Revision as of 15:46, 17 February 2020

Text

Proposed Tech Specs 2.2.1,3/4.3.1.1 & 3/4.3.2.1 & Associated Bases,Reflecting Upgrades & Enhancements to Reactor Protection Sys
	ML20006B389
Person / Time
Site:	Sequoyah
Issue date:	01/24/1990
From:	; TENNESSEE VALLEY AUTHORITY
To:	;
Shared Package
ML20006B388	List: ML20006B387; ML20006B389;
References
	NUDOCS 9002020018
	Download: ML20006B389 (348)
	v • d • e

{{#Wiki_filter:~~ L ! ENCLOSURE 1 I l PROPOSED TECHNICAL SPECIFICATION CEANGE { l SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 l 1 DOCKET NOS. 50-327 AND 50-328 I l (TVA-SQN-TS-89-27)

I LIST OF AFFECTED PAGES i Unit 1 Unit 2 l 1-2 1-2 ; 2-5 2-5 ! 2-6 2-6 ! 2-7 2-7 l 2-8 2-8 ! 2-9 2-9 i 2-10 2-10 '! B2-4 B2-4 : B2-5 B2-5 ! B2-6 B2-6 r 3/4 3-2 3/4 3-2 - 3/4 3-3 3/4 3-3 i 3/4 3-5 3/4 3-5 ' 3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 - 3/4 3-9 3/4 3-9 3/4 3-10 3/4 3-10 3/4 3-11 3/4 3-11 - 3/4 3-12 3/4 3-12 > 3/4 3-13 3/4 3-13 ' 3/4 3-15 3/4 3-15 : 3/4 3-16 3/4 3-16 ' 3/4 3-18 3/4 3-18 i 3/4 3-19 3/4 3-19 i' 3/4 3-21 3/4 3-20 3/4 3-21a 3/4 3-21 3/4 3-22 3/4 3-21a 3/4 3-23 3/4 3-22 3/4 3-24 3/4 3-23 3/4 3-25 3/4 3-24 3/4 3-26 3/4 3 i 3/4 3-27 3/4 3-26 3/4 3-27a 3/4 3-27 3/4 3-28 3/4 3-27a 3/4 3-30 3/4 3-28 3/4 3-31 3/4 3-30 3/4 3-33a 3/4 3-31 ; 3/4 3-34 3/4 3-33a 1 3/4 3-36 3/4 3-34 j 3/4 3-37 3/4 3-36 3/4 3-37a 3/4 3 37 j B3/4 3-1 3/4 3-38

B3/4 3-1 900202001s 900124 :

fDR ADOCK 05000327 FDC

k

         .                                                                                                                       t CHANNEL FUNCTIONAL TEST

1. 6 A CHANNEL FUNCTIONAL TEST shall be:

lR71

a. Analog channels - the injection of a simulated signal into the !

channel as close to the sensor as practicable to verify OPERABILITY t including alarm and/or trip functions, j L kg gt b. Bistable channels - the injection of a simulated signal into the I sensor to verify OPERABILITY includj*n$alprm synt tasand/or trip functions. I c, p.f t e Anneli rfe inieSe~ de si='= de < 4 ~~s t as sku : 4 4 ansea inpd' 4 de f enoss ras k as a ts.sh h wef ongsp.rtzry

                                            '"' M*J d'* "W" CONTAINMENT INTEGRITY i

1. 7 CONTAINMENT INTEGRITY shall exist when: lR71 t

a. All penetrations required to be closed during accident conditions

,                        are either:                                                                                             ;

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated automatic valves secured in their closed positions, except as provided in .

Table 3.6-2 of Specification 3.6.3. I

b. All equipment hatches are closed and sealed.

c. Each air lock is OPERABLE pursuant to Specification 3.6.1.3, i i

d .' The containment leakage rates are within the limits of ; Specification 3.6.1.2, and

e. The sealing mechansim associated with each penetration (e.g.,

welds, bellows, or 0-rings) is OPERABLE. 5 CONTROLLEO LEAKAGE

  ;        1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor coolant pump seals.                                                                                            lR75 C_0RE ALTERATION

1. 9 CORE ALTERATION shall be the movement or manipulation of any component '

R75. within the reactor pressure vessel with the vessel head removed and fuel in the vessel. Suspension of CORE ALTERATION shall not preclude completion of movement of a component to a safe conservative position. t SEQUOYAH - UNIT 1 1-2 Amendment No. 12X 71 l May 18, 1988

TABLE 2.2-1 E .

