Text

Proposed Tech Specs,Allowing Conversion from Westinghouse Fuel to Fuel Provided by Framatome Cogema Fuels
	ML20101N707
Person / Time
Site:	Sequoyah
Issue date:	04/04/1996
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20013A240	List: ML20101N698; ML20101N707; ML20101N731; ML20101N744;
References
	NUDOCS 9604090078
	Download: ML20101N707 (86)
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ENCLOSURE 1 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 l

DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-96-01) {

LIST OF AFFECTED PAGES +

r Unit 1 Unit 2 2-5 2-5 2-8 2-10 ;

2-9 2-11 2-10 8 2-1 ;

8 2-1 B 2-4 B 2-4 8 2-5 B 2-5 3/4 2-4 i 3/4 2-5 3/4 2-5 i 3/4 2-6 3/4 2-6 L 3/4 2-7 3/4 2-6a 3/4 2-8 3/4 2-8 3/4 2-10 3/4 2-9 )

3/4 2-11 3/4 2-10 3/4 2-12 3/4 2-11 3/4 2-13 3/4 2-14 3/4 2-16 B 3/4 1-4 8 3/4 1-4 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 !

B 3/4 2-4 B 3/4 3-2 l B 3/4 3-2 6-22 6-21 1

l s

9604090078 960404 PDR ADOCK 05000327 .

P PDR j

TABLE 2.2-1 m REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS E FUNCTIONAL UNIT S TRIP SETPOINT ALLOWABLE VALUES

$ 1. Manual Reactor Trip Not Applicable e

Not Applicable c 2. Power Range, Neutron Flux

{

Low Setpoint - < 25% of RATED THERMAL POWER -

Low Setpoint THERMAL POWER $ 27.4% of RATED R145l High Setpoint 5 109% of RATED High Setpoint R14 THERMAL POWER 5111.4% of RATED THERMAL POWER

3. Power Range, Neutron Flux, High Positive Rate 5 5% of RATED THERMAL POWER with

< 6.3% of RATED THERMAL POWER a time constant 12 second with a time constant 1 2 second

4. Power Range, Neutron Flux, High Negative Rate 5 5% of RATED THERMAL POWER with 5 6.3% of RATED THERNAL POWER a time constant 1 2 second with a time constant 1 2 second

[ 5. Intermediate Range, Neutron Flux 5 25% of RATED THERMAL POWER 5 30% of RATED THERMAL POWER

6. Source Range, Neutron Flux 5 5 10 counts per second 0 5 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3 gg 8. Overpower AT See Note 2 See Note 4 3> $

--< g 9. Pressurizer Pressure--Low 1 1970 psig

$ 2 1964.8 psig ,

m re 10. Pressurizer Pressure--High 5 2385 psig R145 5 2390.2 psig !

11. Pressurizer Water Level--High 5 92% of instrament span 5 92.7% of instrument span c3 *^

12. Loss of Flow 190% of design flow 2 89.4% of design flow per loop

per loop * ,

t

" Design flow ispf,4)0}gpa per loop.

1 y o<js (87,000 X l.035)

. _ - . - - . _ _ . - - . - . - - - _ _ _ - - _ - . _ _ . . . _ _ _ . _ - - _ _ - - - - _ _ _ - - _ . . _ _ _ _ - - - - - , _ _ . - , - L- ---. 2 - . , - - - - - - - c--. - - - - - .

. . ~ . -, .:.-.

TABLE 2.2-1 (Cc:2 Sundl REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS i

h NOTATION (Continued)

[ NOTE 1: (Continued) x M 1+rS

- The function generated by the lead-lag controller for T,, dynamic compensation R145 1+rS2 R2 n r,&7 i 2

- Time constants utilized in the lead-lag controller for T,,, r, lt 33 secs.,

t as 4 secs. ,

T - Average temperature 'F R145 T' s 578.2*F (Nominal T,, at RATED THERMAL POWER)

K3 - 0.00055 P = Pressurizer pressure, psig P' - 22351,sig (Nominal RCS operating pressure)

S - Laplace transform operator (sec)

and f of the,(AI) is a function power-range nuclearof lonthe chambers; Indicated withdifference gains to be selected between top on based and bottom detectors measured instrument response during plant startup tests such that:

a (1) hq, between 2Merdott and led,06(' 0 (where q and g e are perceh RATED THERMAL POWER in the top and bot' tom ha ves of the, core re,spectively, g and q, + g is total THERMAL POWER in percent of RATED T RMAL POWER).

e

,E QTHL* QW

$G B'

lr n .~

~

w

R

] m. J

- TABLE 2.2-1 (Continued 1 L1 a S REACTOR TRIP SYSTEN INSTRUMENTATION TRIP SETPOINTS Ei j NOTATION (Continued)

QTBI' N-NOTE 1: (Continued) QTW S R23 E! -

Z exceedsL-49 grcWW, the AT trip setpoint shall be (11) for each percent automatically that theby reduced magnitude [kt504e,rfest o_f (qof- its g,)lue va at RATED THERtAL POWER.

exceedsl+ fnefcent, the AT tr setpoint shall be (iii) for automatically each percent that the magnitude reduced byl%56Arfaittiof _o_f (q - q its )luevaat RATED THERMAL POWER. ,

G. #3 Y NOTE 2: ' Overpower AT (' * #' ) s AT, {K4 -K T - K6 (T-T") - f,( AI)} QTPL R145 1+TS3 5(1+rS 3 QTPSk

' = as defined in Note 1 m

Where:

4 1+755 r g,T3 - as defined in Note 1 AT, - as defined in Note 1 Rll8 K, s 1.087 R21 K3 2 0.02/*F for increasing average temperature and 0 for decreasing average 8- temperature

" rS3 - The function generated by the rate-lag controller for T, dynamic R145 g

. 1+T53 compensation

?-

W W

%C + .

'^'

SWcomKo

{~ QTNL, QTPL,qTN5, Ano QTPS A(Le

f:

Tt4E Co t R. Pcg specin c eno,J 4 9. /. // .

Yp ~ _ -

G~

IABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS NOTATION (Continued)

NOTE 2: (Continued)

E R215

% 73 " Tine constant utilized in the rate-lag controller for T.,, i s210 secs.

K6 2 0.0011 for T > T" and K4 2 0 for T s T*

T - as defined I,n Note 1 , ,

I" - Indicated T, at RATED THERMAL POWER (Calibration temperature for

&TinstrumenTation,s578.2*F) j S - as defined in Note 1

. TN SdAT~ n y

fgI)/ -/forpKJF ,

b NOTE 3: The channel's maximum trip setpoint shall not exceed its computed trip point Sv more than 1.9 percent AT span. j R145 NOTE 4: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 1.7 percent AT span. l 37 u .-

r*

= f QPN L, GPfL , Q PM S , Mo c} FPS )%tf SP&umt=D &

.f' * ,

SOLS Pnt. S f&csPlcA Trad Co.9. l . l f ,

g. W ru 9d

v. N

- a

4 I

j J

INSERT A T

and f2(AI) is a function of the indicated difference between tcp and bottom detectors ;

of the power-range nuclear ion chambers; with gains to be selected based on i measured instrument response during plant startup tests such that:

(i) for q, - q, between QPNL* and QPPL' f2(AI) = 0 (where q, and q,are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, l and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (q,- q ) exceeds QPNL' the AT trip setpoint shall be automatically reduced by QPNS* of its value at RATED 1 THERMAL POWER.

(iii) for each percent that the magnitude of (q,- q,) exceeds OPPL' the AT trip setpoint shall be automatically reduced by QPPS* of its value at RATED THERMAL POWER.

l l

) ( w m u. wwcs m &<w acoaccc> ro pa c t u At A .Q Tb rWL- kOD W W 2.1 SAFETY LIMITS un0Aao* wry /Mcru,4h j ' s s , !

BASES 1

x- - . _

l 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 -;

i products to the reactor coolant. Overheating of the fuel cladding is prevented l I by restricting fuel operation to within the nucleate boiling regime where the I

' 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 4

result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer i

coefficient. DNB is not a directly measurable parameter during operation and )

therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been i j related to DNB.ltfiropgh t)1e Wil5-1 @rrylatjen d the W3 cordlat4onnati R142 '

l 4 ptopcitrons Autede/the fance/of itB-Y coptel, iorXI The DNB correlations have i

been developed to predict the DN3 flux and tlie location of DNB for axially j-uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular i core location to th cal heat flux, is indicative of the margin to DNB.

Tmr

! The DNB design asis isL4s f lows: there must be at leastl 95 etc t R 42 pro abili y th the inim DN of t liitinVrod furirg Co it nI nd/

j 4

4 I even is eate than re al t theMesias(DNBlVlimit.

mit s est blist d suc th ther isla 95 percent probability w' th 95 percent e( sia DNBR BR-4 l gy ,) confidence t1at DPB will not occur when the minimum DNBR is at the design DNBR (gY limit.

I N The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum ;

4 DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at '

i the vessel exit is equal _to_the enthalpy of saturated liquid.

l f Nuis 2*.I-Q

~

j ThecurvesareMsedonanenthalpyhotchannelfactor,Fh(M/ecifi.e/in/ R159

{ lth(Cor/Opergtino/imit geport (CffLRJland a reference cosine with a peak of i ~1.55 for axial power shape. An allowance is included for an increase in

' Fh at reduced power based on the expression:

j,7o - t+MM 6a M Fh = FhPgy, p (3.p)) , wu,,3 1

where P = THERMAL POWER yz -warmswun M' l 3

j m, _ _

s R159 N

FhPYt F mit a TED ERMAL OWER TP) eci ed t j CpR, d l

PF =t power actor ultip1 r for , s cif di he 0 ./

j SEQUOYAH - UNIT 1 B 2-1 Amendment No. 19, 114, 138, 155 By letter dated December 8, 1992

INSERT B in meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically such that there is at least a 95 percent probably at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR limit. The uncertainties in the above plant parameters are used to determine the plant uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analysis using values of input parameters without uncertainties.

SAFETY LIMITS '

g, BASES S.

Range Channels will initiate a reactor trip at approximately 25 percent of RATED l THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Intermediate or i 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. l Overtemperature Delta T The Overtemperature Delta T trip provides core protection to prevent DNB for all com_binations of pressure, power, coolant temperature, and axial power t) distribution, provided that the transient is slow with respect to transit, l

thermowell, and RTD response time delays from the core to the temperature R145 kfandLowPressurereactortrips. detectors (about This setpoint 8 seconds),

includes andfor corrections pressureaxial is within the T

k ture and dynamic compensation for transport, thermowell, and RTD re R145l delays from the core to RTD output indication. With normal axial power distri-ki bution, this reactor trip limit is always below the core safety limit as shown l 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

't trip is automatically reduced according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 4 loop P-8 '

setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint.

Delta-T,, as used in the Overtemperature and Overpower AT trips, repre-sents the 100% RTP value as measured by the plant for each loop. This normalizes each loop's AT trips to the actual operating conditions existing at the time of mer.urement, thus forcing the trip to reflect the equivalent full power conditions as assumed in the accident analyses. These differences in R145 RCS loop AT can be due to several factors, e.g. , measured RCS loop flows greater than thermal design flow, and slightly asymmetric power distributions I between quadrants. While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values. Accurate determination of the loop I specific AT value should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e. , power distributions not 1 i

affected by Xenon or other transient conditions). .

Overoower Delta T 1

The Overpower Delta T reactor trip provides assurance of fuel integrity, I e.g. , no melting, under all possible overpower conditions, limits the required range for Overtemperature Delta T protection, and provides a backup to the High Neutron Flux trip. The setpoint includes corrections for changes !

l SEQUOYAH - UNIT 1 B 2-4 Amendment No. 136, 141 pg poux

6T.30

~~D)$ $ dRT'/

l 1

l l

lNSERT C l

The f i (All trip reset term in the Overtemperature Delta T trip function precludes power distributions that cause the DNB limit to be exceeded during a limiting Condition 11 event. The negative and positive Al limits at which the f i(AI) term begins ,

to reduce the trip setpoint and the dependence of f (AI) i on THERMAL POWER are determined on a cycle-specific basis using approved methodology and are specified in the COLR per Specification 6.9.1.14, 1

l l

)

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

Y Y li. Wu-TIOg;M y-* lC & T~o nATnC%L WW WO A CC OR. O IM C, To

TIM N orAr7MS tN TAisus 2. 2. - }

"T- O A C cou a a T- fc"M A ovd A sef A X s 6 g, jQ u z

_ BASES DT'T1 x A ~ ~

density and heat capacity of water with temperature, and dynamic compensation for transport, output thermowell, indication. 4 and RTD response time delays from the core to RTD of various size steam breaks as reported in WCAP-9226, to Excessive Secondary Steam Releases."

O Delta-T,, as used in the overtemperature and Overpower AT trips, repre-hJsentsthe100%RTPvalueasmeasuredbytheplantforeachloop.

Q This n normalizes h each loop's AT trips to the actual operating conditions existing at t e time of measurement, thus forcing the trip to reflect the equivalent full

!(. [i power conditions as assumed in the accident analyses. These differences in R145 RCS loop AT can be due to several factors, e.g. , measured RCS loop flows greater than between thermal design flow, and slightly asymmetric power distributions quadrants. i life, radial power redistribution betweenWhile RCS loop flows are not expected to c small changes in loop specific AT values. quadrants may occur, resulting in specific AT value should be made when performing Incore/Excore quar recalibration and under steady state conditons ;

affected by Xenon or other transient conditions)(.i.e. , power distributions not 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 pressurizar code safety valves for RCS overpressure protection, (2485 psig).and is therefore set lower than the set pressure for these valves in the event of a loss of reactor coolant pressure.The Low Pressure trip p Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to safety pressurizar retain valves.

a steam bubble and prevent water relief through the the accident analyses; however, its functional capability at the specifie of the Reactor Protection System. trip setting is required by this specifica Loss of Flow The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolant pumps.

Above 11 percent of RATED THERMAL POWER, an automatic reactor trip will occur if the flow in any two loops drops below 90% of nominal full loop flow.

