POVER THERMAL P07ED.

High Setpoint - f 109% of RATED High Setpoint - f 110% af RATED THEEMAL POWER THEFJfAL POWER i

3.

Power Range, Neutron Flux, f 5% of RATED THERMAL POVER vith

$ 5.5% of PATED THERMAL POUER j

High Positive Rate a time constant > 2 ' seconds with a time constant > 2 srecnds i

l 4.

Power Range, Neutron Flux,

< 5% of RATED THERMAL P07ER with

< 5.5% of RATFD TIERMAL P07ER j

High Negative Rate a time constant > 2 seconds with a time coiertant 12 seconds i

y 5.

Intermediate Range, Neutron f 25% of RATED THERMAL POWER f 30% of RATED THERMAL POWER i

Flux j

6.

Source Range, Neutron Flux i 10' counts per second f 1.3 X 10' cos:nts per second t

i 7.

Overtemperature'AT See Note 1 See Note 3 8.

Overpover aT See Note 2 See Note 6 l

t 9.

Pressurizer Pressure-Lov

> 1865 psig 2 1855 psig

'{

10.

Pressurizer Pressure-High 5 2385 psig i 2395 psig i

g 11.

Pressurizer Vater 5 92% of instrument snan f 93% of instrument span j

53 Level-High 4

6

+

EE 12.

Loss of Flov

~> 90% of ninimum measured flow

> 88.5% of minimum measured flov l

4 per loop

krloop*

5 1

i y

Minimum measured flow is 89,290 gpm per loop.

l 4

y i

f s

TABLE 2.2-1 (Continued)

REAC X)R TRIP SYSTEM INdikUMENTATION TRIP SETPOLVr5 NOTATION g

s; Note 1: Overtemperature 'E E (1 +.T,s) S E, [K - K, (1 ( +T s)

(T (

1 ) - T* ) + K (P - P') - f (E)l 3

1 3

3 4

(1 + T s)

(1 +T,s) 1 +T}

3 oa where:

E - Measured E by RTD instrumentation; E,"- Indicated E at RATED THERMAL PO7ER; T - Average temperature, 'F; T' f 577.2*F (Maximum Reference T,,, at RATED THERMAL PO7ER);

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

I*Ts - The function generated by the lead-lag controller for T,,, dynamic compe.vation:

Y i

1+TS 2

- 4 rec; T, & T,

- T17e constants utilized in the lead-lag controller for T,,,,

T - 30 se., T, 2

- The function generated by the lead-lag controller for & dynamic cowper satic;

^Ts e 1+Ts 3 T, &T

- Time constants utilized in the lead-lag controller for E. T, -T

- O set:

3 3

I

- Lag compensator on measured T,,,;

1 + T, s g

T,

- Time constant utilized in the measured T,,, lag compensator, T, = 0 sec; g

s - Laplace transform operator, sec ;

g I

E Operatloa with 2 loops Operation with 3 loops 1.14; K, - (values blank pending l

=?

K 1

?

K, - 0.0250; K,. NRC approval of l

co n 0.001275; K, - 2 loop operation)

{

K 3

i 3

TABLE 1.2-1-(Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRYP SETPOINTS n

-g NOTATION (Continued) 5 t

and fp ( E) is a function of the indicated difference between top and bottom detectors of the power-range e

nuclear ion' chambers; with gains to be selected based on measured instrument response dttring plant startup 25 tests such that:

i e

i (i) for q

-q b

f (M) = 0 (where q and q l

THERM 1L P0 b.etween -39 percent and +13 percent, in the top and bottom halves of the core respectively, a,nd c,., are percent RATEDq, is tot g

POL'ER in percent of RATED THERMAL POWER);

(ii) for each percent that the magnitude of (q, - q,) exceeds -39 percent, the & trip setpoint shall be automatically reduced by 1.92 percent of its value at RATED THERMAL PCVER; and (iii) for each percent that the magnitude of (q,f its value at RATED fBERMAL POVER

- q,) exceeds +13 percent, the & trip setpoint shall be automatically reduced by 2.17 percent o Note 2: Overpower E M (1 + T,s) $ E, [ K,- K ( Ts ) (

1

) T - K (T (

1

) - T*) - 1 (M)]

j 3

3 3

i

(.1 + T s) 1 +T s 1 +

T, s 1 + T, s

,o 3

3 where:

& = Measured & by RTD instrumentation.

E,= Indicated C at RATED THERMAL POLTR; T = Average temperature, 'F; T* = Reference T,,, at RATED THERMAL POVER (Calibration temperature for E instrtmentation, 1

$ 577.2*F);

j K, = 1.07; l

4

]

K - 0.02/*F for increasing average temperature and O for decreasing average temperature; 3

a:

h K, = 0.00165/'F for T > T*, K, = 0 for T f T*;

l 4

Ts g

= The functioa generated by the rate lag controller for T,,, dynamic compensation; 3

1+T s j

3

$.D I

i

1-

~

l TABLE 2.2-1 (Continued) m

.. h REACTG1R TR77 STSTEM LMSTRUMENTATION TRIP SETPOINTS m

4 NOTATION (Continued) cz we

- Time constant utilized in the rate lag controller for T,,,,

T - 10 see; T3 3

I ' T s,.The function generated by the lead-lag controller for & dynamic compensation; a

I+Ts3 l

Tg & T - Time constants utilized in the lead-3ag controller for M,

T, T - O sec; 3

3 4

1

- Lag compensator on measured.T,,,;

]

1 + T, s T, - Time constant utilized in the measured T,,,

lag compensator, T, - O see:

2

~*

's - Laplace transform operator, sec -

w I.

f2(aI) - O for all AI.

o Note 3: The channel's r:aximum trip point shall not exceed its computed trip point by more thar.1.8 percent.

l t

i Note 4: Pressure value to be determined during initial startup testice+.