8 REAC10R TRIP SYSTEM I'NSTRUMENTATION TRIP SETPOINTS

                                                                                                                                                  '                      ES L 2i h       FUNCTIONAL UNIT                                      TRIP SETPOINT                               ALLOWA8LE VALUES h         1. Manual Reactur Trip                             Not Applicable                              Not Applicable                                                    '

5 H 279 %

2. Power Range, Neutron Flux Low Setpoint 1 25% of RATED Low Setpoint

             *>                                                                                                                          1-ME of RATED THERMAL POWER                               THERMAL POWER              !

High Setpoint ll1.1% 1 109% of RATED High Setpoint 1--MM-of RATED THERMAL POWER THERMAL POWER

3. Power Range, Neutron Flux, < 5% of RATED THERMAL POWER with < 6.3% of RATED THERMAL POWER High Positive Rate a time constant 2 2 seconds "'8 Uith a time constant 1 2 seconds

4. Power Range, Neutron Flux, < 5% of RATED THERMAL POWER with < 6.3% of RATED THERMAL POWER + a48 High Negative Rate i time constant 1 2 seconds sith a time constant 1 2 seconds

5. Intermediate Range, Neutron i 25% of RATED THERMAL POWER Flux 1 30% of RATED THERMAL POWER

6. Source Range, Neutron Flisx < 105 counts per second 5

i 1.3 x 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note 71 1964 8

9. Pressurizer Pressure--Low 1 1970 psig 14960 psig

                                                                                                                       .n oo.2

10. Pressurizer Pressure--High < 2385 psig

_ 1 6 psig

           ,        11. Pressurizer Water Level--High < 92% of instrument span                                       i         f instrument span i         g f.

Ps 4% gx 12. Loss of Flow 1 90% of design flow per loop

1-694 of design flow per loop
au
Design flow is 91,400 gpm per loop.

           *5$,CP 4

s

        .m y,

TABLE 2.2-1 (Continued) o g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS FUNCTIONAL UNIT TRIP SETPOINT Re l d4 ALLO w/rro g _ _ _ _ _ __ g 13. Steam Generator Water 115 cf nr. /xrs;cp in:;tr =f -r- % c..t r"WA8LE VALUES cf n,.. . a r c.g i tr;;;t R20 y level--Low-Low p : r : - h : t e = ;;..c r a t c r :px ;;;h :te= ;;ccrat-- ' I4. Ste =/F;; icter r?= ' 1 5 Ofrfe?' st--= <? = ct ' 42. 5% f fe!' na: _

                                    . - . .._u.               . . . . .

_2 i__._.c.___

                                                                               -                                         u.  .

v. en., v. o. m . m, na,,ur

                                                                                                                                                      , m. .

__:__:2 _. _ . . . _ . . . u,,,.v.m ra v. .m-.m or - . n.our. -m ter__2..~_1>

                                                                                                                                                                                                                                     '_? =

t . C : r:ter Water Lc;c?

   /IIN                      DeM*/

with ste= ;;nsrator ::ter ?;;;? with :;ts;. g..,.rster acte- 12.;? 2 25% ef n:rc :n;; in:tra 120.5 cf nc. . rn; i ;tra-x .t :p: :::S te = ;;ncratcr .,_c.t :ps ;d.:t = gc.,. rat:-

15. Undervoltage-Reactor -> 5022 volts each bus > 4739 volts each bus 89

Coolant Pumps - rf 16. Underfrequency-Reactor > 56.0 Hz each bus > 55.9 Hz each bus m Coolant Pumps

17. Turbine Trip .

A. Low Trip System > 45 psig > 43 psig Pressure B. Turbine Stop Valve > 1% open > 1% open Closure

18. Safety Injection Input Not Applicable Not Applicable from ESF

19. Intermediate Range Neutron > 1 x 10 -10 amps > 6 x 10~11 amps 27 g, Flux - (P-6) Enable Block y,

a g Source Range Reactor Trip

e g a,s, 12. 4 %

    "" 20. Power         (not P-10) Input to Low Power Range Neutron Flux                                              < 10% of RATED THERMAL POWER
                                                                                                                                                                                  < 4W of RATED THERMAL POWER Reactor Trips Block P-7 E                                                                                                                                                                                                                    g 1, 21 ce O3 C0 Un.

CD on W W __._ m. _ . _ _ . . _ _ _ _ . . _ _ _ _ _ _ _ _ _ __ _ ___m. _ =_____ _. -- - -4 - ., -- .. . - , _ _ , - , . . . . . . - . - _ _ . _ . _ . . _ . . , .