R145 Above the P-8 interlock, automatic reactor trip will occur if the flow in any single loop drops below 90% of nominal full loop flow. This latter trip will prevent the minimum value of the DNBR from going below the safety analysis R142 DNBR limit during normal operational transients and anticipated transients when 3 loops are in operation and the Overtemperature Delta T trip set point is adjusted to the value specified for all loops in operation. '

R145 SEQUOYAH - UNIT 1 B 2-5 Amendment No. 138. 141 May 16, 1990

.Y f h 1000

INSERT D The f2(Al) trip reset term in the Overpower Delta T trip function precludes power distributions that cause the fuel melt limit to be exceeded during a limiting Condition 11 event. The negative and positive Al limits at which the f 2(AI) term begins to reduce the trip setpoint and the dependence of f (AI) 2 on THERMAL POWER are determined on a cycle-specific basis using approved methodology and are specified in the COLR per Specification 6.9.1.14.

h (xM, El POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F m w vmiasto utraw rac Accu rM L4 LIMITING CONDITION FOR OPERATION LiAir5 5 Mc IF,4 o tN T9er C OL. R , ,

)

3.2.2 F q shall be IM'miMd bf thy'foJTow)ng 761ptiogshjpsjj

>M )_ ( f P> .5 W /

q(z <[ _

K(z for <0 9 0.5 er F g TP the q li t at TED T RMAL OWER TP) /

s cifi in e C0 , '

p THER. P0 , and / l R ED T AL P R '

K(z f, in th orma e C0 ed F z) as fung ion of ore eigh spe '- /

/ /

APPLICABILITY: MODE 1 ACTION: Y' With F q exceeding its limit:

Repdce RMA 0WER least for e 1% F exceeds he lim 15 nutes nd simil ly red e the P er Range .eutron i

-Hi rip - tpoints next 4 ours; P0 W

/with F1 y pr eed f up to thin t OPERAT N total 9 72 hourp, subsequ t POWER ERATION g< /may have een re K

rocee provid he Ove power Del 4 T Trip tpoints value of d at 1 st 1% (i AT span or each Fq (z)

I excee the li .

(

2 l . entify corre the cau of the t of li t conditi prior s o inc ing TH MAL POW ; THERMA OWER ma then be i reased prov F s demon rated thr gh inco mapping o be wi in mit q(z it

/

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

R144 SEQUOYAH - UNIT 1 3/4 2-5 Amendment No. 19, 95, 140, 155 00T 231991

I l

INSERT E l f

a. Reduce THERMAL POWER at least 1 % for each 1 % Fo(X,Y,Z) exceeds the limit within 15 minutes, and similarly reduce the following: '

1. Administratively reduce the allowable power at each point along the AFD limit l lines within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and -

t

2. 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 . l

b. POWER OPERATION may proceed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K,) have been reduced at least 1% (in AT span) for each 1% that Fo(X,Y,Z) exceeds the limit specified in the COLR.

)

c. Identify and correct the cause of the out-of-limit condition prior to increasing >

THERMAL POWER above the reduced limit required by Action a. and b., above; l THERMAL POWER may then be increased provided Fo(X,Y,Z) is demonstrated 1 through incore mapping to be within its limits.

l 1

l

)

1 i

I l

POWER DISTRIBUTION LIMITS (f I N SC/2T* f~ j SURVEILLAN '

EMENTS (Continued) r

[_

fr ';

4.2.2.21 F shall evaluat to det ine if F is within s ;

li by:

. Usin he mova e incore etectors obtain a p er distribu- I ti map at y THER POWER gre er than 5% RATED THER L WER.

. Incre ing the easured F component the power istribution m by 3 pe ent to acc nt for manu cturing tol ances and furt nereasin the value 5% to acco t for meas ement uncertai . is.

. Sati ying the fo owing rela 'onship:

i N RTP Fq (z) _ F x K(z for P > .5 R159 PxW N P Fq (z) 5 0 x K(z) for P $ 0.5 / R159 W(z) x 0 /

wh e F"(z) q is e measured F Iincrease by the allow ces for anufacturi tolerances a TP 4, measureme uncertainty q R159 the Fq 1 it, K(z) is t normalize Fq (z) as a f ction of core ]

heigt1, 3 P is the re ive THE POWER,and W is the cycl 1 dependent functio that accou s for power stribution tr sients J ncountered.d ng normal eration. F RT , K(z), and W ) are speci-R159 fied in th "COLR as pe pecificatio 6.9.1.14.

---,e

d. _ . . .Measue ng M(z) q F/ . ording to t following sc ule:

/ Upon ach ving equilib um conditions fter exceedin y 10 per nt or more RATED THERMAL OWER, the THE L POWER at ich Fq(z) w ast determin ,

or

2. t least oncp er 31 effecti full power d s, whichev occurs fir t.

I duringpowe escalation at e beginning of each cycle, ower lev may be increase ntil a power el for extend operation s been ieved and l a power distribution obtained.

l L , R147 SEQUOYAH - UNIT 1 3/4 2-6 Amendment No. 19, 95,14q 155 00T 23191 l'

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

Wit measu ments ndicat g imum ov z Fo"(zh

,}

M determi ) either o the j has nereas since e previo ion of F f lowin etions all be t en:

R220 l

1. s 11 be in eased ov that spec led in 4. .2.2.c by th FO(z) , or appr riate fa or specif d in the be meas ed at leas once per effective f 1 power

2. g(z) sha days il 2 suc ssive maps ndicate t t maximum er z is not ncreasing.

f. ith the r ationship specified n 4.2.2.2.c ove not be g satisfie

1. culate t percent F z) exceeds ts limit by e followin expressio maxim over z

*' -1 x 10 for P t 0. 5 g1gg

,/ ,

/, r P x K(2 i

/

* .5 mum over z -1 x for P < lg359 h 0.5 , ,

Eithe of the foll ing actions a 11 be tak :

a Place the ore in an equi ibrium cc ition where he limit 4.2.2.2. is satisfied Power lev may then increased provi d the AFD lim s of Speci cation 3.2 t, are reduc or 1% 150*

AFD r each perce F (z) exc ded its li

, b. mply with th requiremen of Specif ation 3.2. for F (z) exceeding its imit by t percent c culated abo e.

/ R144 December 11, 1995 SEQUOYAH - UNIT 1 3/4 2-7 Amendment No. 19, 95, 140, 155, 216

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

POWER DISTRIBUTION LXMITS SURVEILLANCE REQUIREMENTS (Continued) -

/

(

, . Th imits cified i 4.2.2.2.c, .2.2.2.e,

s. icable ~ d 4.2.2.2.f ove l{

e not a the follo ' g core pla regions:

1. ower cor region 0 t 5 percent i clusive.

Q . Upper core regio 5 to 100 pe ent inclusi .

4.2f. . When F z) is meas ed for reas s other tha eeting the equiremen pr Specifi ion-4.2.2. an overall m sured F (z) all be obt ned from 9

power tribution p and increa gd by 3 perce to accoun t e r fur,ther increased y 5 percen 6 account f or manufa uring i measureme R l44l I

i 1

P

1 1 1 l

1 J

a j

SEQUOYAH - UNIT 1 3/4 2-8 Amendment No.19,140 Ma 11 MAY 111gfg ,1990

INSERT F F[(X,Y,2) shall be evaluated to determine if Fo(X,Y,Z) is within its limit by:

a. Using the moveable incore detectors to obtain a power distribution map

( F"(X,Y,2) o *) at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b. Satisfying the following relationship:

Fo#(X,Y,Z) s BQNOM(X,Y,Z) l where BONOM(X,Y,Z)*

represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement.

The BONOM(X,Y,Z) factors are not applicable in the following core plane regions ,

as measured in percent of core height from the bottom of the fuel: 1

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

c. If the above relationship is not satisfied, then

1. For that location, calculate the % margin to the maximum allowable design as follows:

l f \ \

F[(X,Y,2)

% AFD Margin . 1- x 100 %

, BQDES (X, Y,2) ,

f i F[(X,YZ)

% t,(bl) Margin . 1- x 100%

g BCDES (X, Y,Z) ,

where BODES (X,Y,Z)*

and BCDES(X,Y,Z)*

represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within Limiting Condition for Operation limits, and include cl!awances for the calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for F[(X,Y,Z),

because the limits include uncertainties.

BONOM (X,Y,Z), BODES (X,Y,Z), and BCDES(X,Y,Z) Data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific !

basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

l l 2. Find the minimum margin of all locations examined in 4.2.2.2.c.1 above.

AFD min margin = minimum % margin value of alllocations examined.

f 2(Al) OPAT min margin = minimum % margin value of all locations examined.

3. If the AFD min margin in 4.2.2.2.c.2 above is <0, either the following l

actions shall be taken, or the action statements for 3.2.2 shall be followed.

l (a) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the negative AFD limit lines at

! each power level by:

l Reduced AFD" = (AFD" from COLR) + absolute value of (NSLOPE^' * % x AFD min margin of 4.2.2.2.c.2) l (b) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the positivo AFD limit lines at each power level by:

i Reduced AFD""* = (AFD" from COLR) - absolute value of (PSLOPE^' *

% X AFD min margin)

4. If the f (AI) 2 min margin in 4.2.2.2.c.2 above is <0, either the following actions shall be taken, or the action statements for 3.2.2 shall be followed.

(a) Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT negative f 2(AI) breakpoint limit by:

Reduced OPAT negative 2f (Al) breakpoint limit = (f2 (Al) limit of Table 2.2-1) + absolute value of (NSLOPE f W" % x f(AI)2 min margin)

NSLOPEAFD and PSLOPE^F are the amount of AFD adjustment required to compensate for each 1 % that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 NSLOPE 8(" and PSLOPE '2(M are the amounts of the OPAT 2f (All limit adj.:stment required to compensate for each 1 % that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 I

1 l

i

3-(b) Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT positive f 2(Al) breakpoint limit by:

Reduced OPAT positive 2f (AI) breakpoint limit = (f2 (Al) limit of Table 2.2-1)- absolute value of (PSLOPE ##" % x f(Al) 2 min margin )

d. Measuring F[(X,YZ) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which Fo(X,Y,Z) was last determined,

or

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

e. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding F[(X,Y,2) > BONOM(X,Y,Z), either of the following actions specified shall be taken.

1. F[(X,Y,2) shall be increased over that specified in 4.2.2.2.a by the appropriate factor specified in the COLR, and 4.2.2.2.c repeated, or

2. F[(X,Y,2) shall be evaluated according to 4.2.2.2 at or before the time when the margin is projected to result in one of the actions specified in 4.2.2.2.c.3 or 4.2.2.2.c.4.

4.2.2.3 When Fo(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured Fo(X,Y,Z) shall be obtained from a power distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the Fo(X,Y,Z) limit specified in the COLR according to Specification 3.2.2.

NSLOPE # and PSLOPE 8# are the amounts of the OPAT f2 (AI) limit adjustment required to compensate for each 1 % that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained.

1 l

l l

POWER DISTRIBUTION LIMITS I 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - F,g (%,Y) !

LIMITING CONDITION FOR OPERATION l

3.2.3 T Nucle Enthal Hot Cha 1 Factor FAH s 1 be li ~ ed by th R142 foil ;ng rel onship:

}

g P

- i

.0 + PF f AH < F (1.0-P k wh T MAL POW 1159 P=R D THERM POWER '

RT

= The F H limit t RATED T RMAL POWE (RTP) specif' d in the F

fb C R, and L PF The p er factor ultiplier rFfg speci ed in the COL .

(

APPLIC. LITY: DE 1 A :

N Fg exc ding its it:

a. educe TH L POWER t ess than 5 f RATED THE L POWER wit 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months nd reduce t Power Range eutron Flux- gh Trip Setp nts to $ % of RATED ERMAL POWER thin the ne 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , {

b. monstrate ru in-core m ping that F N is within it imit

[

within 2 ours after ceedira the mit or reduc HERMAL P0 to les than 5% of ED THERMAL P ER within t next 2 hou , and

c. I ntify and c rect the cau of the out p limit con i ion prior i o increasi THERMAL POW above the dGced limit quired by a or b. abo ; subsequent WER OPERAT may proce provided t F

H i demonstrated rough in-co mapping e within i limit a a nominal 5 of RATED THE L POWER r to excee ng this HERMAL POWE at a nominal % of RAT HERMAL PO prior exceeding is THERMAL R and wit i 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ter att ' ing 95% or ater RATED T RMAL POWE SEQUOYAH - UNIT 1 3/4 2-10 Amendment No. 19, 138, 155 00i n li.il

.'.'.. POWER OfSTRIBUTTON LIMITS

.'g SURVEILLANCE REOUIREMENTS

~

{

i ep vision f Specif' ation 4.0. are not a icable.

I '4.j .2 2

AH sha be dete ned to be thin its mit by usi the mova k 4tcor detect s to obt n a power stributio ap:

1

/ . a. rior tJvoperation ve 75% of load 1,and TEDTHERM) POWER aft each el

b. 7At least on per 3 ctive Full ower Days.

R142 Th ma dF all be in ased by 4%Aor mea rement 3

1 SEQUOYAH - UNIT 1 3/4 2-11 MM os a Amendment No. 138

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

INSERT G Fan (X,Y) shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 1 ACTION:

With Fag (X,Y) exceeding the limit specified in the COLR:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore F asm,Y) to within the limit specified in the COLR, or

2. Reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH* % for each 1 % that Fas(X,Y) exceeds the limit, and

b. Within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore Fasm,Y) to within the limit specified in the COLR, or

2. Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH* % for each 1 % that Fa n(X,Y) exceeds that limit, and

c. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the limit specified in the COLR, either:

1. Restore Fan (X,Y) to within the limit specified in the COLR, or

2. Verify through incore flux mapping that F 3s(X,Y) is restored to within the limit for the reduced THERMAL POWER aliowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

RRH is the amount of power reduction required to compensate for each 1% that Fan (X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

ACTION: (Continued)

d. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of initially being outside the limit soecified in the COLR, reduce the Overtemperature Delta T K, term in Table 2.2-1 by at least TRH*

for each 1 % that Fas(X,Y) exceeds the limit, and

e. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b. and/or c. and/or d., above: subsequent POWER OPERATION may proceed provided that F3 s(X,Y) is demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and ;

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

1 l

l 1

- TRH is the amount of Overtemperature Delta T Ki setpoint reduction required to compensate for each 1 % that Fis(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

i

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F"JX,Y) 3 shall be evaluated to determine if Fay (X,Y) is within its limit by:

a. Using the movable incore detectors to obtain a power distribution map F"JX,Y) 3

at any THERMAL POWER greater than 5% of RATED THERMAL POWER,

b. Satisfying the following relationship:

FAHR"(X,Y) s BHNOM(X,Y)

Where:

F[(X,Y)

FA HR "(X,Y) =

MAP " I AXIAL (X,Y)

And BHNOM(X,Y)*

represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement.

MAPM is the maximum Allowable Peak *

obtained from the measured power distribution.

AXlAL(X,Y) is the axial shape for Fasm,Y).

c. If the above relationship is not satisfied, then

1. For the location, calculate the % margin to the maximum allowable design as follows:

f i F[(X, Y)

%F ag Margin = 1- x 100 %

, BHDES (X, Y) ,

where BHDES(X,Y)*

represents the maximum allowable design peaking I factor which insures that the licensing criteria will be preserved for operation within the LCO limits, and includes allowances for calculational and 1 measurement uncertainties, )

No additional uncertainties are required in the following equations for F"g(X,Y),

3 because the limits include uncertainties.

