ML20076J919

Proposed Tech Specs Re Cycle 2 Reload Operations

ML20076J919
Person / Time
Site:	Sequoyah
Issue date:	07/01/1983
From:	TENNESSEE VALLEY AUTHORITY
To:
Shared Package
ML20076J913	List: ML20076J911 ML20076J916 ML20076J919
References
NUDOCS 8307070011
Download: ML20076J919 (44)

Text

F.

ee e

e APPENDIX A TECHNICAL SPECIFICATION CHANGES 23 8307070011 830701 PDR ADOCK 05000328 P

PDR

.a#

w_au

_m

___A.

m

.w a.a a

(

1, s.

8.

.4 e

D 9

= = _ _

.a...,+_.

..-.._.q f'

..p.6M.e..

94...Nd..--

O

.. -.6

. _.............. -.g......--_

. =.

p.,.._e..

$4_..

.p _W... - _.. _ _.

a.,

og_.

y

.._@.+.._..-

ei

. _ _..h.4_h.

qh.pM M

.e-

p.g ggm g g9 e.

.._.s L.,...

c-..u.

.....s

- N.$

_ * * + * *. - ',.

_4.

...._.*a.

^F.._*

..^

~. _ _

e-..&

g*--,...d.'.6..

-.7 M-h.N

._..*Mesg..e.+._

g....._*+.e

.-.M.._.e.ww W.,_

.e+..+.=.

+e.eh-.m.._.... + _..

- ee e g -._

_y,h 4

-.a.*

670~- _. _

_g.eM.4

+:

e.

O

~

_.e_.

..

• _.. _ -. _.. _..64+_.

e.e y e__e

._.._e._.__.

r _. _ _._ _.....x. - -

i l

L-.=. nm 24002SB._.:

WEccarac-f 6_._

N~_ _.

t.._

uP._cRA uO1_

i...___...._.

o o_.

..-.r _._ _..: _

_.. _. _. _ _. _.. _... -. _.__...__.= o 1

x#.

MO - -

x

-...M j N.

==.- '630_-

- - - - -.. -. _ - + _ _

' ~...

m

~

e..-

.=___

+

g 1

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

-t

_. o. _.. _.

-x

%m

_ _.. _. _. ~

~._

._..___c __._

_. =.

.x N _. -.pS

.o.

y

__m..._

s.

N.

m._s.

.3 _.____..

)

m o._._.._..

. _ x-1_v.

C" _ _ _.

'N.__.

.x x

x___.

1 x__._.s __s N

590.

__'x - ~ _.

x s

._x.

.ACCE.P.TABL:.-

.@~

OpERMION-

=

1.__

+

m.x p_

m. _x -

g.

v_ _

_... _ _ _x.

+_.

(

2...-

i_____....

x

_......u.._.._._

._._2__._

1

__._.a

"=,------

.._._._.50._5._"..__~."__-.."...,~~--r~.."_"~-~---"+""".

e

=

__A_.- _ _am

_ _.8

-..C,

-1.s

-....0.__-

+

c

_.. _ _. _._..- _.._...__.4

...._.4._.i....__.

-I

.x

. ~._ _..:

-_.-.......-..~..a_.__......

__.=t.---

r

._._3_._

3._a._._

~

f_R A...CIlO...N. 4'.O. F R.._A.T.E_D,.T.H..E. R._.M, _A.t_.".PO. _W EE

_r

.. t _ _ t._....a. Ein.-.ure:E.1.._.

_t:: R e..u..._...... S efo. ty..L%......,_. d_ to;.ro rn-O;.a. _ ts... y_....

.. _ ~ ~. _.

actor. Core:

srt =rou sera oa._

___....._1_..

.t._........._.

z.__

... q

(

=ra..x... __4 ~.-

..._..__.c

_1 t

.._,__4._._..._-.

m.~

,.... a w..._... _

2-2

_:=: SEQUOYAH UNIT 2

- H:n:

-+. - - - - ~ ~ - -

n. i s.,.

s

.u.

?/

.~...

%....myer =.w.w.n ;n r.=.wg s

.. ~..

~

TABLE 2.2-1 (Continued) h REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5!

E FUNCTIONAL UNIT TRIP SETPOINT All0WABLE VALUES C5 21.

Turbine Impulse Chamber Pressure -

< 10% Turbine Impulse

< 11% Turbine Impulse (P-13) Input to Low Power Reactor Trips Pressure Equivalent Pressure Equivalent w

Block P-7 22.

Power Range Neutron Flux - (P-8) Low 5 35% of RATED

$ 36% of RATED Reactor Coolant Loop Flow, and Reactor THERMAL POWER THERMAL POWER Trip 23.

Power Range Neutron Flux - (P-10) -

> 10% of RATED

> 9% of RATED Enable Block of Source, Intermediate, THERMAL POWER THERMAL POWER

.and Power Range (low setpoint) Reactor Trips 24.

Reactor Trip P-4 Not Applicable Not Applicable 25.

Power Range Neutron Flux - (P-9) -

< 50% of RATED

< 51% of RATED Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER Trip Below 50% Rated Power NOTATION I

) i AT, {K) - Kg (1 + T b)[T(

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

I 2

NOTE 1:

Overtemperature AT (

3 j

,e+tS 1+T b I * '4b j

3 jf

= Lag compensator on measured AT where:

= Time constants utilized in the lag compensator for AT '1 = 2 secs.

t 3

j AT

= Indicated AT at RATED THERMAL POWER g

K

$ 1.15 j

K

= 0.011 2

l

TABLE 2.2-1 (Continued) sj REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5

NGTATION (Continued)

I g

NOTE 1:

(Continued)

~

-e 1+1S 2

The function generated by the lead-lag controller for T dynamic compensation

=

j-

• T 3 avg 3

= 33 secs.,

2' & '3 Time constants utilized in the lead-lag controller for T,yg, tg

=

1 1 = 4 secs.

3 Average temperature *F T

=

Lag compensator on measured T

=

j, 3

avg 4

Time constant utilized in the measured T,yg lag compensator, 1 = 2 secs.

c I

=

4 578.2*F (Nominal T,yg at RATED THERMAL POWER) 9 T'

0.00055 K

=

3 Pressurizer pressure, psig P

=

2235 psig (Nominal RCS operating pressure)

P'

=

Laplace transform operator (sec-I)

S

=

and f (AI) is a function of the indicated difference between top and bottom detectors j

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

arephrcenkRATEDTHERMALPOWERinthetopandbotlo(AI)=0(whereqmhalvesoftheborereNpe for q q between - 29 percent and + 5 percent f and q (i) is total THERMAL POWER in percent of RATED THERMAL POWER).

and qt*9b

TABLE 2.2-1 (Continued)

~

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS S!

E NOTATION (Continued)

E NOTE 1:

(Continued) w q

exceeds -29 percent, the AT trip set-(ii) for each percent that the magnitude of (qpointshallbeautomaticallyreducedbyIb50pN)rcentofitsval q

exceeds +5 percent, the AT trip set-(iii) for each percent that the magnitude of (qpointshallbeautomaticallyreducedbyOb86pN)centofits r

I S

I S

)(

) T -K D(

) - T9 - f (OI)}

5 (1 + t b NOTE 2:

Overpower AT (

) <_ AT,{K4 -K 6

2 I*Ib I*Ib 1+tS

}

j 5

4 4

I as defined in Note 1 Where:

=

1+rSj as defined in Note 1 T

=

j as defined in Note 1 AT,

=

K F 1.087 4

K

= 0.02/*F for increasing average temperature and 0 for decreasing average 5

temperature IS S

The function generated by the rate-lag controller for T dynamic

=

1+rS avg 5

compensation

_ = -

TABLE 2.2-1 (Continued)

Eg REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS Y

z NOTATION (Continued)

E Q

NOTE 2:

(Continued)

Time constant utilized in the rate-lag controller for T,yg, 15 = 10 secs.