4n,_ m e- ..m,.ek see -sAa< J,- --a m. e a n ,,4a i k \

i O

w v : . 4+ . sh ! b I I d) (

                                                                                ).                        9         9                  I vk
            ,               {  '

I ,' f D vY a tt t[ A' gd' TS NS i 4 h vi M k e f, N

s mg s T
1 s

J K c} 4 .2 4 ' e vt

                                                                               >i d'

A '~ c VI

                                                                                                                    &vs
      't                                                                                                                               l n                                                                         v
                                                                                                                                       )

s b d t e* I

s.

                          .                     1 s                                      Ab                             4                         9          9                  l a

g1 "% 11 d %[S + '* 3 3 '* 9 N s b"' ee f w4 q A aj $ l

     )                   *,h O -                  N!                  D          o4
                                                                              $                            h"       b-p  h                                    a                 vt           41                           vi       v, i

b i i i

                                                     ]k                             T                       -       4 3+ h                                                                               a       s-% 1
                           <w                  xd ;=.
                                                ,                              s w E*
                                                                                                         -s.. .     -4 4           .t Kse g1                             g3-                              -    L' '

yI 47 4'Y E 'd i s L,w%eTs I ,a s is 1 4-4 4

         '$      y Fs
                                             $         1       ).k,  5 N      .i                     ,,

g 1

         'l   d          %d                  l=d')               'l"          d'l~ g3                  d.j        $y w lj j pj    .!

a 8 - n

l 1 l ! l i t i k } l w I I & ! 3 kh NA kI s*s a s w = h : M k \s df , b $4

                                                  .Q4 5

x .! w . .s 9

   't                                                                                                                                 ,

n t D D i I I

  )                                                 %

c1 e q t t t ? Q b ) 5 ( et 9 a f oQ

 .x                                               Mq'g                         c
      &                                           4         s                  y;           c 'as A\

b , Q s s 44 43 3N l' ! I' 44, 4 44 4a i> s$ @^ s d , N .!

                                      $ 'd                                     1            4j   i r      s fs1 ia 4 ereQ                                                       sc                                     ;

l 7 4e

      < ue Yin 4 4*                                                                         74                         -
                                                                                      ,                                               1 V                  N                                                          o Y                     <$

j _ - . _ . ._ . . _ _ . . . . . . _ .. l . . _ _

i m TABLE 2.2-1 (Continued) REACTOR TRIP SYSTEM NSTRUMENTATION TRIP SETPOINTS. gf E FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES E sat.4 se, 3 21. Turbine Impulse Chamber Pressure - < 10% Turbine Impulse < -H% Turbine Impisise

    -             (P-13) Input to Low Power Reactor Trips             Pressure Equivalent                     Pressure Equivalent Block P-7 49 ?,                                     SAV%

22. Power Range Neutron Flux - (P-8) Low < -39% of RATED < -362 of RATED Reactor Coolant Loop Flow, and Reactor -

THERMAL POWER THERMAL POWER Trip

                                                                                                                 - 7. 6 %

23. Power Range Neutron Flux - (P-10) - > 10% of RATED > -9E of RATED Enable Block of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Range (low setpoint) Reactor Trips 7

    "     24. Reactor Trip P-4                                    Not Applicable                           Not Applicable SJL.V5I,

25. Power Range Neutron Flux.- (P-9) - < 50% of RATED < W of RATED Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER Trip Below 50% Rated Power (14 T,S \

t+T,.5) NOTATION NOTE 1: Ov9rtemperature AT ( N< AT, {Ky -K b)[T( M-T'] + K (P-P') 3 - f (AI)) g

                                             #+TS           y           2 (1 + Ig f + T 45 8                           l31             w-*

yg where: = 4eIfcompensator on measured AT f ty,3 = Time constants utilized in the . ., r for AT,*ty* S*C51 7t'3 ##* #' AT, = Indicated AT at RATED THERMAL POWER K y < 1.15 (23 E K = 0.011 2

                                                                                                                                                                ~
      ,                                                                                          TABLE 2.2-1 (Continued)

E g PEACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 x NOTATION (Continued) e . g NOTE 1: (Continued) Z

      "                      1+rp'                                       =

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

             %                                                       J.3 x

RTO . 1 2 ' 'J.z = Time constants utilized in the lead-lag controller for Tavg' *h" 33 **' gI;de 8 t = 4 secs. T = Average temperature 'F 1+ 5 ~ ' #S # ~ E = # t0

                                                                              ~               0" *='= T avg i,                                                  % M ed M :M ' W r : C T,yg M n,x:tr, : - 42 m .-

T' < 518.2'F (Nominal T,,g at RATED THERMAL POWER) K = 0. M 55 W 3 P = Pressurizer pressure, psig P' = 2235 psig (Nominal RCS operating pressure)