- BHNOM(X,Y) and BHDES(X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the I references in Specification 6.9.1.14.

i i

i

! I

2. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above. ;

f ,

! 3. If any margin in 4.2.3.2.c.2 above is < 0, reduce the allowable THERMAL '

I POWER from RATED THERMAL POWER by RRH* % x most negative margin

, from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the Action statements for 3.2.3 apply.

d. With two measurements extrapolated to 31 EFPD beyond the most recent ;

, measurement yielding ;

FAHR"(X,Y) > BHNOM(X,Y) :

1 either of the following actions shall be taken:

1. F)X,Y) shall be increased over that specified in 4.2.3.2.a by the appropriate factor specified in the COLR, and 4.2.3.2.c.1 repeated, or ;

2. FhX,Y) shall be evaluated according to 4.2.3.2 at or before the time when the margin is projected to result in the action specified in 4.2.3.2.c.3.

4.2.3.3 F s(X,Y) i shall be determined to be within its limit by using the incore detectors to obtain a power distribution map

a. Prior to operation sbove 75% of RATED THERMAL POWER after each fiml ,

loading, and

b. At least once per 31 EFPD.

t i

RRH is the amount of power reduction required to compensate for each 1 % that Fas(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14. ,

l l

i

. POSER OfSTRX8UTTON LIMITS 3/4.2.4 OUADRANT POWER TILT RATIO ..,.

.g LIMITING CONDITION FOR OPERATION 1

_ ,- d 3.2.4 l The QUADRANT POWER TILT RATIO shall not . exceed ras umr SPEC t ND i APPLICABILIH: MODE 1 above 50% of RATED THERMAL POWER *

ACTION:

I a.

With but lessthe than QUADRANT POWER or equal to 1.09: TILT RATIO determined to exceed

1. ;

Calculate the QUADRANT POWER TILT RATIO at least once per hou until: -

a)

Either limit, or the QUADRANT TILT RATIO is reduced to within it l

1 b)

[ t>

THERMAL POWER. POWER is reduced to less than 50% of RATE Q8! Db

2. Within 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months s:

i

(' a) f Either reduce within its limit,the orQUADRANT POWER TILT RATIO to (

t% ,

l i

(Q{l- s b)

Reduce THERMAL POWER at least 3% from RATED THER g1 g

for each excess of'1% of indicated QUADRANT POWER TILT RATIO in Flux-High[3 and similarly reduce

% rip Setpoints within the next 4 the Power Range Neutro hours.

3.

Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER w hours and reduce the Power Range Neutron Flux-High Trip setpoints to less next than or equal to 55% of RATED THERMAL POWER within the 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

4.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 50% of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at once per hour fnr 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

"See Special Test Exception 3.10.2.

MAY 0 81990 ' -

SEQUOYAH - UNIT 1 3/4 2-12 R142 '

Amendment No. 138

I POWER DISTRIBUTION LIMITS I

j ACTION: (Continued) i i

b.

With the QUADRANT misalignment POWER TILT of either a shutdown RATIO or control red: determined to exceed 1.09 due 1.

Calculate the QUADRANT POWER TILT RATIO at least once per hour until:

a)

Either its limit,the QUADRANT or POWER TILT RATIO is reduced to within b) i THERMAL POWER.

POWER is reduced to less than 50% of RATED THERMAL l

2. I Reduce THERMAL POWER at least 3% from RATED THERMAL.yPOWER for ach within 1% of30 indicated minutes,QUADRANT POWER T.!Li RATIO in excess 1of b-

3. Verify that the QUADRANT POWER TILT RATIO is within its limit b within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL A r-P0k'ER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Q[

Setpoints ri to less than or equal to 55% of RATED THERMAL POWER ithin the l

next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . p T w-

4. Identify and correct the cause of the out of limit condition

?j(

] prior to increasing THERMAL POWER; subsequent POWER OPERATION o

above 50% of RATED THERMAL POWER may proceed provided that the (%

] QUADRANT POWER TILT RATIO is verified within its limit at least( h o once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c.

i With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to 5 causes other than the misalignment of either a shutdown or control red:

1. Calculate the QUA.DRANT POWER TILT RATIO at least once per hour :

until:

> '6rio i a) Either the QUADRANT POWER TILT is reduced to within

, its limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL l POWER.

1 d :)

u a...:)

4[

I

.MAY 0 R R90 .3,,

R142 SEQUDYAH - UNIT 1 3/4 2-13 Amendment No. 138 ;g fy

\

f f...

1 3

.f.' . ' TABLE 3.2-1 l

1% -

3 f4- DNB PARAMETERS ,-

b [

$g LIMITS 9 &k.

] 7,,F 4 Loops In PARAMETER Operation s ..

l Reactor. Coolant System T,yg 3,583*F

.; y ' Pressurizer Pressure ) 2220 psia

l

,f Reactor Coolant System l 4 y Total Flow f/Ii)4dM/ R142 i i l A .: . . fiquete 3 2~I i .

d

Limit not applicable during either a THERMAL POWER ramp in exce ss of 5% RATED THERMAL POWER, physics test, or performance of su 4.1.1.3.b.

-IfIpcipesA 3g g 3 3% f1p meagiremeg unceptainyy. I R142 1

MAY 081990 sg SEQUDYAH - UNIT 1 3/4 2-16 Amendment No. 41,138

Af00 7?hS #19EC Figure 3.2-1 Flow vs. Power for 4 Loops in Operation 362000 --

A 3.5% measuremert uncertaWy for flow is included in this figure.

(100,360100) l 360000 -- :

358000 --

Acceptable Operation ,

Region p 356000 --

S T i E 354000 -- ;

y ..

.5! i

u. -

E !

I

$ 352000 --

N I li !

y350000" Unacceptable I g Operation !

j Region E 348000 --

346000 --

344000 --

(

(90,342095) i 342000 : : : . :

90 92 94 96 98 100 102 ,

Thermal Power Fraction (% of RTP) x a voynx - u a r .L % z-17

T'H4 donp2n (2co in socie ri o,a u s m , r- 1 /)Ao M %'r Dow rJ g.cg> sr>SM A.1~so J L /M t 7-5 AstG

REACTIVITY CONTROL SYSTEMS 3/2rc (Mo ffv 7- g g)t_fq jd ,Sg, .s,#g c., e c ec,4-77 g

/ (o , 9, / , /"/. I BASES g- ( L /

s / /w >

/

The ACTION statements which permit limited variations from the basic rcquirements are accompanied by additional restrictions which ensure that the original design criteria are met. Misalignment of a rod requires measurement of peaking factors and a restriction in THERMAL POWER. These restrictions /Mi$gg @ D provide assurance of fuel rod integrity during continued operation. In cddition, those accident analyses affected by a bipdlMnpfeph rod are reevalu-cted to confirm that the results remain valid during future eration.

In the event that a malfunction of the Rod Control System renders control rods R219 immovable, provision is made for continued operation provided:

o The affected control rods remain trippable, and o The individual contro'. rod alignment limits are met.

In the event that a malfunction of the Rod Control System renders control, rod b nks immovable during surveillance testing, provision is made for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of continued operation provided:

o The affected control rod banks remains trippable, o The individual control rod alignment limita are met, o A maximum of one control or shutdce. h nk is inserted no more than is steps below the insertion limit, o No reactor coolant system boron concentration dilution activities or power level increases are allowed, and o The SHUTDOWN MARGIN requirements are verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or upon insertion of controlling bank during the period the insertion limit is not met.

The requirements to preclude Reactor Coolant System boron concentration dilution, while a control or shutdown bank is below insert limits, will minimize the impact on shutdown margin.

The controlling bank (s), which is normally Control Bank D, is excluded from the 72-hour provision since insertion of this bank (s) below the insertion limit is not required for control rod assembly surveillance testing. A controlling bank is defined as any control bank that is less than fully withdrawn as defined in the COLR with the exception of fully withdrawn banks that have been inserted in accordance with Surveillance Requirement 4.1.3.1.2. This provision excludes the use o. the 72-hour allowance for control banks that can be exercised 10 steps in either direction without exceeding the insertion limits.

Checks are performed for each reload core to ensure that bank insertions of up to 18 steps will not result in power distributions, which violate the DNB criterion for ANS Condition II transients (moderate frequency transients analyzed in Section 15.2 of the UFSAR) . Administrative requirements on the initial controlling bank position will ensure that this insertion and an edditional controlling bank insertion of five steps or less will not violate the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 during the repair period. If the controlling bank is inserted more than five steps deeper than its initial position, a calculation will be performed to ensure that the SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is met. Since no dilution or power level increases are allowed, shutdown margin will be maintained as long as the controlling bank is far enough above its insertion limit to compensate for the inserted worth of the bank that is beyond its insertion limit.

The 72-hour period for a control rod assembly bank to be inserted below its (i.e., ANS Condition insertion limit restricts the likelihood of a more severe III or IV) accident or transient condition occurring concurrently with the insertion limit violation. November 21, 1995 B 3/4 1-4 Amendment No. 108, 215 SEQUOYAH - UNIT 1

3/4.~2 POWER DISTRIBUTION LIMITS BASES The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) events by: (a) maintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, and (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance Q criteria limit of 2200 F is not exceeded.

M '

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

f

[p' 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.

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

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) '

T s The limits on AXIAL FLUX DIFFERENCE assure that the g upper bound i envelope of the Fq limit specified in the COLR times the normalized axial peak- R15 ing factor is not exceeded during either normal operation or in the event of R144 xenon redistribution following power changer.

1 Prc. visions for monitoring the AFD on an automatic basis are derived from the plaat process computer through the AFD Monitor Alarm. The computer deter-nines the one minute average of each of the OPERABLE excore detector outputs and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER.

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT CHANNEL FACTORS R142 The limits on the heat flux hot channel factor and the nuclear Enthalpy rise hot channel factor ensure that 1) the design limits on peak local power density and minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temparature will not exceed the 2200*F ECCS acceptance criteria limit.

7?M PfAKitKy L m 'rS AM MMD IM

, ng c a tz ,Aga steinc Arion 6.9././4.)

_ A SEQUOYAH - UNIT 1 B 3/4 2-1 Amendment No. 19, 138, 140, 155 00T H E!

POWER OISTRIBUTION LIMITS BASES

(. .

Each of these hot channel factors is measurable but will normally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic surveillance is sufficient to insure that the limits are maintained provided:

k

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

insertion differing by more than + 13 steps from the group demand position. i J ;

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

The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained.

IA9 b

d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits F, 6,y)

+ .

The F limit as a function of THERMAL POWER allo changes in the radial power shape for all permissible rod insertion limits, will be maintained within its limits provided conditions a thru d above, are maintained. ,-

he an F easure nt it, t n, both e erimental rror and ufacturin s 'd y t era e su be all ed for. he 5% is e appropr* te allowa e for a fu l-y ore ap t en wit he inco detector ux mappi system an is the ap opri e. allo nce for nufacturi toleran .

O hen an g is a sured, e rimental ror must b allowed fo and 4% is /

e appro iate al wance fo a full co map taken ith the in re detection syste . The sp*41fied 1' t for F N also cont

s an 8% al wa ce for j u ertaint' s which an that aloperat'nwillresitinFfH P RIM !

he 8% owance based on e followi considerat' ns.

M /1.08.

a. ab al pert ations in e radial p er shape, su as from r N

isalign , effect F more dire ly than F . /

q alth gh rod move t has a di et influence pon limiti wi hin its lim" , such con 1 is not re ly availa e to F

qot R142 limi /

Fh,and

c. error n predicti for control ower shape etected d ing sta up /

ph ics te.t ca e compensa for in F y restri ing axi flux distributiop. This compe ation for g is less eadily ailabl SEQUOYAH - UNIT 1 B 3/4 2-2 Amendment No. 19, 138, 155 00i 2b hu

l l

( INSERT H When an Fo(X,Y,Z) measurement is taken, an allowance for measurement uncertainty is made. An allowance of 5% is appropriate for a full-core map taken with the incore Detector Flux Mapping System, and this allowance is included in the methodology applied to the determination of the core operating limits as described in the reference in Specification 6.9.1.14.

The hot channel factors, F"(X,YZ) o and F"g(X,y),

3 are measured periodically and compared to the nominal design values to provide a reasonable assurance that the core is operating as designed and that the limiting criteria will not be exceeded for l operation within the Technical Specification limits of Sections 2.2 (Limiting Safety l System Settings),3.1.3 (Moveable Control Assemblies),3.2.1 (AXlAL FLUX DIFFERENCE), and 3.2.4 (QUADRANT POWER TILT RATIO). An allowance is provided to account for the expected deviation between the calculation and the measurement. If the measurement is above the maximum expected value for that locatica, it is assumed to not be operating as designed, and a peaking margin evaluation is performed to provide a basis for decreasing the width of the AFD and f(AI) limits, and for reducing THERMAL POWER.

l l

1 l

4

POWER DISTRIBUTION LIMITS BASES F el r bowi redu s the ue of DN ratio. Mar n has bee ained etw n th NBR v ue us in the s ty analysi and the de gn DNBR BR-4

$ li t'to omple y off the rod ow penalty.

e val of rod b penalty i The plic eferenced n the FSAR. J i9 R

argi n exc s of the d bow pena y is avail e for plant esign R L42 fle ility y .

e hot ch nel factor q (z) is sured period ally and in ased by a i c e and h ght depend power fa or,W(z),to ovide assur ce that th l L I imit on e hot cha 1 factor, g(z), is met. (z) accoun for the e ects Di q

of no al operat ma transien and was dete ned from e cted power ontrol vers over e full ra e of burnup c ditions in e core. T W(z) RL59 q nction is ecified i he COLR.

N ,

3/4.2.4 QUADRANT POWER TILT RATIO

/ The q rant power t ratio limi assures th the ra 1 power d tri-bution s 1sfies.the de gn values u d in the p r capab ity analy .

f-Radia ower distrib on measure ts are mad during s rtup test' g and per dically durin power opera on. ('

The two ur time all ance for o ation w a tilt c dition gr ater g than 1.02 less than .09 is prov ed to al w identif tion and or-y rection a dropped misaligne od. In e event s action es not

.a corre the tilt, e margin fo uncertai y on Fn is einstate y redu ng ,

\e the ower by 3 cent from ED THE POWER f5 each perc t of ti in !

A cess of 1.0 3/4.2.5 DNB PARAMETERS ;

The limits on the DNB related parameters assure that each of the para- ;

meters 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 maintain a minimum DNBR of greater than or equal to the safety R142 !

analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their limits following load changes and other expected transient operation.

/gf r- .T SEQUOYAH - UNIT 1 B 3/4 2-4 Artendment No.19 138 155 Byletterdatedbecember8,1992

INSERT I The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such 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 QUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts. The QUADRANT POWER TILT RATIO limit specified in the CORE OPERATING LIMITS REPORT (COLR) is reflected by a corresponding peaking augmentation factor which is included in the generation of the AFD limits.