1

=

5 as defined in Note 1

=

1+T 5 4

as defined in Note 1 I

=

4 K

=

0.0011 for T > T" and K6 = 0 for T 5 T" 6

m Q

T as defined in Note 1

=

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

=

g AT instrumentation, 5 578.2 F)

S as defined in Note 1

=

f (AI) 0 for all AI

=

2 NOTE 3: The channel's maximum trip setpoint shall not exceed its computed trip point by more than 2 percent.

l

--=_- -.

o, 0

4 t

2.1 SAFETY LIMITS BASES 2.1.1 REACTOR CORE The restrictions of this Safety Limit prevent overheating of the fuel and

}

possible cladding perforation which would result in the release of fission Overheating of the fuel cladding is prevented products to the reactor coolant.

by restricting fuel operation to within the nucleate boiling regime where the heat transfer coefficient is large and the cladding surface temperature is l

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 j~

from nucleate boiling (DNB) and the resultant sharp reduction in heat transfer DNB is not a directly measurable parameter during operation and coefficient.

tnerefore THERMAL POWER and Reactor Coolant Temperature and Pressure have been related to DNB tnrough the W-3 correlation. The W-3 DNB correlation has been developed to predict the DNB flux and the location of DNB for axially uniform and non uniform heat flux distributions.

The local DNB heat flux ratio, DNBR, defined is the ratio of the heat flux that would cause DNB at a particular 2

core location to the local heat flux, is indicative of the margin to DNB.

The minimum value of the DNBR during steady state operation, normal This operational transients, and anticipated transients is limited to 1.30.

i value corresponds to a 95 percent probability at a 95 percent confidence level that OhB will not occur and is chosen as an appropriate margin to DNB for all operating conditions.

The curves of Figures 2.1-1 and 2.1-2 show the loci of points of THERMAL POWER, Reactor Coolant System pressure and average temperature for which the minimum CNBR is no less than 1.30, or the average enthalpy at the vessel exit is equal to the enthalpy of saturated liquid.

These curves are based on an enthalpy hot channel factor, F g, of 1.55 i

and a reference cosine with a peak of 1.55 for axial power shape. An allowance is included for an increase in Fh at reduced power based on the expression:

F5H = 1.55 W 0.3 ( W where P is the fraction of RATED THERMAL POWER These limiting heat flux conditions are higher than those calculated for the range of all control rods fully withder.vn to the maximum allowable control rod insertion assuming the axial power imbalance is within the limits of tne f; (delta I) function of the Overtemperature trip. When the axial power i

i imbalance is not within the tolerance, the axial power imbalance effect on the Overtemperature delta T trips will reduce the setpoints to provide protection consistent with core safety limits.

SEQUOYAH - UNIT 2 B 2-1 I

l

i 3/4.2 POWER DISTRIBUTION LIMITS j

3/4.2.1 AXIAL FLUX DIFFERENCE (AFD)

LIMITING CONDITION FOR OPERATION 3.2.1 The indicated AXIAL FLUX DIFFERENCE (AFD) shall be maintained w'ithin the allowed operational space defined by Figure 3.2-1.

APPLICABILITY: MODE 1 above 50% RATED THERMAL POWER

• ACTION:

With the indicated AXIAL FLUX DIFFERENCE outside of the Figure 3.2-1 a.

limits; 1.

Either restore the indicated AFD to within the Figure 3.2-1 limits within 15 minutes, or 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER within 30 minutes and reduce the Power Range Neutron Flux-High Trip setpoints to less than or equal to 55 percent of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

b.

THERMAL POWER shall not be increased above 50% of RATED THERMAL POWER unless the indicated AFD is within the Figure 3.2-1 limits.

SEQUOYAH - UNIT 2 3/4 2-1

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 4.2.1.1 The indicated AXIAL FLUX DIFFERENCE shall be determined to be within its limits during POWER OPERATION above 50% of RATED THERMAL POWER by:

Monitoring the indicated AFD for each OPERABLE excore channel:

a.

1.

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

At least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after restoring the AFD Monitor Alarm to OPERABLE status, b.

Monitoring and logging the indicated AXIAL FLUX DIFFERENCE for each OPERABLE excore channel at least once per hour for the first 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and at least once per 30 minutes thereafter, when the AXIAL FLUX DIFFERENCE Monitor Alarm is inoperable.

The logged values of the indicated AXIAL FLUX DIFFERENCE shall be assumed to exist during the interval preceding each logging.

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

SEQUOYAH - UNIT 2 3/4 2-2

~.

- -:1-I.:_di.::::. '__ ;.iFi:G:idE:!:" # - l

. 23:f.ii. +._._.:f. ; ; !..l..

a_=._-_-_.=.---- - - - - - -

.1. :

w

c

-. :-.... -._ -.=- -.:, =.=: -. w., u ;-

=.

... -.... e o--_

-- --.-.==: ;

a a:

,g-w.

. c: _.

w.

_.._ _ _ c :..-

w :I: ;= = ( S.100 )._

- (- A.e. l,s,0 ) _

=

.c;:-

100 1

i j Eic;;;7satg'7 I

\\E UNACCE m 3LE CPE:ATION y/_

l g

CPE?ATION 1

-j 80 t

~ '.

i 3

s i

i i:

nC.r.e? a n. L.e. =.

o 3

F= OPE?ATICN '-

g

==-

t 60 t

i

=--

i 1

r

-...-__....l

.=_-. =. ( - 3., :;0 )

- ( 2n.. :.0 ) -

- - = - -

q

.q.

4 49w e

_E:_

1 I

$==-

l 20

. y.-

......._._..-j.-_,---

a. ~.....

_ _ _ _ _

~

• s _-~- ; _....

.....;"~' ~.. ~ ~...

..=::=.

... - : = : -- -

=-

2-~~~-~

0

-50

-40

-30

-20

-10 0

10 20 30 40 50 l

Flux Difference (AI):

FIGURE 3.2-1 AX:AL FLUX DIFFERENCE LIMITS As A FUNCTICN OF RATED THERFAL PC'AER S EQ'.:oYAH 1l NIT 2 3/4 2-3

.e.

g e

e n

v y

POWER DISTRIBUTION LIMITS 3/4.2.2 HEAT FLUX HOT CHANNEL FACTOR-F {_ZJ Z

q LIMITING CONDITION FOR OPERATION

3. 2. '- F (Z) shall be limited by the following relationships:

q F (Z) 5 [2.237] [K(Z)] for P > 0.5 0

P F (Z) 5 [2.237] [K(Z)] for P $ 0.5 0

0.5 THERMAL POWER where P =

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

APPLICABILITY: MODE 1 ACTION:

With F (Z) exceeding its limit:

9 a.

Reduce THERMAL POWER at least 1% for each 1% F (Z) exceeds the limit q

within 15 minutes and similarly reduce the Power Range Neutron Flux-High Trip Setpoints within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; POWER OPERATION may proceed for up to a total of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />; subsequent POWER OPERATION may proceed provided the Overpower Delta T Trip Setpoints (value of K ) have been reduced at least 1% (in AT span) for each 1% F (Z) 4 9

exceeds the limit.

b.