                      .S                                                 =   Laplace transform operator (sec-I) and f (al) is a function of the indicated difference between top and bottom detectors i

. > e, of the power-range nuclear ion chambers; with gains to be selected based on measured 5E instrument response during plant startup tests such that: EE 2T (i) for q q arephrcen,between-29percentand+5percentf(aI)=0(whereqtRATEDTHERMALPOWERinthetopandbo and d23

     ?+ "
 ,       ,0                                        and g g +q b is total THERMAL POWER in percent of RATED THERMAL POWER).
     $5                                                                                                                                                   l f3

m TABLE 2.2-1 (Continued) o g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS sx NOTATION (Continued) , E [ NOTE 1: (Continued) (ii) for each percent that the magnitude of (q g exceeds -29 percent, the AT trip set-point shall be automatically reduced by 1.l50 pIr) cent of its value at RATED THERMAL POWER. R23 p0 bf,n - (iii) for each percent that the magnitude of (q - exceeds +5 percent', the AT trip set-gnd# point shall be automatically reduced by 0.l86 hr) cent of its value at RATED THERMAL POWER. ( ,i .r,s

                                                                                                                                                  + %5 NOTE 2:                                      Overpower AT (

C+rS I

                                                                                                                                                        -75 AT, { K4 -K                     N )(

3 (1 + T MT5 h T -K 6D( - T9 - f 2I0III y 4 MT5 4 I * *Ts$ Where: M15

                                                                                                                                                        =   as defined in Note 1 1    ,

ty 7f

                                                                                                                                                        =  as defined in Note 1 Af,                                  =  as defined in Note 1 K

4 5 1.087 118 K 5 = 0.02M for increasing average tvature aM 0 for knasing averay

                             ., 3                                                                                                                         temperature
                            .{g J                                                                                                          T]  y
     ,                      -g g ,.

yT

The function generated by the rate-lag controller for T,,g dynamic compensation SE . _ - . . _ _ - . - - _ . _ _ _ w. - __ . ~ - . + _ ~ , - -.a -..--....n, -. ,n.. . _ . , . . , , ., . , . , . . _ _ _ _ , . , , . . , . . , _ . . .

         ,                                                        TABLE 2.2-1 (Continued)                                                         ~

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS 5! N e NOTATION (Continued)

E a

NOTE 2: (Continued) 2 1 = s Y3 Time constant utilized in the rate-lag controller for T**9, TY = 10 secs. 3 ; g i+T54 ~ "

                                '4 K

g = 0.0011 for T > T" aM Kg = 0 for T $ V U m O o T = as defined in Note 1 T" = Indicated T avg at RATED THERMA 1. POWER (Calibration temperature for AT instrumentation,1578.2'F) S = as elefined in Note 1

                               '/2(AI)   =    0 for all al NOTE 3: -The channel's maximum trip setpoint shall not exceed its computed trip point by more than

, l y

     ,ir               ypercent,. 4r .r     .

3 . f.9 59/. n m=_-- NorE 4: T$e deonarl' auw'*"* SQ Y 'L' "' Y

     ;;6                 a m.+         r,,.
     ?-

Q f NOTE 5: *-INSEMT "l} ' TTO

a . - - - o .- _ -s ->- a - --.---.. - j l m m* a y e l dd 3 l d 3, dTT t tn a A l i 3i f3 ' 1* 9 i' 94$i ! l i

             <                     M    i 4        y            b
                                      *l j, lt 39 Q        s    O         '

t b s y C c 4 l 4 4 "i Te

                                                                              ?
                      ' n' 9                                w    !

y 4 SS 1 4f 4

                                             *i R

l ?? R4 4 r$ e 4 i e w

                     ,7 a         1 m    m v,y s                  ss             s                                          /

Y, T . 2 d4dk

                    .o    .

4 d && .. 0 l h

            &Eh$                                                              ,

4 u 1 1 J

                                                                         .                                                                                    y SAFETY LIMITS i

BASES Range Channels will initiate a reactor trip at e current level proportional to , approxirnately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associated I with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System. Overtemperature Delta T , Or-ead ***' U# 98" gg The Overt ature Delta T trip provides core protection to prevent.DNB i or all to