The 2-hour time allowance for operation with the tilt condition greater than the limit specified in the COLR, but less than 1.09,is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fo(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of the limit specified in the COLR.

i INSERT J The flow parameters indicated in Figure 3.2-1 have been rounded down to bias the analysis in the conservative direction.

l

INSTRUMENTATION BASES [

The measurement of response time at the specified frequencies provides assurance that the protective and ESF action function associated with each channel is completed within the time limit assumed in the accident analyses.

No credit was taken in the analyses for those channels with response times indicated as not applicable in the updated final safety analysis report.

R194 Response time may be demonstrated by any series of sequential, overlapping or total channel test measurements provided that such tests demonstrate the total channel response time as defined. Sensor response time verification may be demonstrated by either 1) in place, onsite or offsite test measurements or

2) utilizing replacement sensors with certified response times.

Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows the breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing R58 maintenance. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on a Westinghouse analysis performed in WCAP-10271, Supplement 1, which determines bypass breaker availability.

3/4.3.3 MONITORING INSTRUMENTATION 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual '

channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

3/4.3.3.2 MOVABLE INCORE DETECTORS The OPERABILITY of the movable incore detectors with the specified minimum complement of equipment ensures that the measurements obtained from use of this system accurately represent the spatial neutron flux distribution of the reactor core. The OPERABILITY of this system is demonstrated by irradiating

- each detector used and determining the acceptability of its voltage curve.

w

[For the purpose of measuring fdp opTL a full incore flux map is used.

Quarter-core flux maps, as defined in WCAP-8648, June 1976, may be used in recalibration of the excore neutron flux detection system, and full incore flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT POWER TILT RATIO when one Power Range Channel is inoperable.

3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit comparison of the measured response to that used in the

~

}~ D> n & of- fsn &> b B 3/4 3-2 Amendment No. 54, 190 SEQUOYAH - UNIT 1 November 9, 1994

ADMINISTRATIVE CONTROLS s- .

MONTHLY REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, R76 including documentation of all challenges to the PORVs or Safety Valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

( CORE OPERATING LIMITS REPORT R159 6.9.1.14 Core operating limits shall be established and documented in the CORE g OPERATING LIMITS REPORT before each reload cycle or any remaining part of a s reload cycle for the following:

g, Y. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, >

J, g. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, 4 * ,8'. Control Bank Insertion Limits for Specification 3/4.1.3.6, rn wir m A. cerm. mTsots its for cification 3/4.2.1, g, f.

4e a or R220 f, ,f[. Heat Flux Hot Channel Factor ( K(z)ff z e ee /

ce tVf tHe pot;afht4K1 c}dfcre/syin

' ication 3/4.2.2, i

s ei2"1 cedlfor R159 7, /. Nuclear Enthalpy Hot ann 1 Factor btidAIowe( Faurfor )t61tig" lier l for Specification 3/4.2.3 r 6.9.1.14.a The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by NRC in: /

{ "

INGHO aE RELO SAFE ALUA ON METH OLOGY",

4 1

- 272- -A,

/ Ju 198r (H Pr rieta .

ion 3 .1.3 - oderato Temperat e

( thodo gy for pecifi

{b

/ oeffi ent, 3 .3.5 - utdo ank I ertion Axial ux mit, 3.1 .6 -

Diffe nee,

/

Cent 1B nserti Limit 3.2.1 df t Ch Fact and 3 .3 - Nuc ar

$. Y .

3.".2 - H Flux thalpr et Cha el Fac r.)

T R220 W -102 -P-A Re sien 1 , "RE TION OF ONSTANT .IAL OF ONTRO FqSUR LLANCE CHNI SPECIF TION", BRUARY 994 3

0 (H P prieta ,

p159 (Meth ology f Spec cation .2.1 - al F1 ifferene t 2 and 3. - Hea lux Hot annel f 0 (Re ed Ax Offs Contro

).)

bb F- tor (W surv, lance quireme s for Faddethodol /

hh' N%

,[ WCAP-10266-P-A, Rev. 2, "THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (H Proprietary).

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel 4h Factor),

t y POPTIN MORE GATIVE t @0 4. W -1363 -A, "S TY EV ATION I.175 ODERAT TEMPE .E CO ICIENT CHNI oPECIF TION FO THE

- ) SEQUO NU PLANT ' MAR 993 (H opriet ). /

4 (Method ogy fo Specif ation 3. .1.3 - derator empera re j

( Coeff ient) 4 December 11, 1995 6-21 Amendment No. 52, 58, 72, 74, 117, SEQUOYAH - UNIT 1 152, 155, 156, 171, 216

INSERT K

1. f i(AI) limits for Overtemperature Delta T Trip Setpoints and f2 (AI) limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

INSERT L

1. BAW-10180P-A, Rev.1, "NEMO - NODAL EXPANSION METHOD OPTIMlZED",

March 1993. (FCF Proprietary)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

2. BAW-10169P-A,"RSG PLANT SAFETY ANALYSIS - B&W SAFETY ANALYSIS METHODOLOGY FOR RECIRCULATING STEAM GENERATOR PLANTS",

October 1989. (FCF Proprietary)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

3. BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs, June 1989. (FCF Proprietary)

(Methodology for Specification 2.2.1,- Limiting Safety System Settings

[f i(AI), f 2(AI) limits), 3/4.1.3.5 - Shutdown Bank Insertion Limits, 3.1.3.6 -

Control Bank Insertion Limits,3/4.2.1 - Axial Flux Difference , 3/4.2.2 - Heat Flux Hot Channel Factor,3/4.2.3 - Nuclear Enthalpy Rise Hot Channel Factor,3/4.2.4 - Quadrant Power Tilt Ratio.)

4. BAW-10168P-A, Rev.2, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (SER expected March 1996).

(FCF Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

5. BAW-10168P-A, Rev. 3, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (SER expected June 15,1996).

(FCF Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

6. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985. (W Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

TABLE 2.2-1 .

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS TRIP SETPOINT ALLOWABLE VALUES j FUNCTIONAL UNIT

1. Manual Reactor Trip Not Applicable Not Applicable

[

z R132

2. Power Range, Neutron Flux Low Setpoint - s 25% of RATED Low Setpoint - s 27.4% of RATED Q THERMAL POWER ro THERMAL-POWER High Setpoint - s 109% of RATED High Setpoint - s 111.4% of lR132 THERMAL POWER RATED THERMAL POWER R36

3. Power Range, Neutron Flux, s 5% of RATED THERMAL POWER with s 6.3% of RATED THERMAL POWER a time constant 2 2 seconds with a time constant 1 2 seconds High Positive Rate s 6.3% of RATED THERMAL POWER R36

4. P'ower Range, Neutron Flux, s 5% of RATED THERMAL POWER with with a time constant 2 2 seconds High Negative Rate a time constant 2 2 seconds

5. Intermediate Range, Neutron s 25% of RATED THERMAL POWER s 45.20% of RATED THERMAL POWER lR177

'?

Flux s 10' counts per second s 1.45 x 105counts per second lRI?7

6. Source Range, Neutron Flux See Note 1 See Note 3

7. Overtemperature AT See Note 2 See Note 4 R132

8. Overpower AT 2 1970 psig 2 1964.8 psig R203 E. 9. Pres:urizer Pressure--Low R s 2385 psig s 2390.2 psig l 5 10. Pressurizer Pressure--High R132 s 92.7% of instrument span k 11. Pressurizer Water Level--High s 92% of instrument span w 2 90% of design flow per loop

2 89.4% of design flow per loop

hl2.LossofFlow.

i C' 0

~ .

1s '

Design flow isNK40M gpa per loop.

~

8 70,045 ( 87, 000 X l.035) e - -

1

l TABLE 2.2-1 (Continued) -

M y REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 7x i NOTATION (Continued) i E 1 % .

t

" ,v.

NOTE 1: (Continued)

S = Laplace transform operator. sec f and fy(AI) is a function of the' indicated difference between top and bottom detectors l of the power-range nuclear ion chambers; with gains to be se lected based on measured instrument response during plant startup tests such th GTML QTtJL+ (1) betweenV/2jlf pdrgfr)t' land 4['/pddnd f (AI) = 0 (where qt and qb

, . for qt ~9b 1 4 are percent RATED THERMAL POWER in the top and bottom halves of the core respe ively. R21 and qt + 9b is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (qt ~ A b) exceedslpf!Vpfp6 the AT trip set-GTNS+ point shall be automatically reduced bylf.)G E6rs6o(1of its value at RATED THERMAL POWER.

1 '

(iii) for each percent that the magnitude of (qt ~ 9 )bexceedsf3 Afe,rt;st>ti, the AT trip set- '

point shall be automatically reduced byLfK)6MerceiA of its value a ATED THERMAL POWER.

QTPS

psyf C b NOTE 2: Overpower AT (1 + ,4r S) < AT,{K4 -K 3 ) T -K6 [T - T"] - f 2(OI)} -

O @& 1+rS5 5 (1 + r3 S R132 C3 e o 5h where:

1+v54

= as defined in Note 1 O' " IT

" 1+TSS

' :s e L; L L QTN L, grPL , (?TN S , AND CPTfS Aec Vaimo M TM ;

w e

~ 8 CotA Ake cAccimenrw 4 7.1.14. >

u ',}

TABLE 2.2-1 (Continuedi REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS j NOTATION (Continuedi l

E Z HOTE 2: (Continued)

N

- as defined in Note 1 R132 7,r3 4

AT, - as defined in Note I c ato4 Kg s 1.087 R201 K3 2 0.02/'F for increasing average temperature and 0 for decreasing average l temperature

3

? - The function generated by the rate-lag controller for T, dynamic R132 L+TS3 compensation r3 - Time constant utilized in the rate-lag controller for T,, r3110 secs.

K6 2 0.0011 for T > T" and K,10 for T s T" p T - as defined in Note 1 18 i T" - Indicated T, at RATED THERMAL POWER (Calibration temperature for ATinstrumenYation,'s578.2*F) 6

(!

E ld W T~ d S - as defin'ed in Note 1

.e 0

y s4/Jg)/=/0/op/a)X SC

~

a

& cpm t- , G tvu , GtN s , M o Qf'PS yx sMe >Fuso

}.5 run GoLK. pa % w.e.maa C 9. /. / Y.

iG i~

INSERT A ;

and f2(AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ion chambers; with gains to be selected based on measured instrument response during plant startup tests such that:

(i) for q,- q, between OPNL* and QPPL* f,(Al) = 0 (where q, and q,are percent RATED THERMAL POWER in the top and bottom halves of the core respectively, and q, + q, is total THERMAL POWER in percent of RATED THERMAL POWER).

(ii) for each percent that the magnitude of (q,- q,) exceeds OPNL* the AT trip setpoint shall be automatically reduced by QPNS* of its value at RATED THERMAL POWER.

(iii) for each percent that the magnitude of (q,- q,) exceeds QPPL* the AT trip setpoint shall be automatically reduced by QPPS* of its value at RATED THERMAL POWER.

l l

(syom,~u vs.ucs H w a 644^) teaoucAo s ro ine Lu ce 4 4 g ,,m. ieo o powocre s 2.1 SAFETY LIMITS dm

)

s' l 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 l products to the reactor coolant. Overheating of the fuel cladding is prevented by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is slightly above the coolant saturation temperature.

Operation above the upper boundary of the nucleate boiling regime could result in excessive cladding temperatures because of the onset of departure from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer coefficient. DNB is not a directly measurable parameter during operation and therefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB.Miro Icshd Wions outsid the cdnce Af W85-106rre)atiori.h thJfWRBf1 coJtelgt'io have The DNB correlations /and fie W-3 hootel R130 been developed to predict the DNB flux and the location of DNB for axially uniform and non-uniform heat flux distributions. The local DNB heat flux ratio, DNBR, defined as the ratio of the heat flux that would cause DNB at a particular core location to the local heat flux, is indicative of the margin to DNB.

C f.% The DNB desian basis isl foll s: Ithere must be at leastI 95 erce R13 L , L, pp6bab ity t t the nim NBR the mitisig roVdurinjfcon ti I d JI e nts i great than equ to t de Van DNBR limiX. e sit NN BR-li is tabli ed suc that ere la 95 percent probability with 55 percent

} confidence Inat DNB will not occur when the minimum DNBR is at the design DNBR limit.

The curves of Figure 2.1-1 show the loci of points of THERMAL POWER, R104 Reactor Coolant System pressure and average temperature for which the minimum DNBR is no less than the safety analysis DNBR limit, or the average enthalpy at R130 the vessel exit is u 1 to the enthalpy of saturated liquid. ,

of Pt 6 u,6 Z. /- (bSL .f l The curves are based on an enthalpy hot channel factor, Fh /s/ec/fi/d M R146 l 1%heXor/Opdatio6 L%it geport (C9tR)/ and a reference cosine with a peak of

~1.55 for ax al power shape. An allowance is included for an increase in g Fh at reduced power based on the expression:

D ,

.3 /.70 - ApM K-dw M' ,

Fh = FhPgy, (1.p))

'#" # 1.n - u sr u ' 46 where P = THERMAL POWER '

RATED THERMAL POWER i FRTP = t[eFg limi at RAT THER POWER P) s cifie in the COL , an 3 the wer fac r mult lier fo Fhsp ified n the LR/

p SEQUOYAH - UNIT 2 B 2-1 Amendment No. 21, 104, 130, 146 By letter dated December 8, 1992

l l lNSERTB in meeting this design basis, uncertainties in plant operating parameters, nuclear and thermal parameters, and fuel fabrication parameters are considered statistically l such that there is at least a 95 percent probably at a 95 percent confidence level that the minimum DNBR for the limiting rod is greater than or equal to the DNBR fimit. The uncertainties in the above plant parameters are used to determine the plant uncertainty. This DNBR uncertainty, combined with the correlation DNBR limit, establishes a design DNBR value which must be met in plant safety analysis using values of input parameters without uncertainties.

l l

aim w.

LIMITING SAFETY SYSTEM SETTINGS i

s BASES Intermediate and Source Range, Nuclear Flux (Continued)

Range Channels will initiate a reactor trip at approximately 25 percent of R129 RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated with either the Inter-mediate 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 l System.

Overtemperature AT The Overtemperature Delta T. trip provides core protection to prevent DNB for all combinations of pressure, power, coolant temperature, and axial power dis-tribution, provided that the transient is slow with respect to transit, thermo-w D (ell, and RTD response time delays from the core to the temperature detectors R132 about 8 seconds), and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for axial power k distribution, changes in density and heat capacity of water with temperature k?anddynamiccompensationfortransport,thermowell,andRTDresponsetimedelays I- ,

S from the core to the RTD output indication. With normal axial power distribu- F gFigure2.1-1. tion, this reactor trip limit is always below the core safety limit as( shown i If axial peaks are greater than design, as indicated by the dif-

, ference between top and bottom power range nuclear detectors, the reactor trip s automatically rc M ed according to the notations in Table 2.2-1.

Operation with a reactor coolant loop out of service below the 4 loop P-8 setpoint does not require reactor protection system setpoint modification because the P-8 setpoint and associated trip will prevent DNB during 3 loop operation exclusive of the Overtemperature Delta T setpoint.