Identify and correct the cause of the out of limit condition prior to increasing THERMAL POWER; THERMAL POWER may then be increased provided F (Z) is demonstrated through incore mapping to be within q

l its limit.

l SURVEILLANCE REQUIREMENTS i

l 4.2.2.1 The provisions of Specification 4.0.4 are not applicable.

SEQUOYAH - UNIT 2 3/42-4

.o POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) 4.2.2.2 F (z) shall be evaluated to determine if F (Z) is within its q

q limit by:

Using the movable incore detectors to obtain a power distribu-a.

tion map at any THERMAL POWER greater than 5% of RATED THERMAL POWER.

b.

Increasing the measured F component of the power distribution qg map by 3 percent to account for manufacturing tolerances and further increasing the value by 5% to account for measurement uncertainties.

c.

Satisfying the following relationship:

FQ (z) < P x W(z)x Mz) for P > 0.5 M

.237 F "(z) < W(z) x 0.5x Mz) for P < 0.5 2.237 Q

where F"(z) is the measured F (z) increased by the allowances for q

manufacturing tolerances and measurement uncertainty, F limit is q

the F limit, K(z) is given in Figure 3.2-2, P is the relative q

THERMAL POWER,and W(z) is the cycle dependent function that accounts for power distribution transients encountered during normal operation.

This function is given in the Peaking Factor Limit Report as per Specification 6.9.1.14.

d.

Measuring F "(z) according to the following schedule:

q 1.

Upon achieving equilibrium conditions after exceeding by 10 percent or more of RATED THERMAL POWER, the THERMAL POWER at which F (z) was last determined,* or q

2.

At least once per 31 effective full power days, whichever i

occurs first.

• 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 a power distribution map obtained.

l l

l SEQUOYAH - UNIT 2 3/4 2-S l

l

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) e.

With measurements indicating f

I" f*)

f maximum Q

k K(z) over z

"(z) either has increased since the previous determinatin of Fq of the following actions shall be taken:

1.

F M(z) shall be increased by 2 percent over that specified in q

4.2.2.2.c, or 2.

F "(z) shall be measured at least once per 7 effective full q

power days until 2 successive maps indicate that F" (z) maximum is not increasing.

over z K(z) j f.

With the relationships specified in 4 2.2.2.c above not being 2

satisfied:

1.

Calculate the percent F (z) exceeds it's limit by the following 9

expression:

)

f

+

Fg (z) x W(z)

-1 x 100 for P > 0.5 maximum

$ over z 2.237 K(z)

.l

\\l q

N Fq (z) x W(z)

-1 f x 100 for P < 0.5 maximum (overz 2.237 x K(z) l 0.5

)

t

./

g 2.

Either of the following actions shall be taken:

a.

Place the core in an equilibrium condition where the limit in 4.2.2.2.c is satisfied.

Power level may then be increased provided the AFD limits of Figure 3.2-1 are reduced 1% AFD for each percent F (z) exceeded its limit, q

or b.

Comply with th6 requirements of Specification 3.2.2 for F (z) exceeding its limit by the percent calculated above.

q SEQUOYAH - UNIT 2 3/4 2-6

j.

5

/

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS (Continued) i g.

Thelimitsspecifiedin4.2.2.2.c[4.2.2.2.e,and4.2.2.2.fabove are not applicable in the following core plane regions:

I 1.

Lower core region 0 to 15 percent inclusive.

2.

Upper core region 85 to 100 percent inclusive.

4.2.2.3 When F (z) is measured for reasons other than meeting the requirements q

of Specification 4.2.2.2 an overall measured F (z) shall be obtained from a q

power distrit otion map and increased by 3 percent to account for manufacturing i

tolerances for further increased by 5 percent to account for measurement uncerta!fity.

l 2

t 4

l I

(

{

l SEQUOYAH - UNIT 2 3/42-7 I

I

~

-...--=-... - --

- ~ - -

-- - ~ ~ ~

POWER DISTRIBUTION LIMITS 3/4.2.3 RCS FLOWRATE AND R J

LIMITING CONDITION FOR OPERATION The combination of indicated Reactor Coolant System (RCS) total 3.2.3 shall be maintained within the regions of allowable flow rate and R), R2 operation shown on Figure 3.2-3 for 4 loop operation:

Where:

N FAH a.

R

=

j 1.49 [1.0 + 0.3 (1.0 - P)]

b.

R 2

[1 - BP (Bu)]

THERMAL POWER p

RATED THERMAL POWER '

l Measured values of F obtained by using the movable d.

F

=

g g

incore detectors to obtain a power distribution map.

The measured values of F shall be used to calculate g

R since Figure 3.2-3 includes measurement uncertainties of 3.5% for flow and 4% for incore measurement of F g, and e.

RBP (Bu) =

Rod Bow Penalty as a function of region average burnup as shown in Figure 3.2-4, where a region is defined as those assemblies with the same loading date (reloads) or enrichment (first core).

APPLICABILITY: MODE 1 ACTION:

utside the regions With the combination of RCS total flow rate and R, R2 j

of acceptable operation shown on Figure 3.2-3:

a.

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

1.

Either restore the combination of RCS total flow rate and R, R to within the above limits, or j

2 2.

Reduce THERMAL POWER to less than 50% of RATED THERMAL POWER and reduce the Power Range Neutron Flux - High trip setpoint'to less than or equal to 55% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SEQUOYAll - UNIT 2 3/4 2-8

.=

t vs m

o C

i o

I 4

40

]

.....6..,,,,eti...t....b,,,,..,4,........asi,,..6..r+.-

i r

' j l j}

! MEASUT1EMENT UNCERTAINTIES OF

! f f-lgl-f j

tj f

'}fi:'1lII

-'- ll:!

MEASUREMENT OF F AREI CLUDED l"

i I~

lj[:1 Il E

E L IN THIS FIGURE.

[ 'l' 7'

i OPE R ATION

.;!.. j.h.,

J j p

.7.

l ACCEPTABLE 46

=

7

..y 7-

.ff h

! h

. [

..j 8

1 L

. 1.

j

}.9}.

j ; REGION FOR,j

-f, L.

g y

L 1....}.

4M[..

.., 7

.}

.,[ f,:ii

- [. l j

-R,ONLY

]j {

. gfj[j ji.. jjj 8

[ q

}i

i 2

i ir :t.au n2.tt::..u rpp

• tj... 4 UNACCEPTABLE j t

'- I b- -

44 4.h.

.i.-42...

.J 11.4..ji

-t

++f

3 l

}

)

I...!

'd,f.

{

l f OPERATION fl." l ~ 4 .. 1[.l . j!..-.,.

7..4l

....l! .. f jp{. -t t' { l REGION i t{. } ~ p:~' j ;+, g . ; g{ q I a m i , il .j.. - t i n: i 1 g 1 ffylt- !{ fit;l{ g 42 L 1 -j-- ACCEPTABLE - 1 -j l l OPE R ATION

]

l. l!..