Pressure value of f 55 pris to be used prior to determination of revised value.

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

i Note 6: The channel's maximum trip point shall not exceed its computed trip point by more than 2.3 percent.

l I

i

+

2

)'

E Ox i

~1 i

O 1

i t

i I

em 2.1 SAFETY LIMITS

+

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

Operation above the upper boundary of the nucleate boiling regime could result in excessive' cladding temperatures because of the onset of dept.rture 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 POVER and Reactor Coolant Temperature and Pressure have been related to DNB through correlations which have been develeped to predict the l

DNB flux and-the location of DNB for axially uniform and non-uniform heat flux distributions.

ihe local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that voald cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

The DNB thermal design criterion is that the probaLility of DNB not occurring on the most limiting rod is at least 95 percent (at a 95 percent confidence level) for any Condition I or II event.

In meeting the DNB design criterion, uncertainties in plant operating parameters, nuclear and thermal parameters, fuel fabrication parameters and computer codes must be considered. As described in the FSAR, the effects of these uncertainties have been statistically combined with the correlation uncertainty. Design limit DNBR values have been determined that satisfy the DNB design criterion.

l Additional DNBR margin is.egintained by per. forming the safety analyses to a higher DNBR limit. This margin between the design and safety analysis limit DNbR values is used to offset known DNBR penalties (e.g., rod bow and transition core) and to provide DNBR margin for operating and design flexibility.

The curves of Figures 2.1-1 and 2.1-2 show the reactor core safety limits for a range of THERHAL POVER, Reactor Coolant System _ pressure and average temperature which_ satisfy the i,'llowing criteria a.

The average enthalpy at the vessel exit is loss than the enthalpy of saturated liquid (far left lina segment in each curve).

b.

-The minimum DNBR satisfieA th 4$ design criterion (all the other line l

segments in each curve).

Eaci. curve reflects the most limiting result l

using either low-parasitic (LOPAh; fuel or VANTAGE 5 fuel. The VANTAGE 5 fuel'is analyzed using the VRB-2 correlation vith design limit DNBR values of 1.24 and 1.23.for the typical and thimble cells, l

respectively. The LOPAR fuel is analyzed using the VRB-1 correlation vita design limit DNBR values of 1.25 and 1.24 for_the typical and thimble cells, respectively.

l c.

The hot channel exit quality is not greater than the upper limit of the quality raage (including the effect of uncertainties) of the DNB correlations. This is not a limiting criterion for this plant.

FARLEY - UNIT 2 B 2-1 AMENDMENT NO. 27 85

f*' LIMITS B_A: ' '

The curves of Figures 2.1-1 and 2.b2 are based on t',4 most limiting result uging an enthalpy hot channel factor, #,,, of 1.65 for VANTAGE 5 fuel and an F,, of 1.55 for LOPAR fuel and a reference cosine with a peak of 1.55 for axial power shape.

An allovance is included for an increase in F at reduced power an based on the expression:

F",, = 1.65 [1 + 0.3 (1-P)] for VANTAGE 5 fuel and l

F"a 1.55 l1 + 0.3 (1-P)] for LOPAR fuel l

vhere P is the fraction of RATED T!!ERMAL POVER.

These limiting heat flux conditions are higher than-those calculated for the range of all-control rods fully withdrawn to the maximum allovable control rod insertion assuming the axial power imbalance is within the limits of the f 3

(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 vill reduce the cetpoints 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, pressurirer and the reactor coolant system piping and fittings are designed.to Section III of the ASME Code for Nuclear Power: Plant vt.S permits a maximum transient pressere of 110% (2735 psig) of design pressto r Se Safety Limit of 2735 psig is therefore consistent with the design crite-tv mssociated code requirements.

The ent,& iNa< <or Coolant System is hydrotested at 3107 psig,125% of

-design press.r"

.a lemonstrate integrity prior to initial operation.

FARLEY_- UNIT 2 B 2-2 AMENDMENT No. 85

s

?.3* LIMITING SAFETY SYSTEM SETTINGS BASES 2.2.(_ REACTOR TRIP' SYSTEM INSTRUMENTATION SETPOINTS The Reactor Trip Setpoint Limits specified in Table 2.2-1-are the values at which the Reactor Trips are set for each functional unit. The Trip Setpoints have been selected to ensure that the reactor core and reactor coolant system are prevented from exceeding their safety limits during normal operation and design basis anticipated operational occurrences and to assist the Engineered Safety Features Actuation System in mitigating the consequences of accidents.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allovable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allovable Value is equal to or less than the drift allowance assumed for~each trip in the safety analysis, danual Reactor Trip The Manual Reactor Trip is a redundant channel to the automatic protective instrumentation channels and provides manual reactor trip capability.

Power Range, Neutron Flux The Power Range, Neutron Flux channel high setpoint provides reactor core protection against reactivity excursions which are too rapid to be protected by temperature and pressure protective circuitry. The lov setpoint provides

-redundant protection in the power range for a power excursion beginning from low power. The trip associated with the lov setpoint may be manually bypassed when P-10 is active (tvo of the four power range channels indicate a power level of above-approximately 10 percent of RATED THERMAL POWER) and is automatically

-reinstated when P-10 becomes inactive (three of the four channels indicate a power level belov approximately 8_ percent of RATED THERMAL POVER).

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

Specifically, this trip complements the _ Power Range Neutron Flu,: High and Lov trips to ensure that the criteria are met-for r:d. ejection from partial power.

The Power Range Negative Rate trip provides protection to ensure that the DNB design criterion is met for control rod drop accidents.