Delta-T,, as used in the Overtemperature and Overpower AT trips, represents the 2

100 percent RTP value as measured by the plant for each loop. This normalizes each loop's AT trips to the actual operating conditions existing at the time of measurement, thus forcing the trip to reflect the equivalent full power condi-tions as assumed in the accident analyses. These differences in RCS loop AT can be due to several factors, e.g., measured RCS loop flows greater than ther-mal design flow, and slightly asymmetric power distributions between quadrants.

While RCS loop flows are not expected to change with cycle life, radial power redistribution between quadrants may occur, resulting in small changes in loop specific AT values. Accurate determination of the loop specific AT value should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by xenon or other transient conditions.).

SEQUOYAH - UNIT 2 8 2-4 Amendment No. 129, 132 OCT 31. .]

INSERT C The f i (AI) trip reset term in the Overtemperature Delta T trip function precludes power distributions that cause the DNB limit to be exceeded during a limiting Condition 11 event. The negative and positive Al limits at which the f i (All term begins to reduce the trip setpoint and the dependence of f,(Al) on THERMAL POWER are determined on a cycle-specific basis using approved methodology and are specified in the COLR per Specification 6.9.1.14.

i l ,

l

1 i

1 l

l l 1 I

l 1

l l

[

t i

1 4

l l

T~t% s<r pow r o s Auro m c M y AtCMfo Accott o,84 7v 7We AHOov-nrvou e to -meux 2.2 -1 TC> A %Qu M T= /~ot V4A bd- s% I A g. ;

/ "L.L 4 x

~Di FACAAWc4 r.

LIMIT r:6 5Mti M GS

. AM% OoM De5 d'6mo^)

' BASES I

Overpower AT The Overpower Delta T reactor trip provides assurance of fuel int rity, e.g.,

no melting, under all possible overpower conditions, limits the rt quired range for Overtemperature Delta T protection, and provides a backup to .he High Neu-tron Flux trip. The setpoint includes corrections for changes inldensity and heat capacity of water with temperature, and dynamic compensation for transport, t thernowell, and RTD response time delays from the core to the RTD output indication.J 0

he Overpower Delta T trip provides protection to mitigate the consequences of various size steam breaks as reported in WCAP-9226, " Reactor Core Response to Excessive Secondary Steam Releases."

O R13:

Delta-T,, as used in the Overtemperature and Overpower AT trips, represents the l )k 100 percent RTP value as measured by the plant for each loop. This normalizes '

each loop's AT trips to the actual operating conditions existing at the time of

j measurement, thus forcing the trip to reflect the equivalent full power condi-

\ tions as assumed in the accident analyses. These differences in RCS loop AT can be due to several factors, e.g. , measured RCS loop flows greater than ther-mal design flow, and slightly asymmetric power distributions between quadrants.

While RCS loop flows are not expected to change with cycle life, radial power j redistribution specific between quadrants may occur, resulting in small changes in loop AT values. !

Accurate determination of the loop specific AT value should be made when performing Incore/Excore quarterly recalibration and under steady state conditions (i.e., power distributions not affected by xenon or other transient conditions,).

Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and is therefore set lower than the set pressure for these valves (2485 psig). The Low Pressure trip provides protection by tripping the reactor in the event of a loss of reactor coolant pressure. j Pressurizer Water Level The Pressurizer High Water Level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level to a volume sufficient to retain a steam bubble and prevent water relief through the pressurizer safety valves. No credit was taken for operation of this trip in the accident analyses; however, its functional capability at the specified trip setting is required by this specification to enhance the overall reliability of the Reactor Protection System.

October 31, 1990 SEQUOYAH - UNIT 2 B 2-5 Amendment No.132

~~

00iE' .

INSERT D The f2(AI) trip reset term in the Overpower Delta T trip function precludes power distributions that cause the fuel melt limit to be exceeded during a limiting Condition ll event. The negative and positive Al limits at which the f 2(Al) term begins to reduce the trip setpoint and the dependence of f (AI) 2 on THERMAL POWER are determined on a cycle-specific basis using approved methodology and are specified in the COLR per Specification 6.9.1.14.

i

POWER DISTRIBUTION LIMITS '

3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F

,r;igiamiaa w wena 77n AccrpAbut LIMITING CONDITION FOR OPERATION L'$'"*j I'I ' I

j 3.2.2 F q shai1 be ><ini{edy/fhVfo}<ogiflodeWipr(s'httipf (x;y,a) z) _ [F [K(z[forP 0.5 F( < [F ] [K

] for < 0.5

0. /

R146 ere F RTP = the RATED T q limit MAL P R (RTP) speci ed in t COLR,

/

p= THERM POWER , and RAT THERMAL OWER i

/

K(z) the nor lized F as a unction gf core ight sp ci- /_

fip in the LR. / /f APPLICABILITY: MODE 1 '

ACTION: (X;h&)

With F exceeding its limit:

fiedu THE POWE at leas 1% for ch 1% F excee the lim

yL /a. w hin lux-inut and si larly r uce the P er Ran Neutron h Tri etpoi s withi he next ours; P ER OPER ON Ni$ may rocee or up a tota of 72 hou ; subse ent POWE PERATIO

,/

R21 y proc d prov ed the 0 rpower D a T Tri Setpoint (value of I~ j 4) h e been duced a east 1% n AT sp for eac 1% Fq (z) y e eds t limit,

b. Ident y and co ect the c se of t out of 1 it condi 'on pri V to nereasin HERMAL P R; THER POWER then be ncrea d #

rovided F z) is dem strated rough in re mappin to be ithin its limi .

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

SEQUOYAH - UNIT 2 3/4 2-4 Amendment No. 21, 95, 131, 146 March 30, 1992

INSERT E

a. Reduce THERMAL POWER at least 1 % for each 1 % Fo(X,Y,Z) exceeds the limit within 15 minutes, and similarly reduce the following:

1. Administratively reduce the allowable power at each point along the AFD limit lines within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , and

2. 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 .

b. POWER OPERATION may proceed for up to 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months . Subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K 4) have been reduced at least 1% (in AT span) for each 1% that Fo(X,Y,Z) exceeds the limit specified in the COLR.

c. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced limit required by Action a. and b., above; THERMAL POWER may then be increased provided Fo(X,Y,Z) is demonstrated through incore mapping to be within its limits.

I I

POWER DISTRIBUTION LIMITS f::-

tinued)

SURVEILLANCE .

/

4.2.2.2I F z) sha be eva ated to d ermine

Fq (z) is w hin its li by:

, Us g the m able inc e detec s to obtai power dis ibu-on map any THE AL POWE greater tha % of RATED ERMAL POWER.

[ R21

[' . Inc asing th measure g(z) compo nt of the p er distribut 3 ap by 3 p cent to ccount for nufacturin olerances an further increasi g the va e by 5% to count for asurement unc tainties.

c. Sat' fying th following r ationship:

N F )5F RTP x (z) for > 0.5 P (z)

M RTP R146 Fq (z) _F x K( for P 5,O.

W(z) 0.5 l

N ere Fq ('z) i the measured q(z) increased y the allowa es for manufactur' g tolerances d measurement neertainty, q is R146 the F 1mit, K(z) is e normalized g(z) as a fun ion of core hei t, P is the r ative THERMAL WER,and W(z) s the cycle d endent functi that account for power dis ibution tra lents I encountered d ng normal op ation. F qRTP (z), and W( are speci- R146 fied in th COLR as per S cification 6. 1.14.

M

d. Measur' g Fg (z) acco ing to the f owing sched e:

1. Upon achievi equilibrium onditions af r exceeding b 10 percent r more of RA THERMAL P0 R, the THERMA POWER 21 at whic q(z) was las determined,* r

2. At ast.once per effective f 1 power days, hichever o urs first.

"Dur g power escal ion at the b inning of each ycle, powe level ma e i reased until power level r extended ope tion has b n achiev and /

power distri tion map obt ned.

SEQUOYAH - UNIT 2 3/4 2-5 Amendt.1ent No. 21, 95, 131, 146 March 30, 1992

POWER-DISTRIBUTION LIMITS

),

SURVEILLANCE REQUIREMENTS (Continued)

R21

e. th me urements ndicatin maxim over z ha nereasect ince the p vious dete ination of 0(z) either of O e fo n@ actions s 11 be take e increase over that ecified in 4. .2.c by

1. 'F (z) shal R)2 6

/

the app priate fac r specified n the COLR, o R21 least once p 7 effective fu

2. g(z) shall e measured power a until ccessive maps - dicate that

c increasing.

imum over is With the ationships a cified in 4.2.2 .c above not be g f.

satisfi :

1.

Calculate e eeds its limit b the [

following theression:

ercent FO(z) A .<

x 100 for P 2 0. 5 maximum ov z -1 , lR14G x (z) r P i

* -1 x 10 for P 4 0.5 lg14 ;

maximum o er z ,

x )

( 0.5 s ions shall be ken:

2. Either [ the following a

a. lace the core an equilibrium ondition whe Power leve may then the R[

limit in 4.2. .c is satisfie .

be increase provided the limits of Sp ification r each percen F lR146 3.2.1 are educed 1% AFD Q(z) exceede its limit, or ,

b. Comp with the requ ements of Spe fication 3 '.2 for limit by th ercent cal ated R21 F z) exceeding i above. /

~

December 11, 1995 SEQUOYAH - UNIT 2 3/4 2-6 Amendment No. 21, 95, 131, 146, 206

T POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued)

The mits sp ified i .2.2.2.c, .2.2.2.e, nd 4.2.2. .f above 7

@ ar not app cable i he follow g core p e regions

1. L er core gion 0 to cercent nelusive.

/ 2. Upper re region to 100 rcent incl ive.

(

4.2 .3 Whe g(z) is easured f reasons her than eting the equireme s

/ Specif ation 4 .2.2 an av all meas ed F9 (z) s 11 be obt ned from /

power istribut n map and 'ncreased 3 percen o account or manuf turing tol ances or urther in eased by percent t account f measure nt certaint . h131

/

l l

l l l l

l l

R131 l

SEQUOYAH - UNIT 2 3/4 2-6a Amedmer.t No. 21, 95 . 131 OCT 20 090 l \

i i

INSERT F t i

F[(X,Y,2) shall be evaluated to determine if Fo(X,Y,Z) is within its limit by:

l i

a. Using the moveable incore detectors to obtain a power distribution map i

( F"(X,Y,Z) a *) at any THERMAL POWER greater than 5% of RATED THERMAL -l POWER.

l

b. Satisfying the following relationship: !

F"(X,Y,2) o s BONOM(X,Y,Z) l where BONOM(X,Y,Z)*

represents the nominal design increased by an !

allowance for the expected deviation between the nominal design and the j measurement.

i The BONOM(X,Y,Z) factors are not applicable in the following core plane regions i as measured in percent of core height from the bottom of the fuel: :

1. Lower core region from 0 to 15%, inclusive.

2. Upper core region from 85 to 100%, inclusive.

t

c. If the above relationship is not satisfied, then !

1. For that location, calculate the % margin to the maximum allowable design as follows:

f i l F[(X,Y,Z)

% AFD Margin . 1- x 100 % .

1 BQDES (X,Y,2) ; ;

f 1 F[(X, YZ)

% f(bl) Margin . 1- x 100% ,

g BCDES (X,Y,Z) ,

where BQDES(X,Y,Z)*

and BCDES(X,Y,Z)*

represent the maximum allowable design peaking factors which insure that the licensing criteria will be preserved for operation within Limiting Condition for Operation limits, and include allowances for the calculational and measurement uncertainties.

No additional uncertainties are required in the following equations for F[(X,Y,Z),

because the limits include uncertainties.

BONOM (X,Y,Z), BODES (X,Y,Z), and BCDES(X,Y,Z) Data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

i i

2. Find the minimum margin of all locations examined in 4.2.2.2.c.1 above. ;

AFD min margin = minimum % margin value of alllocations examined, l i

f 2(AI) OPAT min margin = minimum % margin value of alllocations i examined, j

(

3. If the AFD min margin in 4.2.2.2.c.2 above is <0, either the following '

actions shall be taken, or the action statements for 3.2.2 shall be followed. ;

I (a) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the negative AFD limit lines at l eacli power level by: l Reduced AFDu"* = (AFDu"'" from COLR) + absolute value of (NSLOPE^' * % x AFD min margin of 4.2.2.2.c.2)

(b) Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , administratively reduce the positive AFD limit lines at l each power level by: !

I Reduced AFD""" = (AFD""'" from COLR) - absolute value of (PSLOPE^'D* !

% X AFD min margin) ,

4. If the f (All2 min margin in 4.2.2.2.c.2 above is <0, either the following l actions shall be taken, or the action statements for 3.2.2 shall be followed. [

(a) Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT negative f,(AI) breakpoint limit by:

Reduced OPAT raegative2 f (Al) breakpoint limit = (f2 (AI) limit of !

Table 2.2-1) + absolute value of (NSLOPE 2# #" % x f(AI) 2 min margin) ,

1 NSLOPE^' and PSLOPE^' are the amount of AFD adjustment required to compensate for each 1 % that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 NSLOPE '2" and PSLOPE '2# are the amounts of the OPAT f2 (AI) limit adjustment required to compensate for each 1 % that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14 1

1

^

l l

J

(b) Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months , reduce the OPAT positive f2 (Al) breakpoint limit by:

Reduced OPAT positive2f (Al) breakpoint limit = (f2 (All limit of Table 2.2-1)- absolute value of (PSLOPE ##** % x f (6/)

2 min margin )

d. Measuring F[(X,Y,2) according to the following schedule:

1. Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which Fo(X,Y,Z) was last determined,* *

or

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

e. With two measurements extrapolated to 31 EFPD beyond the most recent measurement yielding F[(X,Y,Z) > BQNOM(X,Y,Z), either of the following actions specified shall be taken.

1. F[(X,Y,2) shall be increased over that specified in 4.2.2.2.a by the appropriate factor specified in the COLR, and 4.2.2.2.c repeated, or

2. F[(X,Y,Z) shall be evaluated according to 4.2.2.2 at or bef'.ere the time when the margin is projected to result in one of tha actions specified in 4.2.2.2.c.3 or 4.2.2.2.c.4.

4.2.2.3 When Fo(X,Y,Z) is measured for reasons other than meeting the requirements of Specification 4.2.2.2 an overall measured Fo(X,Y,Z) shall be obtained from a power distribution map, increased by 3% to account for manufacturing tolerances and further increased by 5% to account for measurement uncertainty, and compared to the Fo(X,Y,Z) limit specified in the COLR according to Specification 3.2.2.