. l. f. lt..! t.! w a t i ..t,.. r s u. }. Ii t j j.] i.2t1. REGION FOR J 7 .j'-- -(( --{- - +- jf. ! ' l j (1.029,40.68[ i R, 8 R, jl, I i S -+t- +t b( 1 t T-l 1 j I+t d 44 t l t- } t 0pl 1 40 ~ t U

o. 1j.l..

v.5,s j

j t

z f 3. q j St...7. p l t 'h ht1 ~ I ' t" H l'T 38 .4t. .. i .i . i i H ' (1.0, 37.84 ) i [j UNACCEPTABLE i t OPERATION h + - REGION !~ .l I illillilllifill!!illilillilli i 3g 0.90 0.92 0.94 0.96 0.98 1.00 1.02 1.04 1.06 R, = Fh/1.49[1.0 + 0 3(1.0 - P)) R = R,/[1 - RBP(Bu)] 2 FIGURE 32 3 RCS Total Flowrate Versus R and R2 - Four Loops in Operation 3

INSTRUMENTATION J MOVABLE INCORE DETECTORS ~~ LIMITING CONDITION FOR OPERATION 1 3.3.3.2 The movable incore detection system shall be OPERABLE with: a. At least 75% of the detector thimbles, b. A minimum of 2 detector thimbles per core quadrant, and c. Sufficient movable detectors, drive, and readout equipment to map these thimbles. APPLICABILITY: When the movable incore detection system is used for: a. Recalibration of the excore neutron flux detection system, l b. Monitoring the QUADRANT POWER TILT RATIO, or c. Measurement of F and F (Z). g q ACTION: With the movable incore detection system inoperable, do not use the system for i the above applicable monitoring or calibration functions. The provisions of Specifications 3.0.3 and 3.0.4 are not applicable. SURVEILLANCE REQUIREMENTS 4.3.3.2 The movable incore detection system shall be demonstrated OPERABLE by normalizing each detector output when required for: r a. Recalibration of the excore neutron flux detection system, or b. Monitoring the QUADRANT POWER TILT RATIO, or c. Measurement of F and F (Z). H q SEQUOYAH - UNIT 2 3/4 3-44

t. CONTAINMENT SYSTEMS AIR TEMPERATURE LIMITING CONDITION FOR OPERATICN I

3. 6.1. 5 Primary containment average aic temperature shall be maintained:

a. between 85*F" and 105*F in the containment upper compartment, anc b. between 100*F" and 125'F in the containment lower compartment. APPLICABILITY: MODES 1, 2, 3 and 4. ACTION: With the containment averaga air temperature not conforming to the above limits, restore the air temperature to within the limits within 8 hours or be in at least HOT STANDSY witnin the next 6 hours and in COLD SHUTDOWN within the following 30 hours. SURVEILLANCE REGUIREMENTS 4.6.1.5.1 The primary containment uppcr compartment average air tamcerature shall be the weignted average"" of all ambient air temperature monitoring stations located in the upper compartment. As a minimum, temoerature readings will be obtained at least once per 24 hours from the following locations: Location a. tiev. 743 ft. b. Elev. 786 ft. I c. Elev. 786 or 845 ft. 4.6.1.5.2 The primary containment lower compartment average air temperature shall te the weighted average"" of all ambient air temperature monitoring stations located in the lower compartment. As a minimum, temperature readings l will be obtained at least once per 24 hours from the follcuing locations: Location a. tiev. 722 ft. b. Elev. 700 ft. l c. Elev. 685 or 703 ft. tower iimit may be reduced to 60*F in MODES 2, 3 and 4. ^

• " The weighted average is the sum of each te.tperature multiplied.by its l

respective containment volume fr' action. In the event of inopersole temperature sensor (s), the weignted average snall be taken as tne reduced total divided by one minus the volume fraction represented by the sensor (s) l out of service. SECC0YAH - UNIT 2 3/4 6-10 w v

C*,MTAI WENT SYSTEMS ' CdNT AIMENT VENTILATICN SYSTEM ^ LIMIT *NG CONDITION FOR OPERATION lines) of con-Three' pairs (three purge supply lines and three purge exhaus 3.6.1.9 taineent purge system lines may be cpen; the cpntainment purge supply anc isolation valves in all othar containment purge lines shall be closec. for either purgir.g ennaust Operation with purga supply or exhaust isolation valves open or venting shall be limited to less than or ecual to 1000 hours per 365 cays. APD L:CA83 f LI~Y: MODES 1, 2, 3, and 4 ACTION: With a purge supply or exhaust isolation valve open in excess of the above cc.ulative limit. or with more than one pair of containment purge system lines open. close the isolation valve (s) ~in the purge line(s) within one hour or ce least HOT STAND 3Y within the next 6 hours and in COLD SHUTCCVN within in at tne following 30 hours. ~ SU4vEILLANCE RECUIREMENTS The position of the containment purge supply and exhaust isolation 4.6.1.9.1 valves shall be determined at least cnce per 31 days. The cumulative time that the purge supply and exhaust isolation 4.6.1.9.2 valves are open during the past 365 dayt saall be determined at least once per 7 days. O A SEQUOYAH - UNIT 2 3/4 6-15 - fn ,j; V[fc y 7 Ip 'e 8, c: C i* e ._:. 3 - -. ~ % -~,L u :. z.;,v. _.3._ s r.s...: _.

l IlYDT0GFft tilTICATI0il SYSTF.M _L1111TIf1G C0!10ITI0tt FOR OPERATION

3. 6. 4. 3 The primary containment hydrogen mitigation system shall be operable.

APPIIn 3ILITY: 1100ES 1 and 2. l ACTI0ti: With one train of hydrogen mitigation system inoperable, restore the inoperable train to OPERABLE status within 7 days or increase the surveillance interval of S.R. 4.6.4.3 frca 92 days to 7 days on the operable train until the incperable train is returned to OPEP.ABLE status. SURVEILLAt1CE REQUIREMENTS 4.6.4.3 The hydrogen mitigation system shall be demonstrated OPERABLE: At least once per 92 days by energizing the supply breakers and a. _ verifying that at least 66 of 68 igniters are energized.* b. At least once per 18 months by verifying the temperature of each igniter is a minimum of 1700 F ~_

• Inoperable igniters must not be on corresponding redundant circuits which provide coverage for the same region.

SEQUOYAH - UNIT 2 3/4 6-26 .2..n ;; a. _ :,.u =n. a...-... = - ~. ~ e.. :- - a-. . -. ~ -. ~ - -. -. = m

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) lirr.iting the fission gas release, fuel pellet temperature and cladding mechanical properties to within assumed design criteria. In addition, limiting the peak linear power density during Condition I events provides assurance that the initial conditions assumed for the LOCA analyses are met and the ECCS acceptance criteria limit of 2200 F is not exceeded. The definitions of certairi hot channel and peaking factors as used in these specifications are as follows: F (Z) Heat Flux Hot Channel Factor, is defined as the maximum local 0 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. F" Nuclear Enthalpy Rise Hot Channel Factor is defined as the ratio of the ikegraloflinearpoweralongtherodwiththehighestintegratedpowccto the average rod power. 3/4.2.1 AXIAL FLUX DIFFERENCE (AFD) The limits on AXIAL FLUX DIFFERENCE assure that the F (Z) upper bound q envelope of 2.237 times the normalized axial peaking factor is not exceeded during either normal operation or in the event of xenon redistribution follow-l ing power changes. Provisions for monitoring the AFD on an automatic basis are derived from the plant process computer through the AFD Monitor Alarm. The computer 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 AI-Power operating space and the THERMAL POWER is greater than 50 percent of RATED THERMAL POWER. i l 3/4.2.2 and 3/4.2.3 HEAT FLUX HOT CHANNEL FACTOR, RCS FLOWRATE AND NUCLEAR ENIHALPY RISE HOT CHANNEL FACTOR The limits on heat flux hot channel factor, RCS flowrate, and 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 temperature will not exceed the 2200 F ECCS acceptance criteria limit. SEQUOYAH - UNIT 2 B 3/4 2-1

i POWER DISTRIBUTION LIMITS BASES Each of these is measurable but will n rmally only be determined periodically as specified in Specifications 4.2.2 and 4.2.3. This periodic j surveillance is sufficient to insure that the limits are maintained provided: Control rods in a single group move together with no individual rod a. insertion differing by more than + 13 steps from the group demand position. b. Control rod groups are sequenced with overlapping groups as described in Specification 3.1.3.6. c. The control rod insertion limits of Specifications 3.1.3.5 and 3.1.3.6 are maintained, d. The axial power distribution, expressed in terms of AXIAL FLUX DIFFERENCE, is maintained within the limits. F"g will be maintained within its limits provided conditions a. through

d. above are maintained. As noted on Figures 3.2-3 and 3.2-4, RCS flow and F"g may be " traded off" against one another to ensure that the calculated DNBR wilI not be below the design DNBR value.