At high power a multiple rod drop accident could cause local flux-peaking which, when in I

conjunction with nuclear power being maintained equivalent to turbine power by action of the automatic rod control system, could cause an unconservative local DNBR to exist. 'The Power Range Negative Rate trip will prevent this from

= occurring by tripping the reactor for multiple dropped rods.- 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.

FARLEY - UNIT 7 B 2-3 AMENDMENT NO.

85

LlHITING SAFETY SYSTEM SETTINGS BASES Intermediate and Source Range, Nuclear, Flux The Intermediate and Source Range, Nuclear Flux trips provide reactor core protection during reactor startup.. These trips provide redundant protection to the lov setpoint trip of the Power Range, Heutron Flux chapnels,' The Source Range Channels vill initiate a reactor trip at about 10' counts per second unicas manually blocked when P-6 becomes active.

The Intermediate Range Channels vill initiate a reactor trip at a current level proportional to approximately 25 percent of RATED TilERHAL POVER unless manually blocked when P-10 becomes active.

No credit was taken for operation of the trips associated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System.

Overtemperature oT The Overtemperature delta T trip provides core protection to prevent DNB be all combinations of pressure, power, coolant temperature, and axial power distribution, provided that the transient is slov vith respect to piping transit, thermovell, and RTD response time delays from the core to the l

temperature detectors (about 4 seconds), and pre':sure is within the range

'between the liigh and Lov Pressure reactor trips. This setpoint includes corrections for changes in density and heat capacity of water with temperature and dynamic compensation for transport, thermovell, and RTD response time delays from the core to RTD output indication.

Vith normal axial power distribution, this resetor trip limit is always below the core safety limit as shovn-in Figure 2.1-1.. If axial peaks are greater than design, as indicated by the difference between top and bottom power range nuclear detectors,-the reactor trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 3 loop P-8 setpoint does not require reactor. protection system setpoint modification because the P-8 setpoint and associated trip vill prevent DNB during 2 loop operation exclusive of the overtemperature delta T setpoint.

Two loop operation above the 3 loop P-8 setpoint is permissible after resetting the K1, K2, and-K3 inputs-to the Overtemperature delta T channels and raising the P-8 setpoint to its 2 loop value.

In this mode of the.P-8 iro rlock and trip functions as a liigh Neutron Flux trip operation, e

at the reduced power level, i-(

FARLEY - UNIT 2 B 2-4 AMENDMENT NO.

85 l

E

_. ~. -

l 1

LIMITING' SAFETY! SYSTEM SETTINGS BASES Overpower AT-The Overpower delta T reactor trip provides assurance of fuel integrity (e.g., no fuel pellet melting) under all possible overpower conditions, l

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 vith temperature, and dynamic compensation for transport, thermovell, and RTD response time delays from the core to RTD output indication.

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.

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. lover than the set pressure for these valves (2485 psig).

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

Pressurizer Vater Level

~The Pressurizer High Vater 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'Flov The Loss of Flov trips proiide core protection to prevent DNB in the event of a loss of-one or more reactos, coolant pumps.-

Above 10 percent of RATED TilthMAL POVER, an automatic reactor trip vill occur if the flow in any two-loops drop belov 90% of nominal full loop flow.

Above 36% 1P-8) of RATED THERHAL POVER, automatic reactor trip vill occur if the flow in any single loop drops belov 90% of nominal full loop flow. This l

FARLEY - UNIT 2 B 2-5 AMENDMENT NO.

65 v= y w

w s

v w-

--s__rs_a L1HITING SAFETY SYSTEM SETTINGS i

BASES latter trip vill ensure that the DNB design criterion is met during normal l

operational transients and anticipated transients when 2 loops are in operation and the Overtemperature delta T trip setpoint is adjusted to the value specified for all loops in operation.

With the overtemperature delta T trip setpoint adjusted to the value specified for 2 loop operation, the P-8 trip at 66%

RATED TilERHAL POVER vill ensure that the DNB design crlierion is met during l

normal operational transients and anticipated transients with 2 loops in operation.

Steam Generator Vater Level The Steam Generator Vater Level Low-low trip provides core protection by preventing operation with the steam generator water level below the minimum volume required for adequate heat removal capacity. The specified setpoint provides allowance that there vill be sufficient water inventory in the steam generators at the time of trip to allow for starting delays of the auxiliary feedvater system.

Steam /Feedvater Flow Hismatch and Lov Steam Generator Vater Level The Steam /Feedvater Flov Hismatch in coincidence with a Steam Generator Lov Vater Level trip is not used in the transient and accident analyses but is included in Table 2.2-1 to ensure the functional capability of the specified trip settings and thereby enhance the overall reliability of the Reactor Protection System.

This trip is redundant to the Steam Generator Vater Level Low-Lov trip.

The Steam /Feedvater Flov Hismatch portion of this trip is activated when the steam flov exceeds the feedvater finv by greater than or equal to 1,55 x 10' lbs/ hour.

The Steam Generatot Lov Vater Level portionoof the trip is activated when the water level drops belov 25 percent, as indicated by the narrow range instrument. These trip values include sufficient allowance in excess of normal operating values to preclude spurious trips but vill initiate a reactor trip before the steam generators are dry. Therefore, the required capacity and starting time requirements of the auxiliary feedvater pumps are reduced and the resulting thermal transient on the Reactor Coolant System and steam generators is minimized.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide reactor core protection against DNB as a result of loss of voltage or under--

frequency to more than one reactor coolant pump. The specified setpoints assure a reactor trip signal is generated before the lov flov trip setpoint FARLEY - UNIT 2 B 2-6 AMENDHENT NO. 85

- - ~

hEACTIVITY CONTROL SYSTEMS H0DERATOR TEMPERATURE COEFFICIENT 1

i LIMITING CONDITION FOR OPERATION l

3.1.1.3 The moderator temperature coefficient (HTC) shall bei

a. Less than or equal to 0.7 x 10-* delta k/k/'F for_the all rods withdravn, beginninF of cycle life (BOL), condition for power levels up to 70% THERHAL POVER vith a linear ramp to O delta k/k/*F at 100%

THERHAL POVER.

b. Less negative than -4.3 x 10 delta k/k/*F for the all rods withdrawn, end of cycle life (EOL), RATED THERHAL POVER condition.