- f NSLOPE (#0 and PSLOPE 2(^0 are the amounts of the OPAT 2f (AI) limit a Jjustment required to compensate for each 1 % that Fo(X,Y,Z) exceeds the limit provided in the COLR per Specification 6.9.1.14

- - During power escalation at the beginning of each cycle, power level may be increased until a power level for extended operation has been achieved and power distribution map obtained.

l2lSG POWER DISTRIBUTION LIMITS 3/4.2.3 NUCLEAR ENTHALPY HOT CHANNEL FACTOR - Fis b,d Q.

('.

LIMITING CONDITION FOR OPERATION R130

_3.2.3 he Nucl r Enthalpy ot Channe Factor, g shall e limite by the wing rel ionship:

' /fol F H FhP[1.0 PFaH (1. -P)]

3 THERMAL OWER R146 where P = RATED TH L POWE P = Th limit t RATED T RMAL P0 (RTP) sp cified in e AH k5 / LR, and

/

[ PF 3

- The pow factor m tiplier rF H spe fied in t COLR.

l EDLICAB ITY: MO 1

\

l ACTI

[ithF H xceeding

s limit:

q Reduce ERMAL P R to less han 50% of RATED THE L POWER wi in 2 ho s and re ce the Pow Range Neufron Flux 'gh Trip Se oints to 55% of R ED THERMA POWER wit n the next hours, b, emonstra thru in- re mappin that F i within its imit P ao ;

i H

within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months af er exceed g the limi or reduce ERMAL POWER to 1 s than 5% RATED T MAL POWER ithin the xt 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months d

c. I ntify and orrect th. cause of e out of l' it conditi prior oincreas'gTHERMAL/0WERabov he reduce limit requir by a.

or b. ab e; subse udnt POWER 0 RATION ma -proceed pro ded that

/ ,

F H

i demonstra d through -core map 'ng to be wi in its limi at nominal of RATED HERMAL P0 prior to ceeding thi T RMAL POW , at a nom al 75% of TED THERMA OWER prior o xceeding is THERMA POWER and ithin 24 ho s after att ning /

95% or ater RATE HERMAL P0 R.

/

3.r SEQUOYAH - UNIT 2 3/4 2-8 Amendment No. 21, 130, 146 March 30, 1992

POWERDISTRIBdTIONLIMITS SURVEILLANCE REQUIREMENTS g pr ons of Speci on 4.0.4 ar applicab .

N 4. 2. g shall be reined to b in its 1 y usin ovable -

n-core dete to obtain er distr on map:

p . Prior eration a % of R ERMAL P0 er each fuel eL30 ng,

b. At sa once Effective ower Days, an -'

N

. The red F e increased b or measurement

~

f uncer,t3 in .

/

l l

i l

l 3/4 2-9 Amendment No.130 SEQUOYAH - UNIT 2 00T 0 2 '.7.9

INSERT G j Fas(X,Y) shall be maintained within the limits specified in the COLR.

APPLICABILITY: MODE 1 ACTION:

With Fas(X,Y) exceeding the limit specified in the COLR:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

1. Restore Fas(X,Y) to within the limit specified in the COLR, or l

2. Reduce the allowable THERMAL POWER from RATED THERMAL POWER at least RRH* % for each 1 % that F3s(X,Y) exceeds the limit, and

b. Within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months either:

1. Restore Fas(X,Y) to within the limit specified in the COLR, or

2. Reduce the Power Range Neutron Flux-High Trip Setpoint in Table 2.2-1 at least RRH* % for each 1% that F3s(X,Y) exceeds that limit, and

c. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of initially being outside the limit specified in the COLR, either:

1. Restore F 3s(X,Y) to within the limit specified in the COLR, or

2. Verify through incore flux mapping that Fasm,Y) is restored to within the limit for the reduced THERMAL POWER allowed by ACTION a.2 or reduce THERMAL POWER to less than 5% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l

RRH is the amount of power reduction required to compensate for each 1 % that F3s(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14. ;

l l l ACTION: (Continued) i d. Within 48 hours5.555556e-4 days 0.0133 hours 7.936508e-5 weeks 1.8264e-5 months of initially being outside the limit specified in the COLR, reduce l

the Overtemperature Delta T K, term in Table 2.2-1 by at least TRH*

for each -

1 % that F3 s(X,Y) exceeds the limit, and

e. Identify and correct the cause of the out-of-limit condition prior to increasing THERMAL POWER above the reduced THERMAL POWER limit required by ACTION a.2 and/or b. and/or c. and/or d., above: subsequent POWER OPERATION may proceed provided that Fas(X,Y) is demonstrated, through incore flux mapping, to be within the above limit prior to exceeding the following THERMAL POWER levels:

1. A nominal 50% of RATED THERMAL POWER,

2. A nominal 75% of RATED THERMAL POWER, and

3. Within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months of attaining greater than or equal to 95% of RATED THERMAL POWER.

l l

- TRH is the amount of Overtemperature Delta T K setpoint reduction required to 3

compensate for each 1 % that Fas(X,Y) exceeds the limit provided in the CCLR per Specification 6.9.1.14.

l l

I l

l l

4.2.3.1 The provisions of Specification 4.0.4 are not applicable.

4.2.3.2 F"/X,Y)3 shall be evaluated to determine if F as(X,Y) is within its limit by: '

a. Using the movable incore detectors to obtain a power distribution map F$fX,Y) 3

at any THERMAL POWER greater than 5% of RATED THERMAL POWER. ,

b. Satisfying the following relationship: ,

FAHR"(X,Y) s BHNOM(X,Y)

Where:

F[(X,Y)

FA HR "(X, Y) =

MAP " l AXIAL (X,Y)

And BHNOMIX,Y)*

represents the nominal design increased by an allowance for the expected deviation between the nominal design and the measurement.

MAPM is the maximum Allowable Peak"* obtained from the measured power distribution.

AXIAL (X,Y) is the axial shape for FanW,Y).

c. If the above relationship is not satisfied, then l 1. For the location, calculate the % margin to the maximum allowable design as l follows:

f T F[(X', Y)

%F ag Ma@n = 1- x%%

BHDES (X, Y) ,

where BHDES(X,Y)*

represents the maximum allowable design peaking factor which insures that the licensing criteria will be preserved for operation within the LCO limits, and includes allowances for calculational and measurement uncertainties.

1 i

No additional uncertainties are required in the following equations for F"y(X,Y),

3 because the limits include uncertainties.

- BHNOM(X,Y) and BHDES(X,Y) data bases are provided for input to the plant power distribution analysis computer codes on a cycle specific basis and are determined using the methodology for core limit generation described in the references in Specification 6.9.1.14.

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

2. Find the minimum margin of all locations examined in 4.2.3.2.c.1 above.

3. If any margin in 4.2.3.2.c.2 above is < 0, reduce the allowable THERMAL POWER from RATED THERMAL POWER by RRH* % x most negative margin from 4.2.3.2.c.2 and maintain the requirements of Specification 3.2.3; otherwise the Action statements for 3.2.3 apply.

d. With two measurements extrapolated to 31 EFPD beyond the most recent !

measurement yielding i

FAHR"(X,Y) > BHNOM(X,Y)

)

either of the following actions shall be taken: l

1. F"JX,Y) 3 shall be increased over that specified in 4.2.3.2.a by the )

appropriate factor specified in the COLR, and 4.2.3.2.c.1 repeated, or i

2. F"JX,Y)3 shall be evaluated according to 4.2.3.2 at or before the time when l the margin is projected to result in the action specified in 4.2.3.2.c.3. -

4.2.3.3 Fas(X,Y) shall be determined to be within its limit by using the incore detectors to obtain a power distribution map:

a. Prior to operation above 75% of RATED THERMAL POWER after each fuel j loading, and ,

i

b. At least once per 31 EFPD. i l

l i

RRH is the amount of power reduction required to compensate for each 1 % that Fas(X,Y) exceeds the limit provided in the COLR per Specification 6.9.1.14.

4

POWERDISTRIBbTIONLIMITS 3/4.2.4 QUADRANT POWER TILT RATIO

(' j LIMITING CONDITION FOR OPERATION 3.2.4 The QUADRANT POWER TILT RATIO shall not exceed' T%C Lt es s T st%coctto j APPLICABILITY: MODE 1 above 50% of RATED THERMAL POWER

s ~ re+ s Coc/(,

ACTION: l

a. With the QUADRANT POWER TILT RATIO determined to exceed but less than or equal to 1.09:

l

1. Calculate the QUARANT POWER TILT RATIO at least once per hour i O until either: ,

a) The QUADRANT POWER TILT RATIO is reduced to within its N limit, or H aI b) THERMAL POWER is reduced to less than 50% of RATED THERMAL l Y POWER.

2. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months either:

g a) Reduce the QUADRANT POWER TILT RATIO to within its limit, i

]Q> or

qs b) Reduce THERMAL POWER at least 3% from RATED THERMAL POWER i for each 1% o" indicated QUADRANT POWER TILT RATIO in j -

excess of],2 4 and similarly reduce the Power Range Neutron Flux-High Trip Satpoints within.the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Trip set-points to less than or equal to 55% of RATED THERMAL POWER l within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . I i

4. Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION above 505 of RATED THERMAL power may proceed provided that the QUADRANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

"See Special Test Exception 3.10.2.

R130 SEQUOYAH - UNIT 2 3/4 2-10 Amendment No. 130 00T 02 M0

I

~

4 POWER DISTRIBUTION LIMITS ACTION: (Continued)

b. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to misalignment of either a shutdown or control rod: -}

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: p. j a) The QUADRANT POWER TILT RATIO is reduced to within its limit, or i h; I ?b b) THERMAL POWER is reduced to less than 50% of RATED THERMAL b g POWER.

h

2. Reduce THERMAL POWER at least 3% from RATED THERMAL POWER for g s

each 1% of indicated QUADRANT POWER TILT RATIO in excess of

% within 30 minutes. o

3. Verify that the QUADRANT POWER TILT RATIO is within its limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after exceeding the limit or reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and reduce the Power Range Neutron Flux-High Set-points to less than or equal to 55% of RATED TH RMA POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

-l

4. Identify and correct the cause of the out o limit condition prior to increasing THERMAL POWER; subsequent POWER OPERATION a; 've 50% of RATED THERMAL POWER may proceed provided that the QUa0 RANT POWER TILT RATIO is verified within its limit at least once per hour for 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or until verified acceptable at 95%

or greater RATED THERMAL POWER.

c. With the QUADRANT POWER TILT RATIO determined to exceed 1.09 due to causes other than the misalignment of either a shutdown or control red:

1. Calculate the QUADRANT POWER TILT RATIO at least once per hour until either: ,

a) The QUADRANT POWER TILT RATIO is reduced to within its j 1

limit, or b) THERMAL POWER is reduced to less than 50% of RATED THERMAL POWER. .

l 1

1 R130 3/4 2-11 Amendment 130 SEQUOYAH - UNIT 2 00T 0 2 ,J,

m TABLE 3.2-1 E DNB PARAMETERS

,8 sz .

e E, LIMITS

-4 m

4 Loops In PARAMETER Operation Reactor Coolant System T,yg 5,583*F Pressurizer Pressure > 2220 psia

Reactor Coolant System Flow Rate %[8%g R130 w

A p qw aA- 3. 2 - 1 Z

N 9

38 or 5

w. .

q

co I _. c.

,co

Limit not applicable during either a THERMAL POWER ramp in excess of 5% of RATED THERMAL POWER Der minute or a THERMAL POWER step in excess of 10% of RATED THERMAL POWER, physics tes performance of surveillance requirement 4.1.1. 7%

' , 30 finc M s / 3 g f g m M r g nt/cer, tai'n% g p< e

1 k&O 77f15 l 5 c- ~

I I

l

~

Figure 3.2-1 Flow vs. Power for 4 Loops in Operation -

l l

i 362000 A 3.5% measurement uncertainty for flowis included in this figure. '

(100,360100) 360000 --

l 1

358000 --

Acceptable j Operation .

Region g 356000 --

c.

S!

2 se

~% 354000 --

il:

o IE E

$ 352000 --

N c

(

}350000" Unacceptable 3 Operation j Region E 348000 --

346000 --

344000 --

(90,342095) 342000 : : : . ;

s 90 92 94 96 98 100 102 ,

l Thermal Power Fraction (% of RTP)

?

f L

~

.ssa voya - var 2 % z-tr

TM c were.m. /t%o msat47%7709 4.s m rs M Skr oow~ too mska rwow L

rm es MG REACTIVITY CONThqL SYSTEMS Sp4c,g g,qo fg mc CMA Soc SWe mcex BASES k 9, l . /( f s 4 MOVnnf2 CONTROL ASSEMRLTES (Continued)

The ACTION statements which permit limited variations from the basic . .

rcquirements are accompanied by additional restrictions which ensure that the .

i criginal design criteria are met. Misalignment of a rod requires measurement '

of peaking factors and a restriction in THERMAL' POWER. These restrictions provide assurance of fuel rod integrity during continued operation. In ;

i cddition, those safety analyses affected by a misaligned rod are reevaluated to c:nfirm that the results remain valid during future operation.

In the event that a malfunction of the Rod Control System renders control rods R205 immovable, provision is made for continued operation provided: ;

o The affected control rods remain trippable, and o The individual control rod alignment limits are met. ,

In the event' that a malfunction of the Rod Control System renders control rod '

b:nks immovable during surveillance testing, provision is made for 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months of ;

continued operation provided: t i o The affected control rod banks remains trippable,

~

) o The individual control rod alignment limits are met, ,

o A maximum of one control or shutdown bank is inserted no more than !

'18 steps below the insertion limit,

~No reactor coolant system boron concentration dilution activities 4

!- o

[ or power level increases are allowed, and i i o The SHUTDOWN MARGIN requirements are verified every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or !

< upon insertion of controlling bank during the period the insertion ,

R limit is not met. ,

! The requirements to preclude Reactor Coolant System boron concentration f dilution, while a control or shutdown bank is below insert limits, will !

, minimize the impact on shutdown margin. .

The controlling bank (s)', which is normally Control Bank D, is excluded from the

72-hour provision since insertion of this bank (s) below the insertion limit is '

not required for control rod assembly surveillance testing. A controlling bank

' in defined as any control bank that is less than fully withdrawn as defined in

the COLR with the exception of fully withdrawn banks that have been inserted in cecordance with Surveillance Requirement 4.1.3.1.2. This provision excludes l the use of the 72-hour allowance for control banks that can be exercised 10 steps in either direction without exceeding the insertion limits.

,I

checks are performed for each reload core to ensure that bank insertions of up to 18 steps will not result in power distributions, which violate the DNB criterion for ANS Condition II transients (moderate frequency transients

} analyzed in Section 15.2 of the UFSAR). Administrative requirements on the i initial controlling bank position will ensure that this insertion and an j cdditional controlling bank insertion of five steps or less will not violate

the SHUTDOWN MARGIN requirement'of Specification 3.1.1.1 during the repair

. period. If the controlling bank is inserted more than five steps deeper than

! its initial position, a calculation will be performed to ensure that the

. SHUTDOWN MARGIN requirement of Specification 3.1.1.1 is met. Since no dilution or power level increases are allowed, shutdown margin will be maintained as lang as the controlling bank is far enough above its insertion limit to compensate for the inserted worth of the bank that is beyond its insertion limit.