               @p                                                            ations distribu on, provided       of pressure, power, coolant temperature, and axial power that the transient is slow with respect to pi; h:

transit delays from the core to the temperature detectors (about A'teconds),f and bressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections-for axial power distribution, changes in density and heat capacity of water with temperature and dynamic . compensation f I 4eMn Wi pip h- d; hy; ft:: th; :.cr; t th; h;.; t;;p;rctur; d;t;;- below ormal e core safet axial power distribution, this reactor trip limit is always gre er than design,y limit as shown in Figure 2.1-1. If axial _ peaks are as indicated by the difference between top and bottom p er range nuclear detectors, the reactor trip-is automatically reduced i ccording to the notations in Table 2.2-1. , 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. Thr;; h:p Op;r;thn 05:,;; th; i hop P-S ;;tp;bt h p;7;h;ith ofkr 7;;;tth; th; 4 K1, K2 P-S :nd,t K3

                                                                 ;;tp;i         input:,

te it; 3 he;:to V:h; th; Ov;rt;;;;

                                                                                                          . h thh  rah r: Delt: T henn;h :nd r:hing th; j                           h;k :nd trip fun;t hn; ::                                 ::d; cf Op;rcti:n, th: P-0 ht;r-                   ,

f 4 eve 4. High N;;tr n P hn trip t th; r:du;;d pon t i Dwf "c " Overpower Delta T The Overpower Delta T reactor trip provides assurance of fuel integrity, 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 in-density and heat capacity of water with temperature, and dynamic comoesation {' for ip k g d; hy; ft ; th; ;;r; h th: h p t;;;;r:tur; d;t;;ter;. ,,0 ;r:dit j

                                                         =J;g h;n 107 Op;r;thn Of thh trip h th; :: id;nt t
                                                       +,v44,~.4,J m mp.4,Ay L M - A m          ody / isden.

SEQUOYAH - UNIT 1 B 2-4 f

  #4a4 - 4,,A -md   s   4MLu4a4d.-4me 4-    &-A.-A-4    ,4am,-,,e.       da.m,          .m         ma     -  n,me  -es.,  .m   w  _...   ,w-m,- , -     .,--w~    m-, - ----,- ----, -

t Jul.., 2.0 a.ru i

Ovee 4 pua.v /u<e asea / ' Dovrpow.- De//, -7~ ' l

                                                                                                                                                                                           'i
                                                                  .Tisser/       C

f j QJ 4/JW /M de

SW **ye N as Y 0 W-f*W"" ' b r "br.lnd.t j Nfre.ren /s Y /O O $o ATJ" Vdkre ' AJ **9da/wrend bj Ye j? .* l
estd 7 Aop. - 75'is nors,a hers eaa.4 /oop 'r s T frjv 4 4 ocde/ }

opeu d'y e o.,ad><>.a exnf;y sf A -ffma o / /****.rwe.~a~6

                   -M.u .            .

4a,:y M. /,,, ; A n 4 d 4 oyat.<h.f Afu ,sn e . m H ;o,a a., s.umas' - i,, -th suis;,,ra /y.,as. ria'u , afAn.,<<i ,;. 40 /oop s T a,, 4 da da . sana / AAn, : e.g. ineaia<<a/ . 4u /oop .4 ., p.4, M., d-~ / ~4/g,, , M4 o ,d' .r/jt/g sy ,arafaa. pm. .,d:< A tha.db~r L A wa ., pasus,. u a su /o., r, a e -o/ a,,. u s 4 w/N e</a Avf.i rad l:,/ po.sv< reafd;Ja/>a, 4 Asa.-, fusnS* b may ocwe.,- metocy a, .<,,,aH a.Jayat. i., /o90

                   .s p a,A .a T vaku. Acwa4 A &<,i.,aiu., o,# & snap                                                                                                                           t peuA oT ~Ar                                                                                                                                                               ;

r.,co,a/&cora pan-4,/y .doddme.ath,ady., k m,d uJa.,; a.d' J pa .. 4.19s.sy' f.

                   .<&4                  &L                             Oa. , po--aM d48~
                                                                                                                  ,,, /          ,,    w                    J. ,                             ,

xenon oe oNe< /,w,,. sis.,f eo.,att. .,). l\ l l 1 l i

'4

                                                                                                                                         '1, e    Ovupmr /1/h-T -h                                      >$

Y' '**'y"""' *l "**" yprov.dr pro nanV \" 'mi/V< b" b'"$' SAFETY L1 HITS U " WW - 922 6 Aeoube

l' ore despute A Swan .kody-BASES #/**' A'Id'u, 1 -

m!y:::; / ; ren"ir!d by S::::r, thi: p it; fun:ti:n:1 ::p:bility :t the :p;;ified trip :stting i: h ::t;r Pret:;tien Sy:t;;. :ific:ti:n t; :nh;n;; th; ;;;rell reli;bility ;f the t- l