APPLICABILITY: Specification 3.1.1.3.a - H0 DES 1 and 2* only#

Specification 3.1.1.3 b - H0 DES 1, 2 and 3 only#

1 ACTION:-

a. Vith the HTC more positive than the limit of 3.1.1.3.a above, operation in H0 DES 1 and 2 may proceed provided:

-1. Cont.ol rod withdraval limits are established end maintained sufficient to restore the HTC to within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or be in HOT STANDBY vjthin the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

These vithdraval limits shall be in addition to the insertion limits of Specification 3.1.3.6.

2. The control rods are maintained within~the withdraval limits established above until a subsequent calculation verifies that

{

the HTC has been restored to within its limit for the all rods t

vithdrawn-conditlon, j

i

3. A Special Report is prepared and submitted to the Commission pursuant a Specification 6.9.2 vithin 10 days, describing the i

value of the measured HTC, the interim control rod withdrawal limits and the predicted average core burnup necessary for restoring the positive HTC to within.its limit for the all rods withdrawn condition, b Vith the HTC more negative than the limit of 3.1.1.3.b above, be in HOT S!!UTDOVH within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

Vith K,, greater than or equal to 1.0

? See Special Test Exception 3.10.3 FARLEY - UNIT 2 3/4 1_4 AMENDHENT NO. 49, SO 85

REAdTIVITYCONTROLSYSTEMS

. ROD DROP TIME-

-LIMITING CONDITION FOR OPERATION 1.1.3.4 The individual full length (shutdovn and control) rod drop time

om the fully withdrawn position (225 to 231 steps, inclusive)* shall be less than or equal to 2.7 seconds from beginning of decay of stationury tripper l-coil voltage to dashpot entry with:

T,,, greater than or equal to 541'F, and a.

b.-

All reactor enolant pumps operating.

APPLICABILITY:

MODES 1 and 2.

ACTION:

a.

Vith the drop time of any full length rod determined to exceed the above limit, restore the rod drop time to within the above limit prior to-proceeding to MODE 1 or 2.

b.

- Vith the rod drop times within limits but determined with 2 reactor coolant pumps operating, operation may proceed provided TilERMAL POWER is restricted to less than or equal to 66% of RATED THERMAL.POVER.

SURVEILLANCE REQUIREMENTS 4.1.3.4 The rod drop time of full length rods shall be demonstrated through measurement prior to reactor criticality:

.a.

For all rods following each removal of the reactor vessel head, b.

For specifically affected individual rods following any maintenance on or modification to sie control. rod drive system which could affect the drop time of those specific rods, and' c.

At least once per 18 months.

The fully withdrawn position used for determining rod drop time shall be greater than or equal to the fully. withdrawn position used during subsequent plant operation.

FARLEY - UNIT 2 3/4 1-19 AMEJDMERT NO.

76, 85

l POVER DISTRIBUTION LIMITS

'3/4.2.2 HEAT FLUX HOT CilANNEL FACTOR - F,(y LIMITING CONDITION FOR OPERATION 3.2.2 F,(Z) shall be limited by the following relationships:

F,(Z) j [2.45] [K(Z)] for P > 0.5 for VANTAGE 5 fuel l

P F,(2) f [4.9) [K(Z)] for P f 0.5 for VANTAGE 5 fuel and l

F,(Z) ~< [ T ] [K(Z)] for P > 0.5 for LOPAR fuel l

2.32 F,(Z) < [4.64] lK(Z)] for P f 0.5 for LOPAR fuel l

vhere P THERMAL POVER RATED TilERHAL POVL7 and K(Z) is the function obtained from Figure (3.2-2) for a given core height location.

APPLICALILITY:

H0DE 1 ACTION:

Vith F,(Z) exceeding its limits Reduce THERMAL POVER at least 1% for each 1% F,(Z) exceeds the limit a,

within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months ; POVER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months ;. subsequent POVER OPERATION may proceed provided the Overpower delta T Trip Setpoints have been reduced

-at least 1% for each 1% F (Z) exceeds the limit. The overpower delta T Trip 5etpoint reduction shall be performed with the reactor in at least HOT STANDBY.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POVER above the reduced limit required by a, above;

.THERHAL POVER may then be increased provided'F,(Z) is demonstrated through incore mapping to be within its limit l.

l FARLEY - UNIT 2 3/4 2-4 AMENDHENT NO. fJ, S$,

85 1

j,.

1.2 0,0,1.0 6.0,1.0 12.0,0.933 n6o 0.8 u.

O uJ N

D 4 - 0.6 3m O-Z e

n.

E 0.4 0.2 0

0-2 4

6 8

10

-12 CORE HEIGHT (FEET)

Figure 3.2.2 ~K(2) Normalized F,(2.) as a Function of Core Height FARLEY - UNIT 2 3/4 2-7 AMENDMENT NO.