The relaxation of F as a H function of THERMAL POWER allows changes in the radial power shape for all permissible rod insertion limits. When RCS flow rate and F are measured, no additional allowances are g necessary prior to comparison with the limits of Figures 3.2-3 and 3.2-4. Measurement errors of 3.5 percent for RCS total flow rate and 4 percent for F g have been allowed for in determination of the design DNBR value. R), as calculated in Specification 3.2.3 and used in Figure 3.2-3, accounts for F"g less than or equal to 1.49. This value is the value used in the N various safety analyses where F influences parameters other than DNBR, e.g. 3g peak clad temperature, and thus is the maximum "as measured" value allowed. R, as defined, allows for the inclusion of a penalty for Rod Bow on DNBR 2 only. Thus, knowing the "as measured" values of F and RCS flow allow for g " trade off" in excess of R equal to 1.0 for the purpose of offsetting the Rod Bow DNBR penalty. f I SEQUOYAH - UNIT 2 B 3/4 2-2

POWER DISTRIBUTION LIMITS THIS FIGURE DELETED l l l [ Figure B 3/4 2-1 TYPICAL INDICATED AXIAL FLUX DIFFERENCE VERSUS THERMAL POWER SEQUOYAH - UNIT 2 8 3/4 2-3

POWER DISTRIBUTION LIMITS BASES N The penalties applied to F toaccouniforRodBow(Figure 3.2-4)asa g function of burnup are consistent with those described in Mr. John F. Stolz's (NRC) letter to T. M. Anderson (Westinghouse) dated April 5, 1979 and W 8691 Rev. 1 (partial rod bow test data). When an F measurement is taken, both experimental error and manufacturing q tolerance must be allowed for. 5 percent is the appropriate allowance for a full core map taken with the incore detector flux mapping system and 3 percent is the appropriate allowance for manufacturing tolerance. M The hot channel factor Fq (z) is measured periodically and increased by a cycle and heignt dependent power factor, W(z), to provide assurance that the limit on the hot channel factor, F (z), is met. W(z) accounts for the effects ) q of normal operation transients and was determined from expected power control 1 maneuvers over the full range of burnup conditions in the core. The W(z) function for normal operation is provided in the Peaking Factor Limit Report per Specification 6.9.1.14. ~ 3/4.2.4 QUADRANT POWER TILT RATIO The quadrant power tilt ratio limit assures that the radial power distri-bution 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 two hour time allowance for operation with a tilt condition greater than 1.02 but less than 1.09 is provided to allcw identification and cor-rection of a dropped or misaligned rod. In the event such action does not correct the tilt, the margin for uncertainty on F is reinstated by reducing n the power by 3 percent from RATED THERMAL POWER f5r each percent of tilt in excess of 1.0. SEQUOYAH - UNIT 2 8 3/4 2-4

i POWER DISTRIBUTION LIMITS BASES i 3/4.2.5 DN8 PARAMETERS 1 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 1.3 throughout each analyzed transient. The 12 hour periodic surveillance of these parameters through instrument readout is sufficient to ensure that the parameters are restored within their j limits following load changes and other expected transient operation. l l i e i d 1 i i l SEQUOYAH - UNIT 2 8 3/4 2-5 l s.4 %,_ _, z e*,

~_ t. CONTAINMENT SYSTEMS BASES c n 3/4.6.1.8 EMERGENCY GAS TREATMENT SYSTEM (EGTS) The OPERABILITY of the EGTS cleanup subsystem ensures that during LOCA conditions, containment vessel leakage into the annulus will be filtered through the HEPA filters and charcoal adsorber trains prior to discharge to the atmosphere. This requirement is necessary to meet the assumptions used in the accident analyses and limit the sitt boundary radiation doses to within the limits of 10 CFR 100 during LOCA conditions. Cumulative operation of the system with the heaters on for 10 hours over a 31 day period is suf ficient to reduce the buildup of moisture on the absorbers and HEPA filters. ANSI N510-l' will be used as a procedural guide for surveillance testing. 3/4.6.1.9 CONTAINMENT VENTILATION SYSTEM Use of the containment purge lines is restricted to three pairs (one supply line and one exhaust line) of purge system lines at a time to ensure that the site boundary dose guidelines of 10 CFR Part 100 would not be exceede-in the event of a loss of coolant accident during purging operaticos. The analysis of this accident assumed purging through the largest pair of lines (a 24 inch inlet line and a 24 inch outlet line), a pre-existing iodine spike in the reactor coolant and four second valve closure times. 3/4.6.2 DEPRESSURIZATION AND COOLING SYSTEMS 3/4.6.2.1 CONTAINMENT SPRAY SYSTEM The OPERABILITY of the containment spray system ensures that contair. ment depressurization and cooling capability will be available in the event of a LOCA. The pressure reduction and resultant lower containment leakage rate are l consistent with the assumptions used in the accident analyses. 3/4.6.3 CONTAINMENT ISOLATION VALVES ^ The OPE.RABILITY of the containment isolation valves ensures that the containment atmosphere will be isolated from the outside environment in the event of a release of radioactive material to the containment atmosphere or pressurization of the containment. Containment isolation within the time limits specified ensures that the release of radioactive material to the environment will be consistent with the assumptions used in the analyses for a LOCA. 3/4.6.4 COM8USTISLE GAS CONTROL The OPERABILITY of the equipment and systems required for the detection and control of hydrogen gas ensures that this equipment will be available to maintain the hydrogen concentration within containment below its flammaole l l limit during post-LOCA conditions. Either recombiner unit or the purge syster SEQUOYAH - UNIT 2 B 3/4 6-3 s .= : aw ux a : =a.: a n..-. ...ww.. a :...;..

a, CONTAINMENT SYSTEMS BASES COMBUSTIBLE GAS CONTROL (Continued) is capable of controlling the expected hydrogen generation associated with

1) zirconium water reactions, 2) radiolytic decomposition of water and

3) corrosion of metals within containment. These hydrogen control systems are consistent with the recommendations,of Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Co'ntai,nment Following a LOCA", March 1971.