                        .Tnsed "C

t l Pressurizer Pressure The Pressurizer High and Low Pressure trips are provided to limit the pressure range in which reactor operation is permitted. The High Pressure- , trip is backed up by the pressurizer code safety valves for RCS overpressure protection, and.is therefore set: lower than the set pressure for these valves-(2485 psig). The Low Pressure trip provides protection by tripping the ' reactor in the event of a loss of reactor coolant pressure. Pressurizer Water level The Pressurizer High Water Level trip ensures prctection 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 l the accident analyses; however, its functional capability at the specified ! tripthesetting of Reactor is Protection required by this specification to enhance the overall reliability System. .

i l

l Loss of Flow 1 1 The Loss of Flow trips provide core protection to prevent DNB in the event of a loss of one or more reactor coolan s. <

                                                                                             ~ 90 %

Above 11 percent of RATED THERMAL POW

                                                                                            , anJautomatic reactor trip will occur if the flow in any two loops drop b ow Mr of nominal full loop flow.

Abnve25% (P- 0) of n',T:0 T",En"'i P0i':n, tomatic reactor trip will occur if t he flow in any single loop drops below6 of nominal full. loop flow. This-latter trip will prevent the minimum value of the DNBR from going below 1.30 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. L'ith th: Omrt;;perr.tur; 0;1t; i > trip ::t p: int :dj;;t d i: th: clu: :p::ifi:d f;r 0 leep ;;;re.ti;n th; P 0 trip :.t 75% 0,^,TE0 "ER"^1 P0i'ER will pr;;:nt th; mini;;; =1;; ;f th; 000 fr;; g;ing3b:h; with la r 1.20 . during =r=1 ;per:ti:=l' tr:=i:nt , =d :nticip;t;d tr= ,icat: Op;r;tia. l tbs P-8 lde lal, SEQUOYAH - UNIT 1 B 2-5 y ,

SAFETY LIMITS BASES SA 7 Steam Generator Water Level- ~ - - f .Lud "o"

         'h: Ste:: C;n r:t;r h'et;r L; vel Lew Lew trip pie.ide; ;;re pr;t::ti:n by pr:v:ntin; Op:.r:tien with th: :t::: ;;n:r;;;r .;;ter 1;;;l b;1:w th: mini: =                                 j V

1;;; 7;;;ir:d 107 d;;;:t: 5::t 7;;;V:1 ::p::ity. 'h: :p::ifi:d ::t;;'"t preeidee eli; wen: that th:re will b; ;;ffi:i:nt .;;t:r inv:nt:ry '- th: tte 4 j

 ;f;;d

r :.t;r ter: ;y;t::.

t th: ti : Of trip t: :ll w for :t:rting d:1:y; Of th: :=iliary i 1

4t;; /r::d ; tar ric "i;;at;h and Le., Steem Ge,e..to, Weter Le.ei T eam/Feedwater Flow Mismatch in coincidence with a Steam Ge or Low Water le . trip is not used in the transient and accident ses but is= included in Table . -1 to ensure the functional capabilit the specified trip settings and there nhance the overall reliabi of the Reactor Protection System. This tri

redundant to th eam Generator Water Level- 1 Low-Low trip. The Steam /Feedwate ow Mis portion of this trip is 1 activated when the steam flow exceeds eedwater flow by greater than or

' squal- to 1.5 x 106 lbs/ hour. Th eam Gene r Low Water level portion of l the. trip is activated when water level . drops w 24 percent, as indicated by the narrow range in ment. These trip values inc sufficient allowance ! in excess of norm perating values to preclude: spurious-t ' but will initiate a r or trip before the steam generators are dry. Th ore, the 1 requir apacity and starting time requirements.of the auxiliary fee i p r are reduced and the-resulting thermal transient on the~ Reactor Coolan

   -yst;; and eteam generatore ie minimi;ed.

Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips provide ! reactor core protection against DNB as a result of loss of voltage or under- ! frequency to more than one reactor coolant pump. The specified set points assure a reactor trip signal is generated before the . low flow trip set point is reached. Time delays are incorporated in the underfrequency and undervoltage i' trips to prevent spurious reactor trips from momentary electrical power transients. For undervoltage, the-delay is. set so that the time required for a signal to 4 reach the reactor trip breakers fdllowing the simultaneous trip of two or more reactor coolant For underfrequency,the pump busiscircuit delay set sobreakers that theshall time not exceed required for1.2 seconds. a signal to reach BR the reactor trip breakers after the underfrequency trip set point is reached shall not exceed 0.6 seconds. SEQUOYAH - UNIT 1 B 2-6 Revised 03/18/87'