85 l

l'

-~,

mu---

POVER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTilALPY HOT CHANNEL FACTOR _- F",,

LIMITING CONDITION FOR OPERATION 3.2.3-F",,

shall be limited by the following relationships F"anf1.65[14 0.3 (1-P)) for VANTAGE 5 fr-and l

F",, f 1.55 [1 + 0.3 (1-P)] for LOPAR fuel l

vhere P = TilERHAL POVER RATED THERHKL POVER APPLICABILITY:

H0DE 1

-ACTION:

Vith F",n exceeding its limits a.

Reduce THERHAL POVER to less than 50% of RATED TilERHAL POVER within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Pover Range Neutron Flux-High Trip Setpoints to f 55% of RATED THERHAL POVER vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , Demonstrate through_in-core mapping that F",,THERHAL POVER to less than b.

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 5% of RATED THERHAL POVER vithin 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 TilERHAL POVER above the reduced limit required by a or b, abovel subsequent POVER OPERATION may proceed provided that F",,

is demonstrated through in-core mapping-to be within its limit at a nominal 50% of RATED THERHAL POVER-prior to exceeding this TilERHAL POVER, at a nominal 75% of ' RATED THERHAL POVER prior to exceeding this TilERHAL POVER and within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months af ter attaining 95% or greater RATED TilERHAL POVER.

FARLEY - UNIT 2 3/4 2-8 AMENDHENT NO. 77,. g7 85

r POVER DISTRIBUTION LIMITS

'DNB PARAMETERS LIMITING CONDITION FOR OPERATION

_3.2.5 The following DNB related parameters shall be maintained within the limits shown on Table 3.2-1:

Reactor Coolant System T,,,

a.

b.

Pressurizer Pressure c.

Reactor Coolant System Total Flow Rate.

APPLICABILITY:

MODE 1 ACTION:

Vith any of the above parameters exceeding its limit, restore the parameter to within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POVER to less than 5% of RATED THERMAL POVER vithin the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.5.1-Each of the parameters of Table 3.2-1 shall be verified to be within

.their limits at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.2.5.2 The Reactor Coolant System total flow rate shall be determined to be within its limit by measurement at least once per 18 months.

4.2.5.3 The indicated RCS flow rate shall-be verified to be within the acceptable limit _at least once per 31 days.

FARLEY - l! NIT 2 3/4 2-14 AMENDMENT NO. 35

1

=

a

'{

.)

~

' TABLE 3.2-1' s2 DNB PARAMETERS Ei i

LIMITS c:

'E

-i PARAMETER 3 Loops in operation 2 Loops in Operation u

Indicated Reactor Coolant System T,,,

f 580.7'F

(**)

l Indicated Pressurizer Pressure 1 2205'psig*

(**)

j' Indicated Reactor Coolant System 1 267,880 gpm***

(**)

l

' Total Flow Rate M

v.

Y U

Limit not applicable during either a THERMAL POVER ramp in excess of 5% of RATED THERMAL POVER per 6

minute or a THERMAL POVER step in excess of 10% of RATED THERMAL POLTR.

55 Values blank pending NRC approval'of 2 loop operation.

'z Value includes a 2 4% flow uncertainty (0.1% feedvater venturi fouling bias included).

l F

M

TABLE 3.3-2.

5' REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES

.E FUNCTIONAL UNIT.

RESPONSE TIME l@

U l.

Manual Reactor Trip' Not Applicable w

2.-

Power Range, Neutron Flu'x a.

High f 0.5 seconds

b.

Lov Not Appilcable 3.

Pover' Range,. Neutron Flux, High Positive Rate Not Applicable 4,

Power Range,-Neutron Flux, High Negative Rate Not Applicable l

[

15.

Intermediate Range, Neutron Flux

.Not Applicable 6.

Source Range, Neutron. Flux Not Applicable ga 7.

Overtemperature BT f 6.0. seconds

l E.

'Overpover ar Not Applicable 9.

Pressuriter Pressure--Lov f 2.0 seconds 10.

Pressurizer Pressure--Bigh f 2.0 seconds 11.

Pressurizer Vater Level--High Not Applicable E

e h

Neutron detectors are exempt from response time testing. Response time of the neutron flux signal portion of the-channel shall be measured from detector output or input of first electronic component in channel.

z?

!"n

'l TABLE 3.3-4-(Continued)

-t s

ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP'SETPOINTS

?

0 FUNCTIONAL UNIT

- TRIP SETPOU1T ALLOVABLE VA. LUES h

2.

CONTAINME!TT SPRAY a

N a.

Manual Initiation Not Applicable Not Applicable b.

Automatic Actuation Logic Not Applicable Not Applicable c.

Containment Pressure- -

$ 27 psig

< 23.3 psig l

High-High-High' 3.

COTEUNMEffr ISOLATION a.

Phase "A" Isolation

[

1.-

-Manual Not Applicable Not Applicable y

2.

From Safety Injection Not Applicable Not Applicable g

. Automatic Actuation' Logic b.

Phase "B" Isolation 1.

Manual Not Applicable Not Applicable 2.

Automatic Actuation Logic Not Applicable Not Applicable 3.

Containment Pressure--

f 27 psig

$ 28.3 psig l

High-High-High 1

c.

Purge and Exhaust Isolation 1.

Manual Not Applicable Not Applicable C

E 2.

Automatic Actuation Logic Not Applicable Not Applicable l

4 5

-~

D Y,

'e

~ TABLE 3.3 4.(continued)'

y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

- FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUiLS.

-4.'

STEAM LINE ISOLATION u

a.

Manual Not Applicable Not Applicable b.

Automatic Actuation Not Applicable Not Applicable

. Logic c.

Containment Pressure--

f 16.2 psig f 17.5 psig i

High-High d.