The 72-hour period for a control rod assembly bank to be inserted below its insertion limit restricts the likelihood of a more severe (i.e., ANS Condition III or IV) accident or transient condition occurring concurrently with the j

i . insertion limit vi olat ion.

4 November 21, 1995 SEQUOYAH - UNIT 2 B 3/4 1-4 Amendment.No. 98, 205

i 3/4.2 POWER DISTRIBUTION LIMITS '

BASES 1

The specifications of this section provide assurance of fuel integrity during Condition I (Normal Operation) and II (Incidents of Moderate Frequency) i events by: (a) maintaining the calculated DNBR in the core at or above design during normal operation and in short term transients, anc (b) limiting the fission gas release, fuel pellet temperature and cladding mechanical i properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides. assurance that the Q initial conditions assumed for the LOCA cnclyses are met and the ECCS w jacceptance 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:

/ . Heat Flux Hot Channel Factor, is defined as the mastimum local

/

heat flux on the surface of a fuel roc at core elevation z divided by the average fuel rod heat flux, allowing for manufacturing tolerances on fuel pellets and rods.

l Nuclear Enthalpy Rise Hot Channel Factor, is defined as the ratio of i

the integral of linear power along the rod with the highest integrated power to the average rod power.

3 /4.2.,1 AXIAL FLUX DIFFERENCE (AFD)

'&t 4 The limits on AXIAL FLUX DIFFERENCE (AFD) assure that the bound envelope of the F g limit specified~in the COLR times the normalized

@ upper R146l axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution following power changes. !

Provisions for monitoring the AFD on an automatic basis are derived from R21 the plant process computer through the AFD Monitor Alarm. The compuer deter-mines the one minute average of each of the OPERABLE excore detector outputs '

and provides an alarm message immediately if the AFD for at least 2 of 4 or 2 of 3 OPERABLE excore channels are outside the allowed WI-Power operating space i and the THERMAL POWER is greater than 50~ percent of RATE ERMAL POWER.

l s E,,

3/4.2.2 and 3/4.2.3 HEAT FLUX AND NUCLEAR ENTHALPY HOT ANNEL FACTORS R130 The limits on heat flux hot channel factor and nuclear enthalpy hot chan-nel factor ensure that 1) the design limits on peak local power density and R21 minimum DNBR are not exceeded and 2) in the event of a LOCA the peak fuel clad temperature will not exceed the 2200*F ECCS acceptance criteria limit.

M PEh KlesC, Lini s AM SPEC.tesco ia ras CALR. /M. s/E c.t ee care 6.9././4. _

W SEQUOYAH - UNIT 2 B 3/4 2-1 Amendment No. 21, 130, 131, 146 March 30, 1992

POWER DISTRIBUTION LIMITS l gy 1 BASES }~

~

Each of these hot channel factors is measurable but will normally only be R130 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 + 13 steps from the group demand position.

R21

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

d* c. The control rod insertion limits of specifications 3.1.3.5 and 1 !

3.1.3.6 are maintained.

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

The limit as a function of THERMAL POWER allows chances in the ra la power shape for all permissible rod insertion limits, will be maintained ,

within its limits provided conditions a thru d above, are maintained. E s When Fqmea rement is en, both erimental e or and manuf dring erance st be allo d for. Th % is the a opriate all nce for a 11 core ma aken with e in core etector flu apping syst and 3% is ~ e appr riate allo nce for ma acturing to rance. /

g' hen F H

i easured, erimental e or must be owed for 4% is th approp ' ate allowa for a ful ore map tak with the -core det 1on /

s em. The s cified limi orFhals ontains a allowan for unc M

{

tainties w ch mean tha normal oper 1on will r it in F <F 1. 0 . N RIM The llowance i ased on th ollowing siderati .

a. abn al pertur ions in t radial p r shape uch as fr rod y l N

isalignmen , effect F more dir y than q.,

l g . alt gh rod mo ent has direct i - uence up limiting to /

g thin its ' it, suc control not read availabl o limit N

Fg,a ,

c. rrors i redicti for contye power s e detected ring st up j physi test ca e compensated for i by res ing axip flux -

q l tributi . This sation/rFfg is 1 readily 4vailap .

SEQUOYAH - UNIT 2 8 3/4 2-2 Admendment No. 21, 130, l 146 March 30, 1992

INSERT H When an Fo(X,Y,Z) measurement is taken, an allowance for measurement uncertainty is made. An allowance of 5% is appropriate for a full-core map taken with the incore Detector Flux Mapping System, and this allowance is included in the methodology applied to the determination of the core operating limits as described in the reference in Specification 6.9.1.14.

The hot channel factors, Fo(x,y,Z) and F"g(X,Y),

3 are measured periodically and .

compared to the nominal design values to provide a reasonable assurance that the core is operating as designed and that the limiting criteria will not be exceeded for !

operation within the Technical Specification limits of Sections 2.2 (Limiting Safety l System Settings),3.1.3 (Moveable Control Assemblies),3.2.1 (AXIAL FLUX i DIFFERENCE), and 3.2.4 (QUADRANT POWER TILT RATIO). An allowance is provided to account for the expected deviation between the calculation and the measurement. If the measurement is above the maximum expected value for that location, it is assumed to not be operating as designed, and a peaking margin evaluation is performed to provide a basis for decreasing the width of the AFD and f(AI) limits, and for. reducing THERMAL POWER.

i 1

I i

h i

l l

l POWER DISTRIBUTION LIMITS -

BASES Miiel rod ing redu he value NB ratio. rgin has be etained y betw the DNBR ue used i e safety ysis and th sign DNBR BR-e t to co ely offs e rod bo nalty.

d T pplicable ue of rod penalty is erenced in FSAR. R,1,46 Margi n excess o e rod bow ty is ava e for p design 1

, ,. ibility. R,1,30 e hot chan factor F M is meas periodic and incr d by a cycle a eight depe power f or W(z), t rovide ass ce that th i limi the hot c el factor q(z),is .

W(z) acc s for the cts L, f normal o ion tra nts and w etermined expected er control maneuv over the range o rnup condi s in the c . The W(z Rio6 f f ion is s fied in t LR.

T 3/4.2.4 QUADRANT POWER TILT RATIO quadrant assures that th

& u[bon sat Radi tiwer

r tilt ratio s the design va distribu used in the po al power dist -

apability anal

R21 easurements arg mad during start ng and - ,

riodically d power operatio e two hour t '

owance for o on with a tilt tion greater than 1.02 bu than 1.09 is ed to allow i ication and corre of a dropped saligned rod. e event such n does not margin for unce

rect the tilt y on F is r q ated by reducin the y 3 percent from THERMAL P0 r each percent of

~

in

$ ess of 1.0. / /

3/4.2.5 DNB PARAMETERS The limits on the DNB related parameters assure that each of the para-meters 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 I adequate to maintain a minimum DNBR of greater than or equal to the safety R130 analysis DNBR limit throughout each analyzed transient.

The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their R21 limits following load changes and other expected transient operation.

} & 5418r- y SEQUOYAH - UNIT 2 B 3/4 2-4 Admendment No. 21, 130, 146 By letter dated December 8, 1992

l INSERTI The QUADRANT POWER TILT RATIO limit assures that no anomaly exists such that i the radial power distribution satisfies the design values used in the power capability j analysis. Radial power distribt. tion measurements are made during startup testing and l periodically during power operation. ;

1 The OUADRANT POWER TILT RATIO limit at which corrective action is required provides DNB and linear heat generation protection with x-y plane power tilts. The QUADRANT POWER TILT RATIO limit specified in the CORE OPERATING LIMITS REPORT (COLR) is reflected by a corresponding peaking augmentation factor which is included in the generation of the AFD limits.

l The 2-hour time allowanco for operation with the tilt condition greater than the limit specified in the COLR, but less than 1.09, is provided to allow identification and correction of a dropped or misaligned control rod. In the event such action does not correct the tilt, the margin for uncertainty on Fo(X,Y,Z) is reinstated by reducing the allowable THERMAL POWER by 3 percent for each percent of tilt in excess of the limit i specified in the COLR. j INSERT J The flow parameters indicated in Figure 3.2-1 have been rounded down to bias the analysis in the conservative direction. l I

i 1

l I

I

\

i i

i INSTRUMENTATION BASES

{

! REACTOR TRIP SYSTEM AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM i INSTRUMENTATION (Continued) 1 The measurement of response time at the specified frequencies provides

assurance that the protective and the engineered safety feature actuation 4 associated with each channel is completed within the time limit assumed in the

! accident analyses. No credit was taken in the analyses for those channels

with response times indicated as not applicable in the updated final safety i analysis report.

! R182 Response time may be demonstrated by any series of sequential, overlapping l or total channel test measurements provided that such tests demonstrate the

! total channel response time as defined. Sensor response time verification may

! be demonstrated by either 1) in place, onsite or offsite test measurements or l 2) utilizing replacement sensors with certified response times.

Action 15 of Table 3.3-1, Reactor Trip System Instrumentation, allows 1

the breaker to be bypassed for up to 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months for the purpose of performing R46

maintenance. The 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months is based on a Westinghouse analysis performed in j WCAP-10271, Supplement 1, which determines bypass breaker availability.

1 3/4.3.3 MONITORING INSTRUMENTATION I i

I i

j 3/4.3.3.1 RADIATION MONITORING INSTRUMENTATION The OPERABILITY of the radiation monitoring channels ensures that 1) the radiation levels are continually measured in the areas served by the individual

('

channels and 2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

i 3/4.3.3.2 MOVABLE INCORE DETECTORS j The OPERABILITY of the movable incore detectors with the specified minimum

] complement of equipment ensures that the measurements obtained from use of l this system accurately represent the spatial neutron flux distribution of the

! reactor core. The OPERABILITY of this system is demonstrated by irradiating l each detector used and determining the acceptability of its voltage curve.

For the purpose of measurdgM88 Hll incore flux map is used.

) Quarter-core flux maps, as defined in wcAP-sseu, June 1976, may be used in d

recalibration of the excore neutron flux detection system, and full incore

l flux maps or symmetric incore thimbles may be used for monitoring the QUADRANT e POWER TILT RATIO when one Power Range Channel is inoperable. l 3/4.3.3.3 SEISMIC INSTRUMENTATION The OPERABILITY of the seismic instrumentation ensures that sufficient capability is available to promptly determine the magnitude of a seismic event and evaluate the response of those features important to safety. This capability is required to permit. comparison of t measured response to that used in the ;

i R.

h (X,Y 04. IM(,X,y) _

i SEQUOYAH - UNIT 2 B 3/4 3-2 Amendment No. 46, 72, 182 November.9, 1994

~

ADMINISTRATIVE CONTROLS MONTHL7 REACTOR OPERATING REPORT 6.9.1.10 Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the PORVs or Safety Valves, shall R64 y be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report.

pk R146 j CORE OPERATING LIMITS REPORT J 6.9.1.14 Core operating limits shall be established and documented in the CORE N OPERATING LIMITS REPORT before each reload cycle or any remaining part of a i reload cycle for the following:

7.* f. Moderator Temperature Coefficient BOL and EOL limits and 300 ppm surveillance limit for Specification 3/4.1.1.3, 3, g. Shutdown Bank Insertion Limit for Specification 3/4.1.3.5, l f , .8. Control Bank Insertion Limits for Specification 3/4.1.3.6, m,s a A u- t' .* P o 1-M- s xmsn.%

g, g. Limits for Specification 3/4.2.1, )

f. g. . Heat Flux Hot Channel Factory K(2)l,jQ (z , d e act/r tpf:it R206 s e pdteJrtiaVdecyta% 1r .a , i et%en / /

u il Jfor S cation 3/4.2.2, a

. R146

7. g. Nuclear Enthalpy Hot Chann 1 Factor [ptidgowpf F9(top 41uly/fp1;idrl f or Specification 3/4.2.3 , pp.a o 6.9.1.14.a The analytical methods used to determine the core operating limits g shall be those previously reviewed and approved by NRC in:

1. W -927 N

-P-A, / ESTIN OUSE LOAD S ETY EV ATION THODOLO ,

j

'\ ly 1 5 (W Propriet Metho - ogy f y).

Spec' icatio 3.1.1.3 Moder or Temp ature Coeff' ient, .1.3. - Shut wn Bank nsertio Limit, .1.3.6 -

Con el Ba Inser ion Lim

s, 3.2. - Axia Flux Di erence, b 3. .2 - H t F1 Hot Ch el Fac r, and .2.3 1 clear

/

g thalpy et C nnel Fa or.)

_ vision "RE TION O CONST AXIAL 0 SET R206 q ~'. W -1021 P-A, ,

y NTROL g SUR LLANCE ECHNI SPECIFI TION", EBRUARY 994

'd 3hS f (W Pro .ieta .

Methe logy fo Specifi tion 3. .1 - Ax 1 Flux fferenpe 7146 0 (Rel ed Axia Offset ntrol) d 3.2. - Heat F ux Hot /hannel Facy r (w(z) surveill ice reqb rements, or Fo M hodolop9).)

(}

g [. WCAP-10266-P-A, Rev. 2, THE 1981 REVISION OF WESTINGHOUSE EVALUATION MODEL USING BASH CODE", March 1987, (W Proprietary).

fqg (Methodology for Specification 3.2.2 - Heat Flux Hot Channel q Factor). _.

T' j 4. WCAP-1 316-P-A, SAFETY ALUATI J SUPPO NG A NEG VE EOL 161

~5 MODE OR TEMP TURE C FFICIEl TECHNI SPEC ICATIO- FOR THE

@O k SEQ YAH NUCL PLANT " Marc 1993,

_ Propr tary).

(Metho logy fo Specif ation 3 .1.3 oderat Temper ure d Coeff'cient). /

N December 11, 1995 6-22 Amendment No. 44, 50, 64, 66, 107, SEQUOYAH - UNIT 2 134, 142, 146, 161, 206 !

INSERT K

1. f i (All limits for Overtemperature Delta T Trip Setpoints and f2 (All limits for Overpower Delta T Trip Setpoints for Specification 2.2.1.

INSERT L

1. BAW-10180P-A, Rev.1, "NEMO - NODAL EXPANSION METHOD OPTIMlZED",

March 1993. (FCF Proprietary)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

2. BAW-10169P-A,"RSG PLANT SAFETY ANALYSIS - B&W SAFETY ANALYSIS METHODOLOGY FOR RECIRCULATING STEAM GENERATOR PLANTS",

October 1989. (FCF Proprietary)

(Methodology for Specification 3.1.1.3 - Moderator Temperature Coefficient.)