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

( Sov/;on 2, D /Jm

                                                                                                   .Zo~ved 'b "         (hj< /o /.2)

! 1 l SteamGeneratorWaterLevel j 1 The Steam Generator Water Level Low-Low trip protects the reactor from l loss of heat sink in the event of a sustained steany'feedwater flow mismatch resulting from loss of normal feedwater or a feedwater system pipe break, inside or outside of containment. . This function also provides input to the steam generator level control system. IEEE 279 requirements are satisfied i by 2/3 logic for protection function actuation, thus allowing for a. single failure of a channel and still performing.the protection function. Control / protection interaction is addressed-by the use.of the' Median Signal , Selector which prevents a single failure of a channel providing input'to the' control system requiring protection function action. That is, a single failure of a channel. providing input to the control system does not result t in the control system initiating a condition requiring protection function action. The Median Signal Selector performs this by ng1 selecting the channels indicating the highest or lowest steam generator, levels as input to the control system. With the transmitters located inside containment and thus possibly. experiencing adverse environmental conditions (due to a fe'edline break), , the Environmental. Allowance Modifier (EAM) was devised. ..The EAM function: . (Containment Pressure (EAM) with a setpoint of 5 0.5 psig) senses the presence of adverse containment conditions (elevated pressure) and enables the Steam Generator Water Level-- Low-Low trip setpoint- (Adverse)> which _ l reflects the increased transmitter uncertainties due t'o this environment. The EAM allows the use of a lower Steam Generator Water Level - Low-Low i (EAM) trip setpoint when these conditions are not present, thus allowing more margin to trip for normal operating conditions. The Trip Time Delay '(TTD) creates additional operational margin when the-plant needs it most, during early escalation to power, by allowing the. operator time to recover level when the' primary side load is sufficiently small to allow such action. _The TTD is based on continuous monitoring of primary side power through the use of Ac3 /.y AT. Two time delays are- l calculated, based on the number of steam generators indicating less the 1 4 l

t t L ,

                                                       - Sedbn       2. 0 Amr
                                            .Tn.wd           "D '              (Apa 2 ol' 2)                                   ;

L low Low Level trip setpoint and the primary side power level. The magnitude of the _ delays decreases with increasing primary side power level, up to 50 % RTP. Above 50 % RTP there are no time delays for the Low-Low i: level trips. L In the event of failure of a Steam Generator Water Level channel,< it is , placed in the trip condition'as input to the Solid State. Protection System and does not affect either the EAM or TTD setpoint calculations for the remaining operable channels. It is then necessary for the operator to force the use of the shorter TTD time delay by adjustment of the single _ steam generator time' delay calculation (TS ) to match the multiple steam- ' generator time delay calculation (Tg)- for the affected protection set, , through the MMI. Failure of the Containment Pressure (EAM) channel to a ' protection set also does not affect the EAM setpoint. calculations. This results in the requirement that the operator adjust the affected Steam Generator Water Level - Low-Low (EAM) trip setpoints to the same value as <

       'the Steam Generator Water Level - Low Low (Adverse). - Failure of the Ac14.g AT channel input (failure of more than one T H RTD or failure of a                      ,

TC RTD) does not affect the TTD calculation for a protection set. This - results in the requirement that the operator _ adjust the threshold power l 1evel for zero seconds time delay from 50 % RTP to 0 % RTP,; through the MMI. l l

                                   , - ,       .~e---,          ,, ,v.---..>r-        .

TABLE 3.3-1 i y ' o REACTOR TRIP SYSTEM INSTRUMENTATION 8 Y

                '                                                                                                                                 MINIMUM                                             .

TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL-UNIT' 0F CHANNELS TO-TRIP OPERABLE MODES ACTION m e 1. Manual Reactor Trip 2 1 2 1, 2, and

1

2. Power Range, Neutron Flux 4 2 3 1, 2 2

3. Power Range, Neutron Flux 4 2 #

3 1, 2 2 High Positive Rate

4. ' Power Range, Neutron Flux, 4 #

2 3 1, 2 2 High Negative Rate R a

5. Intermediate Range, Neutron Flux 2 1- ~2 1, 2, and

3 T 6. Source Range, Neutron Flux N

A. Startup 2 1 2 2 , and

4 B. Shutdown. 2 0 1 3, 4 and 5 5

7. Overtemperature Delta T Four. Loop Operation 4 2 3- 1. - 2 6, '

                                                                                                                                                                                          .lR45

8. 0verpower Delta T.