Steam Flov in Two Steam f A function defind as follovs:

$ A function defined as follovs:

Lines-High, Coincident A op corresponding to 40% of full A op corresponding to 44% of full u

S with T

-Lov-Lov steam-flow between 0%'and 20% load steam flow between G% and 20% load and then a op increasing linearly and then a op increasing linearly Y

to a op corresponding to 110% of to a op corresponding to 111.5% of full steam flow at full load with full steam flov at full load with T, y 2 543*F T,,, 1 540'F l

e.

Steam Line Pressure-Lov.

1 585 psig 2 575 psig 5.

TURBINE TRIP AND FEED VATER ISOLATION a.

Steam Generator Vater.

I 75% of narrov range instriment f 76% of narrow range instrument Level-High-High span each steam generator span each steam generator E

E m

5

n..

TABLE 3.3-4 (Continued)

' ENGINEERED SAFEIT FEATURE ACTUATION SYSTEM INSTRUMENTATION NtIP SETPOINTS ng r--

FUNCTIONAL UNIT TRIP SETPOINT ALLOVABLE VALUES N.

6.

AUXILIARY FEEDVATER a.

Automatic Actuation N.A.

N.A.

H' Logic w

b.

Steate Generator Vater 2 17% of r.arroi., range instrument 2 16% of narrow range instrument

. Level-Lov-Lov span ear.'n steam generator span each steam generator c.

Undervoltage - RCP 2 26F0 volts 1 2640 volts d.

S.I.

See 1 above (all SI Setpoints) e.

Trip of Main Feedvater N.A.

N.A.

Pumps i

.w

')

7.

LOSS OF POVER y

a.

4.16 kv Emergency Bus 2 3255 volts bus voltage

2 3222 vo?ts bus voltage

g Undervoltage (Loss of f 3418 volts bus voltage

Voltage)-

b.

4.16 kv Emergency Bus 2 3675 volts bus voltage

2 3638 volts bus voltage

Undervoltage (Degraded f 3749 volts bus voltage

Voltage)

8.. ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS a.

Pressurizer Pressure, f 2000 psig

$ 2010 psig P-ll E

E3 b.

(Increasilid,P-12 Lov-Low T Ei 544'F

< 547'F (Decreasing) 543'F E540*F 4

2 c.

Steam Generatcr Level, (See 5. above)

P P-14 N

f*

d.

. Reactor Trip, P-4 N.A.

N.A.

Refer to-appropriate relay setting sheet calibration requirements.

L.

--------- - -.a- -----

TABLE 3.3-5 (Continued)

ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS 3.

--Pressurizer Pressure-Lov a.

Safety Injection (ECCS)

$ 27.0/12.0

b.-

Reactor Trip (from SI) f 2.0 c.

Feedvater Isolation f 32.0

d.-

Containment Isolation-Phase "A" f 17.0

e.

Containment Purge Isolation 5 5.0 f.

-Auxiliary Feedvater Pumps Not Applicable g.-

Service Vater System f 77.0/87.0'*'

4.

Dif ferential Pressure Between Steam Lines.-High a.

Safety Injection (ECCS) f 12.0/22.0

b.

Reactor Trip (from SI)

$ 2.0 c.

IFeedvater Isolation f 32.0

d.

. Containment-Isolation-Phase "A"

$ 17.0/27.0

c.

Containment Purge Isolation Not Applicable f.

Auxiliary Feedvater Pumps Not Applicable g.

Service Vater ?ystem j 77.0/87.0

5.

Steam Flov in Two Steam Lines-High Coincident with T, g -Lov-Lov s.

Steam Line Isolation Not Applicable l

'6.

Steam Line Pressure-Lov a.

Safety Injection (ECCS) f 12.0/22.0

'b.

Reactor Trip-(from SI) f 2.0 c.

Feedvater Isolation f 32.0

i y

d.

Containment 1 solation-Phase "A" f 17.0I /27.0'

i I

e.

Containment Purge Isolation Not Applicable l

l-f.

Auxiliary Feedvater Pumps Not Applicable g.

' Service Vater System j 77.0'i'/87.0

h.

Steam Line Isolation f 7.0

'FARLEY - UNIT 2 3/4 3-30 AMENDMENT NO.

85 1

.REACIOR COOLANT SYSTEM HOT STANDBY LIMITING CONDITION FOR OPERATION 3.4.1.2 At least two of the Reactor Coolant Loops listed below-shall be OPERABLE and in operation when-the rod control system is operationni or at least two Reactor Coolant Loops listed belov shall be OPERABLE vith one Reactor Coolant Loop in operation when the rod control system is disabled by opening the Reactor Trip Breakers or shutting down the rod drive motor / generator sets:*

1.

Reactor Coolant Loop A and its associated steam generator and Reactor Coolant pump, 2.

Reactor Coolant Loop B and its associated steam generator and Reactor Coolant pump, 3.

Reactor Coolant Loop C and its associated steam generator and Reactor Coolant pump.

APPLICABILITY: HODE 3 ACTION:

a.

With less than the above required Reactor Coolant Loops OPERABLE, restore the required loops to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or be in HOT SHUTD0VN vithin the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

b.

Vith only one Reactor Coolant Loop in operation and the rod control system operational, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months open the Reactor Trip Breakers or shut-down the rod drive motor / generator sets, c.

With no Reactor Coolant Loops in operation, suspend all operations involving a reduction-in boton concentration of the Reactor Coolant System-and immediately initiate corrective action to return the. required coolant loop to operation.

SURVEILLANCE REQUIREMENTS 4.4.1.2.1-

-At least the above required Reactor-Coolant pumps, if not in operation, shall be determined:to be OPERABLE once per 7 days by verifying correct breaker alignments and indicated power availability.