The hydrogen mixing systems are provided to ensure adequate mixing of the containment atmosphere following a LOCA. This mixing action will prevent localized accumulations of hydrogen from exceeding the flammable limit. The operability of at least 66 of '68 igniters in the hydrogen control distributed ignition system will maintain an effective coverage throughout the containment. This system of ighitors will initiate combustion of any signifi-cant amount of hydrogen released after a degraded core accident. This system is to ensure burning in a controlled manner as the hydrogen is released instead of allowing it to be ignited at high concentrations by a random ignition source. 3/4.6.5 ICE CONDENSER The requirements associated with each of the components of the ice condenser ensure that the overall system will be available to provide sufficient pressure suppression capability to limit the containment peak pressure transient to less than 12 psig during LOCA conditions. 3/4.6.5.1 ICE BED The OPERABILITY of the ice bed ensures that the required ice inventory will 1) be distributed evenly through the containment bays, 2) contain suffi-cient boron to preclude dilution of the containment sump following the LOCA and 3) contain sufficient heat removal capability to condense the reactor system volume released during a LOCA. These conditions are consistent with the assumptions used in the accident analyses. The minimum weight figure of 1200 pounds of ice per basket contains a 10% conservative allowance for ice loss through sublimation which is a factor of 10 higher than assumed for the ice condenser design. The minimum weight figure of 2,333,100 pounds of ice also contains an additional 1% conservative allowance to account for systematic error in weighing instruments. In the event that observed sublimation rates are equal to or lower than. design predictions after three years of operation, the minimum ice baskets weight may be adjusted downward. In addition, the number of ice baskets required to be weighed each 9 months may be reduced after 3 years of operation if such a reduction is supported by observed sublimation data. SEQUOYAH - UNIT 2 83/46-4

. w

+ .... ~. _. -

s. ADMINISTRATIVE CONTROLS e. An unplanned offsite release of 1) more than 1 curie of radioactive material in liquid effluents, 2) more than 150 curies of noble gas in gaseous effluents, or 3) more than 0.05 curies of radiciodine in gaseous effluents. The report of an unplanned offsite release of radioactive material shall include the following information: 1. A description of the event and equipment involved. 2. Cause(s) for the unplanned release. 3. Actions taken to prevent

• recurrence.

4. Consequences of the unplanned release. f. Measured levels of radioactivity in an environmental sacoling medium determined to exceed the reporting level values of Tacle 3.12-2 when averaged over any calendar quarter sampling period. 'l RADIAL PEAKING FACTOR LIMIT REPORT 6.9.1.14 The W(z) function for normal operation shall be provided to the Direc-tor, Nuclear Reactor Regulation, Attention, Chief of the Core Performance Branch, U.S. Nuclear Regulatory Commission Washington, D.C. 20555 at least 60 days prior to cycle initial criticality. In the event that these values would be submitted at some other time during core life, it will be submitted 60 days prior to the date the values would become effective unless otherwise exempted by the Commission. Any information needed to suport W(z) will be by request from the NRC and need not be included in this report. SPECIAL REPORTS 6.9.2 Special reports shall be submitted to the Director of the Office of Inspection and Enforcement Regional Office within the time period specified for each report. 6.10 RECORO RETENTION In addition to the applicable record retent1on requirements of Title 10 Code of Feceral Regulations, the following records shall be retained for at least the minimum period indicated. 6.10.1 The following records shall be retained for at least five years: Records and logs of unit operation covering time interval at each a. power level. b. Records and logs of principal maintenance activities, inspections, repair and replacement of principal items of equipment related to nuclear safety. SEQUOYAH - UNIT 2 6-27

4 -_--e An--- w l. 9 t APPENDIX B PEAKING FACTOR LIMIT REPORT l l l i i i 'l ? i r f l l

Sequoyah Nuclear Plant Unit 2 - Peaking Factor Limit Report This peaking factor limit report is provided in accordance with paragraph 6.9.1.14 of the Sequoyah unit 2 technical specifications. The cycle 2 W(z) function for RAOC operation is shown in figure 1. W(z) was calculated using the method described in NS-EPR-269, letter from E. P. Rahe (Westinghouse) to C.11. Burlineer (NRC) August 31, 1982. This U(z) function is used to confirm that the heat flux hot channel factor, F (z), will be limited to the technical specification values of: g 2.237 [ K m): for p > o. 5, and F (z) s o r (=) 1 4.474.K (a)] fo r P s 0 5' ~ o This W(z) function, when applied to a power distribution measured under equilibrium conditions, demonstrates that the initial conditions assumed in the LOCA are met along with the ECCS acceptance criteria of 10 CFR 50.46. e l i e e 1-+ p:._

4. HEIGHT MAA t.ie (FEET) w(2) hae I 15 0.000

• e 45 0.000
• t.

75 0.000

• 1.05 0.009
• t *8 1.35 0.000
• 1.65 1.379 u,

1.95 1.328 i.s 2.25 1.280 2.55 1.237 km 2.85 1.198 3,15 1.174 3.45 1.177 g, y 3.75 1.177 4.05 1.180 t.m 4.35 1.183 3 4.65 1.179 1*

• 4.95 1.179

,,1 g3 5.25 1.185 D 5.55 1.202 5.85 1.218 a. g 6.15 1.233 g 6.45 1.243 5.# 8 6.75 1.247 8" 8 7.05 1.245 8 7.35 1.237 t.te 8 ,a 7.65 1.223 7.95 1.205 t.te 8.25 1.185 l 1* 8.55 1.165 8.85 1.167 g g, 9.15 1.192 t.se 9.45 1.209 9.75 1.234 8" 10.05 1.279 g,g 10.35 1.329 10.65 0.000

• t.o.

10.95 0.000 + 11.25 0.000 + 88 11.55 0.000

• 11.85 0.000 +

w e as t.se a.m se %se has too se a,as s.es ia m the sa m m 410# FET)

• T0p and bottom 15% excluded as per Technical Specification 4.2.2.2.g l

FIGURE 1 SEQUOYAH UNIT 2, CYCLE 2 RAOC W(z) FOR CYCLE BURNUPS BETWEEN 0 AND 1000' l nd

e.- O i.m Lee H E IGl4T MAX UEET) w(2) .15 0.000

• i,,,

45 0.000

• t.e 75 0.000
• 1.05 0.000
• 8*
• 1.35 0.000
• 1.65 1.349 g,

1.95 1.303 2.25 1.259 t.m a 2.55 1.220 LD 2.85 1.185 3,15 1.168 3* 8 3.45 1.173 a 3.75 1.175 i g g,z 4.05 1.181 .m A.35 1.186 y IX r 4.65 1.184 a ' 4.95 1.191 s b 5.25 1.201 g,3 3 ~ 5.55 1.218 .-y i.a 5.85 1.236 6.15 1.252 km 6.45 1.262 ,a r' 6.75 1.266 8*38 7.05 1.263 a aa E 7.35 1.255 .se 7.65 1.240 L+ 7.95 1.221 8.25 1.196 Lu 8.55 1.174 8.85 1.173 8 9.15 1.195 9.45 1.209 .s. 9.75 1.229 t.se 10.05 1.268 10.35 1.318 8 8' 10.65 0.000 + 10.95 0.000 4 g, 11.25 0.000

• Lee 11.55 0.000 +

e.as t.as a.se s.se e.se ses e.as 1.as e.as e.as ie.oo suas ta.as 11.85 0.000

• cme Eloft rtrn
• Top and bottom 15% excluded as per Technical Specification 4.2.2.2.g FIGURE 2 SEQUOYAH UNIT 2, CYCLE 2 RAOC W(z)

FOR CYCLE BURNUPS BETWEEN 1000 AND 2000 l l

' de t.m HEIGHT MAX t*** (FEET) W(Z) 1.as .15 0.002 t. 45 0.000

• 75 0.000
• t **

1.05 0.000

• 1.35 0.000,

1.65 1.307 t.m 1.95 1.270 2.25 1.231 12 2.55 1.193 2.85 1.168 8" " 3.15 1.159 .g 3.45 1.1C 3.75 1.173 a t.m a: 6 4.05 1.182 r r E 4.35 1.192 88 x 4.65 1.195 g,,, a 4.95 1.213 5.25 1.231 ,,J t.3 5.55 1.248 r 5.85 1.269 t.n a 8 6.15 1.287 6.45 1.297 3* "

4 6.75 1.300 3,i, 7.05 1.296 8

7.35 1.286 t.te 7.65 1.269 7.95 1.246 ta t* 8.25 1.216

1. 12 8.55 1.191 8.85 1.186 t.m 9.15 1.199 9.45 1.208 1m 9.75 1.222 10.05 1.254 I""

10.35 1.301 t.o. 10.65 0.000 10.95 0.000 +

t..