3. BAW-10163P-A, Core Operating Limit Methodology for Westinghouse-Designed PWRs, June 1989. (FCF Proprietary)

(Methodology for Specification 2.2.1,- Limiting Safety System Settings

[f i (AI), f 2(Al) limits),3/4.1.3.5 - Shutdown Bank Insertion Limits,3.1.3.6 -

Control Bank Insertion Limits, 3/4.2.1 - Axial Flux Difference , 3/4.2.2 - Heat Flux Hot Channel Factor,3/4.2.3 - Nuclear Enthalpy Rise Hot Channel Factor, 3/4.2.4 - Quadrant Power Tilt Ratio.)

4. BAW-10168P-A, Rev.2, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (SER expected March 1996).

(FCF Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

5. BAW-10168P-A, Rev. 3, RSG LOCA - B&W Loss of Coolant Accident Evaluation Model for Recirculating Steam Generator Plants, (SER expected June 15,1996).

(FCF Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

6. WCAP-10054-P-A, Westinghouse Small Break ECCS Evaluation Model Using the NOTRUMP Code, August 1985. (W Proprietary)

(Methodology for Specification 3.2.2 - Heat Flux Hot Channel Factor.)

._ . . _ . . . . .,_s.. . .m.; . _ . .--

4 ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE

SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 i

DOCKET NOS. 50-327 AND 50-328 ,

(TVA-SON-TS-96-01) i DESCRIPTION AND JUSTlFICATION FOR THE CONVERSION FROM WESTINGHOUSE ELECTRIC CORPORATION FUEL TO FRAMATOME COGEMA FUEL (FCF) i l

l l (MARK-BW17) ,

l l l

1 i

l l

I I 4

Descriotion of Chance TVA proposes to modify the Sequoyah Nuclear Plant (SQN) Units 1 and 2 technical specifications (TSs) to allow the use of nuclear fuel provided by Framatome Cogema Fuels (FCF). The appropriate sections of the Safety Limits and Safety Limits Bases will be revised to include new terms that are to be added to the Core Operating Limits Report (COLR). As part of the Safety Limits changes, the reactor coolant system (RCS) design flow rate listed in Table 2.2-1 will be reduced to 87,000 gallons per minute per loop (without uncertainty). TSs 3.2.2, Heat Flux Hot Channel Factor -

Fo(X,Y,Z), 3.2.3, Nuclear Enthalpy Hot Channel Factor - Fas(X,Y), and 3.2.4, Quadrant Power Tilt Ratio, will be revised to include FCF terminology and methodology.

Surveillance Requirements (SRs) 4.2.2.2, .3, 4.2.3.1, .2, .3 will be revised to agree with the associated TSs. SR 4.2.5.1 will have Figure 3.2-1 added to display appropriate RCS flow rates. Additionally, the corresponding Bases sections will be revised to reflect this new terminology and methodology. TS 6.9.1.14 will also be revised to indicate the FCF methodology used to support the fuel conversion.

Reason for Chanae -

SQN's present fuel contract will expire at the end of each unit's Cycle 8 refueling outage. Based on the cost savings provided by FCF, TVA will change fuel vendors.

This TS change is required to implement the FCF methodology associated with the new fuel assembly design.

Justification for Chanaes FCF has provided Topical Report, BAW-10220P, Revision 0, to support the conversion to FCF fuel. This report is contained in Enclosure 5. This report also provides justification to reduce the RCS design flow rate for SON when fueled with Mark-BW fuel. Westinghouse Electric Corporation has additionally evaluated the impact of the reduced RCS flow rate on the resident Westinghouse fuel assemblies and determined that the reduced flow rate was acceptable for previously burned Westinghouse fuel assemblies. This information is contained in Enclosure 9.

The following provides a summary of Topical Report BAW-10220P.

The Mark-BW 17x17 fuel assemblies will include the following features to ensure compatibility with the presently installed Westinghouse fuel assemblies:

Leaf-type holddown springs

Dashpot region in the guide thimbles for control rod deceleration

Mixing vanes on selected intermediate spacer grids

Floating spacer grid restraint system l This design was previously approved at Duke Power Company's Catawba and l McGuire units and Portland General Electric's Trojan plant, in these applications, the i Mark-BW assembly was shown to be compatible with the Westinghouse standard fuel assembly designs.

For application to SON, tests, measurements, and evaluations were conducted to establish that the Mark-BW fuel assembly is structurally, hydraulically, and neutronically compatible with the Westinghouse-supplied VANTAGE SH fuel assembly currently in use at SQN. Using two VANTAGE SH fuel assemblies supplied by TVA, data were collected from the interface dimensions and key performance features.

Damping and harmonic tests were performed on these assemblies to verify similarity with the Mark-BW and to obtain results for future use in mixed core mechanical analyses.

The hydraulic tests were performed in FCF's cold-water loops, which has been shown to be acceptable through previous hot-to-cold loop results comparisons. These flow tests provide a distribution of pressure drops through the length of the assembly.

Resultant assembly and component pressure drop values were used to determine form loss coefficients for the VANTAGE 5H assembly and were applied in the thermal-hydraulic evaluation of a mixed core of VANTAGE SH and Mark-BW assemblies.

From a neutronic standpoint, the VANTAGE SH and the Mark-BW fuel assemblies are almost identical. The structural materials within the active fuel region are similar in composition and weight. The slight differences in uranium loading will be modeled such that isotopic composition and burnup differences are properly calculated. Thus, the use of the Mark-BW assembly, in conjunction with the Westinghouse VANTAGE SH assembly in the core, does not adversely affect plant operation or neutronic parameters, in accordance with requirements of 10 CFR 50.46, an evaluation of the emergency core cooling system (ECCS) performance was performed for FCF Mark-BW reload fuel at SON utilizing the guidelines of 10 CFR 50, Appendix K. At the time of the initial operation, SON was fueled by Westinghouse with standard fuel. Later, the Westinghouse VANTAGE 5H design was implemented at SON. Compliance was demonstrated by Westinghouse for both fuel types. FCF calculations and evaluations documented in Enclosure 5 demonstrate that SON continues to meet these critaria when operated with Mark-BW fuel assemblies. Large break loss-of-coolant-accident (LOCA) calculations performed in concurrence with an approved evaluation model demonstrate compliance up to and including the double-ended severance of the largest primary coolant pipe. The small break LOCA calculations also demonstrate that the plant meets 10 CFR 50.46 criteria for small breaks when loaded with Mark-BW fuel.

The coexistence of Mark-BW fuel with resident Westinghouse fuelis shown to be inconsequential and does not cause the calculated temperatures for the different assembly types to approach the limits of 10 CFR 50.46.

i 1

All Final Safety Analysis Report (FSAR) Chapter 15 non-LOCA transient events were evaluated using approved methodology. For those transients potentially affected by operation of Mark-BW fuel, the bounding cases were reanalyzed. For other transients, the relevant core-related parameters, pertinent to future reloads, were evaluated for their effect on transient events. The events specifically analyzed were the following:

Control rod bank withdrawal at power l-

Loss of forced reactor coolant flow I

Locked reactor coolant pump rotor

Loss of electricalload

Main steam line break

Steam line break with coincident rod withdrawal The non-LOCA analyses and evaluations confirmed that the operation of the SON l units for reload cycles with Mark-BW fuel continues to be bounded by the previously l reviewed and licensed safety limits. Reanalyses of the transients affected by the fuel l reloads demonstrate that the acceptance and design criteria specified in Regulatory l Guide 1.70 continue to be met.

Thermal-hydraulic analyses were performed to support the Mark-BW fuel application -

and future Mark-BW reload fuel use at SON. The result of the thermal-hydraulic l analyses is a set of operating limits that insure fuel and clad integrity are maintained during normal operation and transients of moderate frequency. The design criteria i that were established and met to achieve this goal were as follows: )

1. During Condition I and 11 events, there must be at least a 95 percent probability i I

with a 95 percent confidence level that the hot pin will not experience a Departure From Nucleate Boiling (DNB); or a 99.9 percent probability that DNB will not occur

! core wide.

2. During Condition I and ll events, there must be at least a 95 percent probability l with a 95 percent confidence level that no fuel will experience centerline melting.

l The structural design requirements for the Mark-BW fuel assembly were derived in large part from FCF experience, both in design and in-core operation of similar designs. For application to SON, plant specific design requirements and parameters augmented the established Mark-BW design criteria. These requirements in total are consistent with the acceptance criteria of the Standard Review Plan (NUREG-0800),

Section 4.2, and follow the guidelines established by Section 111 of the American Society of Mechanical Engineers (ASME) code. Code Level A criteria are used for normal operation and Code Level D criteria are used for LOCA/ seismic operation.

l l

l t

4 i

l l

The design basis analyses performed to verify the adequacy of the Mark-BW fuel assembly in SON were as follows:

Normal operations Growth allowance

- Fuel assembly holddown

- Guide thimble buckling

- Spacer grid loads

- Interface with adjacent assembly

- Lateral seismic and LOCA loading

- Fuel assembly vertical LOCA loading

- Fuel assembly component stress

- Shipping and handling loads The above analyses and the review of the content of the NRC-approved Topical Report BAW-10172P," Mark-BW Mechanical Design Report," confirm that the Mark-BW fuel assembly maintains its mechanicalintegrity when operated in SON either as a full complement of Mark-BW assemblies or in conjunction with the resident fuel assemblies.

An evaluation of the long-term containment integrity as detailed in Chapter 6 of SON's FSAR shows that when the SON core contains Mark-BW fuel the existing analysis )

results remain bounding. The important aspects of the fuel change that potentially '

l impact the analysis are changes in the flow characteristics past the fuel, the RCS average operating temperature, the core stored energy and fuel heat capacity, and the decay heat. Each aspect was evaluated considering detailed assembly testing and measurement results and/or comparisons with existing analyses. Based on these '

evaluations, there is no consequence to the containment systems when the SON units are fueled with FCF-supplied Mark-BW assemblies.

All analyses and evaluations presented in Enclosure 5 confirm and justify the operation of TVA's SON units with Mark-BW fuel reloads in combination with resident fuel design or a complete core of Mark-BW assemblies.

The specific changes to the TS are based upon those approved for a similar transition from Westinghouse Electric Corporation to Babcock & Wilcox (B&W) fuel on September 30,1991, for the Trojan Nuclear Plant. B&W Topical Report BAW 10163P-A also provides the basis for many of the proposed TS changes. The following significant exceptions are noted:

1. Specific values for the OTNL, QTPL, OTNS, OTPS, QPNL, OPPL, QPNS, and OPPS have been relocated to the COLR consistent with Generic Letter 88-16.

2. The SON TS change request includes separate peaking limits for the resident Westinghouse fuel and the new FCF fuel because the FCF analysis has been performed to a higher peaking limit.

l i

3. Action 1.a to Limiting Condition of Operation 3.2.2 has increased from 15 minutes to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months to allow for the performance of actions in a controlled manner. These actions are administrative, which involve other organizations besides Operations. It should be noted that this time limit is consistent with the two hour time limit provided in SR 4.2.2.2.c.3.a and .b of the Trojan amendment discussed above.

Other differences to the Trojan amendment are considered enhancements for clarity and to facilitate the implementation of the TSs or are a result of the different initial-TSs starting point.

Environmental Imoset Evaluation The proposed change does not involve an unreviewed environmental question because operation of SON Units 1 and 2 in accordance with this change would not:

1. Result in a significant increase in any adverse environmentalimpact previously evaluated in the Final Environmental Statement (FES) as modified by NRC's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmentalimpact appraisals, or decisions of the Atomic Safety and Licensing Board.

2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SON that may ha'n' a significant environmentalimpact.

l i

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ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-96-01)

DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATION FOR THE CONVERSION FROM WESTINGHOUSE ELECTRIC CORPORATION FUEL TO FRAMATOME COGEMA FUEL (FCF) (MARK-BW17)

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l Significant Hazards Evaluation l

TVA has evaluated the proposed technical specification (TS) change and has I determined that it does not represent a significant hazards consideration based on criteria established in 10 CFR 50.92(c). Operation of Sequoyah Nuclear Plant (SON) in accordance with the proposed amendment will not: l

1. Involve a significant increase in the probability or consequences of an accident previously evaluated. :

The analyses provided in Topical Report BAW-10220P show that the changes do not significantly change the results of previously evaluated events. These '

analyses provide the template for accident analyses assumptions that must be met by the cycle-specific reload analysis.

The SON Units 1 and 2 Cycle 9 reload cores with Mark-BW fuel will be designed to operate within the approved limits for accident analysis. The limits provided in the TS and described in the Updated Final Safety Analysis Report (UFSAR) j provide the framework for accident analyses. By maintaining these limits, the probability or consequences of accidents related to the core changes do not i significantly change. Thus, it is concluded that there is no significant increase in the probability or consequences of an accident previously evaluated.

2. Create the possibility of a new or different kind of accident from any previously analyzed.

The change to Mark-BW fuel cores and mixed (transition) cores has been ;

evaluated in the Topical Report BAW-10220P. It was concluded that the change i did not create new or different kinds of accidents. The change in fuel suppliers has been evaluated for consideration of the effects of power distribution and peaking factors such that there are no restrictions on the use of Mark-BW fuel assemblies beyond those already established in the UFSAR and TS. Adherence to the safety analysis limits restricts the possibility of new or different accidents.

Historically, new accidents have not been associa2d with changes in fuel suppliers as long as safety analysis limits continue to be met. It is concluded that transition to Mark-BW fuel does not create the possibility of a new or different kind of accident from any previously analyzed. 1 i

3. Involve a significant reduction in a margin of safety.

The margin of safety is established by the acceptance criteria used by NRC.

Meeting the acceptance criteria assures that the consequences of accidents are within known and acceptable limits. The loss-of-coolant accident (LOCA) acceptance criteria are unchanged: peak cladding temperature of s2200 degrees Fahrenheit, peak cladding oxidation of s 17 percent, average clad oxidation of s 1 percent, and long-term coolability. These requirements continue to be met.

The methods used to demonstrate conformance with these limits have changed, and were reviewed to assure that the methods, as well as the results, are I

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2-acceptable. The acceptance criteria for Departure from Nucleate Boiling (DNB) events has not changed and is still the 95 percent probability and 95 percent confidence interval that DNB is not occurring during the transient. The DNB correlation, and methods used to demonstrate that DNB limits are met have changed, and these changes were reviewed to assure conformance wi:h acceptable practices. Other changes, as well as the changes discusr.d above, have been evaluated in the referenced safety analyses and are shown to meet applicable acceptance criteria. Other margins, such as avoiding fuel centerline l melting, are not significantly changed. Based on these results, it in concluded that the margin of safety is not significantly reduced.

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4 ENCLOSURE 4 PROPOSED TECHNICAL SPECIFICATION (TS) CHANGE

SEQUOYAH NUCLEAR PLANT (SON) UNITS 1 AND 2

! DOCKET NOS. 50-327 AND 50-328 (TVA-SON-TS-96-01) j APPLICATION FOR WITHHOLDING AND AFFIDAVIT

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ML20101N707

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