4.4.l'.2.2 The required Reactor Coolant Loop (s) shall be verified to be in operation and circulating Reactor Coolant at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

4.4.1.2.3 The required steam generator (s) shall be determined OPERABLE by verifying secondary side water level to be greater than or equal to 10% of vide range indication at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

All Reactor Coolant pumps may be de-energized for up to I hour provided (1) no operations 6te permitted that vould cause dilution of the Reactor Coolant System boron concentration, and (2) core outlet temperature is maintained at

~1 east 10'F below saturation temperature.

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

58 85

,3/4.2 POVER DISTRIBUTION-L1HTTS-pASES

_ __The specifications of this section provide assurance _of fuel integrity during condition I (Normal' Operation) and 11 (Incidents of Moderate Frequency) events by: (a) meeting the DNB design criterion during normal operation and l

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.

I The definitions of_certain hot channel and peaking factors as used in these specifications are as follows:

F,(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.

F(11 Nuclear Enthalpy Rise llot 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.

F,(Z)

Radial Peaking Factor, is defined as the ratio of peak power density y

to average power density in the horizontal plane at core elevation Z.

3/4.2.1 AXIAL-FLUX DIFFERENCE The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the F (Z) upper bound envelope of 2.45 for VANTAGE 5 and 2.32 for LOPAR times th,e normalized l

axial peaking factor is not exceeded during either normal operation or in the event-of xenon redistribution following power changes.

Target-flux difference is determined at-equilibrium xenon conditions. The full length rods may be positioned within the core in accordance with their respective insertion limits and should be inserted near their normal position for steady state operation at high power levels.

The value of the target flux difference obtained under these conditions divided by the fraction of RATED Ti!ERHAL POWER is the target-flux difference at RATED THERMAL POVER for the

associated core burnup conditions. Target flux differences for other'THERHAL POVER levels are obtained by multiplying the RATED THERHAL POVER'value by the appropriate fractional THERHAL POVER level. The periodic updating of the target flux difference value is necessary to reflect core burnup considerations.

i i

FARLEY -' UNIT 2 B 3/4 2-1 AHENDMF.NT NO.

13, 65, 85

w 3

POVER DISTRIBUTION L1MTTS BASES AXIAL-FLUX DIFFERENCE (Continued)

Although it is intended that the plant vill be operated with the AFD vithin the 4(5)% target band about the target flux difference, during rapid plant-Ti!ERMIL POVER reductions, control rod motion vill cause the AFD to deviate outside of the target band at reduced THERMAL POVER levels. This deviation vill not affect the xenon redistribution sufficiently to change the envelope of peaking factors which may be reached on a subsequent return to RATED THERMAL POVER (with the AFD within the target band) provided the time duration of the deviation is limited.

Accordingly, a 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months penalty deviation limit cumulative during the previous 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months is provided for operation outside of the target band but within the limits of Figure (3.2-1) while at TilERMAL POVER levels between 50% and 90% of RATED TilERMAL POVER.

For TilERMAL POVER levels between 15% and 50% of RATED TilERMAL POVER, deviations of the AFD outside of the target band are less significant. The penalty of 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months actual time reflects this reduced significance.

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 avetage of each of the OPERABLE excote detector outputs and provides an alarm message immediately if the AFD for 2 or more OPERi.t_.E excore channels are outside the target band and the THERMAL POVER is greater than 90% of RATED TilERMAL POVER.

During operation at THERMAL POVER levels between 50% and 90% and between 15% and 50% RATED TilERMAL POWER, the computer outputs an alarm message when the penalty deviation accumulates beyond the limits of I hour and 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , respectively.

Figure B 3/4 2-1 shows a typical monthly target band.

3/4.2.2 and 3/4.2.3 IIEAT FLUX ll0T CllANNEL FACTOR, NUCLEAR ENTilALPY 110T CifANNU TXUITR 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 vill not' exceed the 2200*F ECCS acceptance criteria limit.

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

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

a.

Control rods in a single group move together with no individual rod insertion differing by more than 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.

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

85

.POVER DISTRIBUTION LIHlTS BASES F"

through,il vill be maintained within its limits providgd conditions a.

d.-above are maintained. The relaxation of F.11 as a function of TilERHAL POVER allows changes in the radial power shape ~for all permissible rod insertion limits.

Vhen an F measurement is taken, an allowance for both experin.cntal errorandmanu}acturingtolerancemustbemade.

An allowance of 5% is appropriate for a full co.e map taken with the incore detector flux mapping system and a 3% allowance is appropriate for manufactering tolerance.

When F" 11 is measured, experimental error must be allowed for and 4% is the appropriate allovance for a full core map taken with the incore detection system. The :pecified limit for F 11 contains an 8% allovance for uncertainties. The8%allovanceisbasedonkhefollowingconsiderations:

a.

Abnormal perturbations in the radial power shape, such as from rod misalignment, affect F",Il more directly than F,,

b.

Although rod movement has a direct influence upon limiting F to within its limit, suchcontrolisnotreadilyavailabletollmit F"a ' ""d II c.

Errors in prediction for control power shape detected during startup by restricting axial flux physics tests can be compensated for ip F,is less readily available.

distribution. This compensation for F,11 FARLEY - UNIT 2 B 3/4 2-4 AMENDHENT NO.

57, 85

" 'POVER DISTRIBUTION L1 HITS BASES The radial peaking factor F,El factor, F (Z), remains within its limit.(Z), is measured periodic assurance that the hot chann limit-for-RATED TilERHAL p0VER (FRIP)ashrovidedintheRadialPeakingFactotThe F'#

' limit report per Specification 63.1.11 was determined from expected power control maneuvers over the full range of burnup conditions in the core.