11.25 0.000 * .g:ggg:

t..,.

s. m. t. t,.. m. CM E!CHT (FETI

• Top and bottom 15% excluded a5 per Technical Specification 4.2.2.2.g FIGURE 3 SEQUOYAH UNIT 2, CYCLE 2 RA0C W(z)

FOR CYCLE BURNUPS BETWEEN 2000 AND 4000 I

6. ,e i i t.m i.ee HEIGHT MAX (FEET) W(Z) s.es .15 0.000

• 8 **

45 0.000

• gg 75 0.000
• 1.05 0.000
• t.ao 1.35 0.000
• 1.65 1.246

'8 1.95 1.220 g,, 2.25 1.191

• 8 6>

~ 2.55 1.167 2.85 1.144 3,15 1.146 1.2 3.45 1.157 3.75 1.168 9 8' " 4.05 1.182 2 g,a 4.35 1.201 4.65 1.224 E t.x 4.95 1.254 i b a a 5.25 1.279 5" t 5.55 1.299 ,_o 5.85 1.323 3,3 d 8 ,x 6.15 1.341 - La 6.45 1.351 8 6.75 1.353 .te 7.05 1.347 8 7.35 1.334 L' 7.65 1.312 N 7.95 1.282 a i, n 8.25 1.245 Lu 8.55 1.24 3 8.85 1.206 Lie 9.15 1.201 9.45 1.206 g,,, 9.75 1.214 Le 10.05 1.234 10.35 1.275 s.co 10.65 0.000 4 10.95 0.000 + ^1 11.25 0.000 + - um 11.55 0.000

• a,e i.as a.es sm e,e ses e.se t.se a,e see te.m thee than 11.85 0.000
• CIFE EIGif FEET)
• T0p and bottom 15% excluded as per l

Technical Specification 4.2.2.2.g l FIGURE 4 SEQUOYAH UNIT 2, CYCLE 2 RA0C W(z) FOR CYCLE BURNUPS BETWEEN 4000 AND 8000 l 1

'e. s.5s HEIGHT MAK l FEET) WCZ) g, q t.e .15 0.000

• 45 0.000
• 75 0.000
• l' "

1.05 0.000

• g,,

1.35 0.000

• t.e l

1.65 1.272 1.95 1.243 8 2.25 1.218 3 2.55 1.193 a s 2.85 1.166 3.15 1.136

t. >

3.45 1.147 t.st 3.75 1.162 4.05 1.181 4.35 1.203 9 8* 8 a 2 4.65 1.253 x 4.95 1.281 ^ 8 5.25 1.307 I t.m 8 c 5.55 1.330 g 83 8 8 5.85 1.353 E 6.15 1.370 d 6.45 1.380 6.75 1.380 1 a.as 8 7.05 1.372 t.is 7.35 1.355 7.65 1.329 t.ts 7.95 1.295 I 8.25 1.256 g,,, x 8.55 1.232 8.85 1.218

8. t 9.15 1.201 a.is 9.45 1.207 9.75 1.231 10.05 1.250 10.35 1.255 8.as 10.65 0.000 **

a.m 10.95 0.000* 11.25 0.000

• 11.55 0.000 +

11.85 0.000 + 8.se sm t.as a.m sm e.as ses s.m saa se sm is m it.se ta.se CIFE EIGHT FET)

• Top and bottom 15% excluded as per Technical Specification 4.2.2.2.g FIGURE 5 SEQUOYAH UNIT 2, CYCLE 2 RA0C W(z)

FOR CYCLE BURNUPS BETWEEN 8000 AND EOL 1 I l

.A x --a., --_-a v v ~ mw h E O e APPENDIX C RELOAD TEST PROGRAM i i l e 9

o. e The reload core design will be verified by performance of the following i tests: 1. Control rod drop times, I 2. Critical boron concentration measurements, 3 Control rod bank worth measurements using rod swap method, 4. Moderator temperature ccarricient measurements, and 'I 5. Flux distribution measurements using the incore flux mapping l system. j l I t 1 a i I i i t l ,_.,_,..m. .m.. ,.,,,m.,

06 a 4. e e APPENDIX D REMOVAL OF ROD CONTROL RESTRICTIONS e S e l i

s. ) 400 Chentnut Street Tower II July 22, 1982 Director of Nucicar Reactor Regulation Attention: Ms. E. Adensam, Chief Licensing Branch No. 4 Division of Licensing U.S. Nuclear Regulatory Commission Washington, DC.20555

Dear Ms. Adensam:

In the flatter of ) Docket Nos. 50-327 Tennessee Valley Authority ) 50-328 In 1979 Uestinghouse Electric Corporation identified to the NRC, Core Performance Branch, by letters dated November 15 and November 28, 1979 (Rcrerence letter numbers NS-TMA-2162 and NS-TMA-2167), a concern with regard to certain assumptions utilized in the dropped rod accident safety i analysis applicable to some Westinghouse NSSS designs. This concern was derived primarily from the potential for an unanalyzed power overshoot while in automatic control following selected dropped rod events which did not result in a reactor trip. The concern was applicable to all Westinghouse plants which rely upon the power range neutron flux high-negative rate reactor trip to mitigate the consequences of the drooped rod accident. Operating plants were notified of an unreviewed safety question under 10 CFR 50.59 and nonoperating plants notified of a significant deficiency under 10 CFR 50.55(e). Westinghouse recommended, and NRC subsequently required, certain operational restrictions above 90-percent power (either manual rod control or restricted rod insertion limits when in automatic rod control) to address this concern on an interim basis and to provide further evaluation. It is our understanding that a meeting was held between memhers of the Core Performance Branch staff and Westinghouse to discuss the Westinghouse dropped rod evaluation process. This process demonstrated that the DNB design basis can be me't for this FSAR Chapter 15 condition II event. We have been notified by Westinghouse that this evaluation process results in conclusions that will allow removal of the interim operating requirements en rod control and insertion. 4 It is also our understanding that an agreement has been reached between Westinghouse and members of the Core Performance Branch staff that the removal of operating requirements would take place after the NRC review of the information subsequently submitted by Westinghouse letter dated 4 . January 20,1982 (Reference letter number NS-EFR-2545). This letter, ' serves as notification that the debpped rod evaluation process documented by Westinghouse letter NS-EPR-2545, dated January 20, 1982 applies to Sequoyah unit 1, cycle 2 and Sequoyah unit 2, cycle 2. l The results confirm that the DN3 design basis is met for the dropped rod accident. Based upon-this method, it can be concluded that the interim restrictions on rod control and insertion will no longer be necessary. M O'7 M() l N = ,cz,

e

-:l:.:.;;.".. '