,3/4.2.4 00ADRANT p0VER TIL'l RATIO The quadrant pover 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 power operation.

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

The two hour time allovance 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 F is teinstated by reducing the maximum allowed power by 3 percent for each perce,nt of tilt in excess of 1.0.

~

For purposes of monitoring QUADRANT p0VER TILT-RATIO vhen one excore detector is inoperable, the movable incore detectors are used to confirm that the normalized symmetric power distribution is consistent with the QUADRANT POVER TILT RATIO.

The incore detector monitoring is done with a full incore flux map or two sets of four symmettic thimbles.

The two sets of four symmetric thimbles is a unique set of eight detectot locations. These locations are C-8, E-5 E-ll, H-3, H-13, L-5, L-11, and N-8.

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 value of $80,7'F is based on the average of two control board readings and,In indication urcertainty of 2.$'F.

The indicated pressure value of 2205 psig 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% flov (0.1% flov is included for feedvater venturi fouling).

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months surveillance of Tavg and pressurizer pressure through the control board readings are sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

The 18 month surveillance of the total RCS flow rate is a precision measurement-l-

that verifies the RCS flow requirement at the beginning of each fuel cycle and ensures correlation of the flov indication channels with the measured loop flows. The monthly surveillance of the total RCS flow rate is a reverification L

of the RCS flow requirement usi:, loop elbow tap measurements that are l

correlated to the precision RCS ilov measurement at the beginning of the fuel cycle. ^The 12 hour-RCS flov surveillance is a qualitative verification of significant flow degradation using-the control board indicators and the loop elbov_ tap measurements that are correlated to the precision RCS flow measurement at the beginning of each fuel cycle.

l FARLEY - UNIT 2 B 3/4 2-5 AMENDHENT No.

U, 85

.. - ~ - -.

. ~

)

3/4.4 REACTOR COOLANT SYSTEM I

BASES 3/4.4.1 REACTOR COOLANT LOOPS AND COOLANT CIRCULATION The plant is designed to operate with all Reactor Coolant Loops in operation, and meet the DNB' design criterion during all normal operations and l

anticipated transients.

In HODES 1 and 2 vith one Reactor Coolant Loop not in operation this specification requires that the plant be in at least HOT STANDBY vithin I hour.

In MODE 3, tvo Reactor Coolant Loops p: ovide sufficient heat removal l

capability for removing core heat even in the event of a bank withdraval accident;-however, a single Reactor Coolant Loop provides sufficient decay heat removal capacity-if a bank vithdraval accident can be prevented:

1.e.,

by opening the Reactor Trip Breakers or shutting dovn the rod drive motor / generator sets.

l In MODE 4, a single reactor coolant or RHR loop provides sufficient heat removal capability for removing decay heat, but single failure considerations require that at least two loops be OPERABLE. Thus, if the Reactor Coolant Loops are not OPERABLE, this specification requires two RHR loops to be OPERABLE.

In H0DE 5, single failure considerations require two RUR loops to be OPERABLE.-

The operation of one Reactor Coolant Pump or one RHR pump provides adequate flov to_ ensure mixing, prevent stratification and produce gradual reactivity changesLduring boron concentration reductions in the Reactor Coolant System.

The_ reactivity change rate associated with boron reduction vill, therefore, be within the capability of operator recognition and control.

The restrictions on starting a Reactor Coolant Pump.vith one or more Reactor Coolant System cold Icgs less than or equal to 310'F are provided to

. prevent Reactor Coolant System pressure transients, caused b, energy additions from the secondary _ system, which could exceed the limits of Appendix G to 10 CFR Part 50.

The Reactor Coolant-System vill be protected against overpressure transients and vill not exceed the limits of Appendix G by either (1) restricting the water volume in the pressurizer and thereby providing-a volume for the primary coolant to expand into, or-(2) by' restricting starting of the Reactor Coolant Pumps to when the secondary water temperature of each steam

generator is less than 50'F above each of the Reactor Coolant System cold leg temperatures.

FARLEY - UNIT 2 B 3/4 4-1 AMENDMENT NO. SS, 85 A

ADMINISTRATIVE CONTROLS Type of container (e.g., LSA, Type A Type B, large Quantity) and e.-

f.- Solidification agen_t (e.g., cement, urea formaldehyde).

lhe radicactive effluent release reports shall include unplanned releases from effluents on a quarterly basis.the site to unrestricted areas of radioactive

-PROCESS CONTROL PROGRAM (PCP) made during the tLONTHtL OPERAllNG REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, be_ submitted on_a monthly basis to the Commission, later than the 15th of each month following the calendar month covered by the report.

Monthly-Operating Report within 90 days in which the chan effective.

treatment systems shall be submitted with the Monthly Opera period in which the change was implemented.

RADIAL PEAKING FACTOR tIMIT REPORT 6.9.1;11 Tho' fxy limit for Rated Thermal Power (F3TP) for all core planes containing bank D" control rods and all unrodded core planes shall be y

established and documented in the Radial Peaking factor Limit Report before each reload cycle (prior to: MODE 2) and provided to the Commission, pursuant to 10 CFR 50.4, upon issuance.

In the event that the limit would be submitted otherwise exempted by the Commission.at some other time during core life, Anyinformationneededtosupportf[willbebyrequestfromtheNRCandneed not be included in this report.'

MNUAl' DIESEL GENLRATOR RELIABILITY DATA REPORT

\\

6.9.1.12 start on demand for each diesel generator shall be submitted to

)

annually.

Position C.3 b of NRC Regulatory Guide 1.108, Rev,, ion:1,1977.

)

FARLEY-UNIT 2 6-19 AMENDMENT NO. #, 62, 74, 85

_

ML20094L364

Text

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