.., ^

• * ' _ :-llI_

Diractor of Nucicar Reactor Fesulation .T'il r ??, 10D We form 111 requent the I;3C to review the raterial ruh::itted t'y Westinghouno (!!S ~'PR-29fi9, Januarv ?0, 1982), ami cu5sequently renovo the int.-in coerationil restrictions effective with the ntartup of crele 2 for both units. Appro*11 in n=eded b-fore startup of S*quoyah unit 1, cycle 2, presently schedulard for refueling cuta ;- in Santenher 103?. ' ~ If you have any quentienn concerning this mtter, please get in te sch uith J. E. WO.13 at FTS 36R-?f,31. Very truly ycntn, TTPl?.SS8:2 VALLP.Y AUTHORITY 1. s 4 (.,-] *, y L. M. !itll:, 'Snager , (.. * * .,2, !!ucicar Licensing s vo-n fi^. a d subacathed before :e '. : Ctsin g%2 ay of ShtAtr 1932 .l' I (h/h ) lr flotar7 Public g "'/ ?,ff.f,chw1:sf on Expires 7 d" ~ / ~ _ cc: U.S. Nuclear Regulatory Cc :.issien Bezion II Attn: Mr. Ja'es P. O'9e1117, Reginaal.toninistr:ter 101 !'arietta Street, Guite ~1100 Atinnta, Ccorgia 30303 l o }* e age e ge .e.,+.- .-w+** = = = - %.ww e.een,a y

e. ENCLOSURE 2 SIGNIFICANT HAZARDS CONSIDERATION DETERMINATION FOR PROPOSED TECHNICAL SPECIFICATIONS FOR UNIT 2 SEQUOYAH NUCLEAR PLANT 4 P l 4 i I \\ l

1 Item A. The reload safety evaluation changed the value of K from 0.2 to 0.3 in the following relationship: F < 1.55 [( 1.0 + K(1.0 - P)). This in AH turn affects Figure 2.1-1, Overpower and Overtemperature Delta-T Trip Setpoints, f(AI), the Rj, R, FAH relationship, and the 2 associated bases. 1. The probability of an occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in the FSAR is unaffected based on the findings presented in the reload safety evaluation. 2. The changes involve implementing the requirements necessary to ensure the analytical work performed by Westinghouse for cycle 2 remains valid. No equipment modifications are involved and no possibility for an a,ccident or malfunction of a different type than previously evaluated in the FSAR is created. 3 The margin of safety defined in the bases to all technical specifications is maintained. No significant variations in thermal margins result from the cycle 2 reload; however, the reactor core safety limits, figure 2.1.1 in the technical specifi-cations, and the axial offset limits have been revised to reflect the increase in K from 0.2 to 0.3 Increasing the slope of the allowable FAH as a function of the power design limit from 0.2 to 0.3 requires reevaluation of the DNB protection setpoint. The setpoints for Sequoyah unit 2 cycle 2 have been updated to account for this increase in slope. The maximum calculated FAH through the power range of Sequoyah unit 2 cycle 2 has been verified to be less than the value allowed with the 0.3 FSH slope multiplier. The effect on specific parameters is discussed in the Reload Safety Evaluation Report. B. Relaxed axial offset control (RAOC) will be implemented for cycle 2. The axial flux difference technical specification and associated bases ~ must be revised to implement this new control philosophy. i 1. The probability of an occurrence or the consequences of an accident or malfunction of equipment important to safety previously analyzed in the FSAR is unaffected based on the findings presented in the reload safety evaluation. 2. These changes involve tsplem'enting the requirements necessary to ensure the analysis supporting the RAOC philosophy is conformed to in actual plant operations. No equipment changes are involved and no possibility for an accident or malfunction of a differer.t type tha3 previously evaluated in the FSAR is created. l 3 The margin of safety defined in the bases to all technical specifications is maintained. M e n

-~e _g_ Limits on allowable operating axial flux imbalance as a function of l power level considering limiting normal operation and operational transient power distributions are less restrictive than those from LOCA Fg considerations. Axial power distributions used to gener-ate DNB core limits for incidents of moderate frequency were also evaluated, and no changes to these limits were c.equired for RAOC operation. Fxy(z) surveillance currently in the technical specifications is being C. To implement F (z) surveillance replaced by F (z) surveills e. 0 n Lc0 3 2.2 and its associat 2 surveillance requirements, peaking factor limit report and bases must1>e revised. The probability of an occurrence or the consequences of an accident 1. or calfunction of equipment bsportant to safety previously analyzed is unaffected. These changes only change the parameter =onitored to verify Fg remains within its limits. 2. These changes involve replacing the Fxy( ) surveillance requirements with F (z) surveillance. Fxy( ) is L:plicitly g included in the Fn(z) measurement, and the intent of the technical specification is to monitor Fo using a measured parameter. No modification is involved, and no possibility for an accident or malfunction of a different type than previously evaluated in the FSAR is created. 3 The cargin of safety defined in all technical specifications ir maintained. F surveillance is only a change to the plant q surveillance requirements and has no bspact on the results of the cycle 2 or safety parameters. D. The ACTION statement to LCo 3.2.2 has been revised to delete the requirement that the overpower delta-T trip setpoint be adjusted in HOT STANDBY. 1. The probability for and consequences of an accident or malfunction are unaffected by this change. The technical specifications already allow operations with the overpower AT trip setpoint reduced. This change only affects the timing of the satpoint reduction. 2. Reduction of the overpower ST trip setpoint at power does not generate the possibility of any accident or malfunction different from those already analyzed in the FSAR. 3 This technical specification change maintains the margin in the technical specification by requiring the same actions,, i.e., reduction of the overpower AT trip point. i - - -'Mu--- -- -,. M Ib

m.+ ' E. The upper limit on the containment upper and lower. compartment temperatures is being changed to those values used in the new LOCA analysis. 1. This technical specification change does not effect the probability for or the consequences of an accident or malfunction. It only involves raising the upper limits on containment air te=peratures, and the new LOCA analysis shows this is accomplished. 2. This technical specification change does not effect the possibility for an accident or create a malfunction of a different type than previously evaluated in the FSAR. 3 The margin of safety in the basis of this technical specification is maintained. The new LOCA analysis shows the conclusions presented in the FSAR are still valid. F. Change the number of purge lines allowed open for three pairs at any one time (four 24-inch - upper and lower containment, two 12-inch - instrument room). 1. The probability for and the consequences of an accident or malfunction are not increased based on the results of the new LOCA analysis. 2. Changing the number of the purge lines allowed open at any one ti=e does not create the possibility for an accident. or malfunction of a different type. 3 The margin of safety of the technical specifications is =aintained. The new LOCA analysis shows all conclusions presented in the FSAR are still valid. G. Remove the state =ent in the Bases of the Quadrant Power Tilt Ratio which relates the 1~*02 limit to DNB and linear heat generation rate protection. 1, 2, and 3 This is purely an editorial change. The paragraph removed defines the purpose of the 1.02 limit and is only superfluous information. The margin of safety defined in the basis of all technical specifications, the probability of and the consequences for all analysed accidents, and possibility for any new accident are all unaffected. H. Reflect additional hydrogen igniters in the technical specifications. 1, 2, and 3 The igniters perform a mitigation function, and they do not increase the probability of an accident. l 4 i n-m-..:..X .. ~. f

.7#..

.:. l; L ~"~~7> .}}