Four. Loop Operation 4. 2 3 1, 2 6, l R45

9. -Pressurizer Pressure-Low 4- 2 3- 1, 2 6

10. Pressurizer Pressure--High 4

       ,                                                                                                                                   2         3          - 1, 2               6

11. Pressurizer Water level--High 3 2 2 1, 2 /d

      'EE NY en                                                                                                                                      -
      .zu                                                                                                                                                                          woG -TOM
       ?*

a_. - -a. %---.mm. tama.iaj an-.i.s - m - ----

                                                                          . --. _Lw_-m n--- , - - --m.
                     %.---" " _ . ,  ap. m. **   _;m----wwgs ,-rii.es.
    . v.-._4m

m m TABLE 3.3-1 (Continued) o C o REACTOR TRIP SYSTEM INSTRUMENTATION.

                   .c
                     .                                                                                                         MINIMUM e                                                                TOTAL NO.              CHANNELS             CHANNELS-      APPLICABLE                           '

5 FUNCTIONAL-UNIT OF CHANNELS TO TRIP- OPERABLE- MODES . ACTION a-

12. Loss of Flow - Single Loop 3/ loop 2/ loop in 2/ loop in 1 6 . - _ = _

(Above P-8) any oper- each oper- g pg

                                                                                                     - ating loop              ating loop

13. Loss of Flow - Two Loops 3/ loop 2/ loop in 2/ loop 1 /6 #

(Above P-7 and below P-8) two oper- each oper-ating loops ating loop

14. Main Steam Generator 3/1: p 2/ hsp in 2/1;;; i- 1, 2 . 7 Ee/r7a Water-Level--Low-Low _T,,mf "E " 9 : y ep r- ::cS per-R t! g 1: p ating 1::;

o.IM 1/10 p level 1/10 p 1, 2' # MJJ T 15. Stes:/fct S:ter Fi= 2/ hsp level ~ 7

                                -P.i:::tch and L = Ste =                                   :nd'       ::i cident'              10:01 : d C ncreter ":ter L:::1                             2/h p-f hu                      with       2/ hop-f?
                                                                                   =i:=:t:5 -         1/h:p-f!= = E=:tch er
                                                                                     .   -i._.__,,
                                                                                               .      _1. _ .. .. _. .o. ...,si.._.--,

i__

                                                                                                                                        ... i.

= 1: p :nd 1/h:p-fleu

                                                                                                                              ;;;h; ;tch
                         -16. 'Undervoltage-Reactor Coolant Pumps                                          1/ bus                          2             3                 1        6,    ,

17. Underfrequency-Reactor Coolant-

  ..                             Pumps                                   .       4-1/ bus                         2              3                 1        6, 1

18. -Turbine Trip' g

                                                                                                                                                           - /6 A.              Low Fluid Oil Pressure           3                              .2               2                 1 4            f, B.              Turbine Stop Valve Closure      .4        .                       4.             4                 1        1W6                         y e
                                                                                                                                                .s

--.-. - .i~.-~-; A- -L--- - - - - - - - - -

$_ ,.. ____2__ .L-. - -

a e,A - ar s. 2 +aw;...w 4-as-a-.o +.k- k 1.- t' 4 s s. R m 7 T e yy I o o U. e 4 .; 4% ** i h I . 5- =47 44 2 4 M1 ! i i N; *{ L4 i b4 m ") ,

          ,                                   ,      jf,
                                                      .                    -                            X,                                                       .

m M . 'e

                       -                        E p
         $     9 ew 44 t

l4 ; .? - ; d *t- -Q t dD. i

               ,I R                                   59
                                                      .                                                                  d          et                         ,       a
                *Y                                                        n
         -                                          . p4,             N  n,))4,L b                                                                                                                                                             .l s          j                          <                             e                                         m
          .h                                                            $

t ( a 4 x h 4 'N C- ] st , - w w Nk 4 y

                                !5                                                                                               d a+>                                      a, i

s4 s : r 1 ,r . s, l 9 . R -

                    '-        .t     s        gg h       veTs
                                                                   .g ,g 'Q gi       (j               94L                     nw                                                M        48                                  ,

q 1 4 y e c:i ! R x N Ap l

 .------.-._.-_---.L__L-------.--_-_--__--_-_----.
                                                                                                                           , , -          ,    , ,-     r    ,     e

TABLE 3.3-1 (Continued) TABLE NOTATION With the reactor trip system breakers in the closed position and the control rod drive system capable of rod withdrawal, and fuel in the reactor vessel. The channel (s) associated with the protective functions derived from the out 3 of service Reactor Coolant Loop shall be placed in the tripped condition. The provisions of Specification 3.0.4 are not applicable. High voltage to detector may be de-energized above the P-6 (Block of Source Range Reactor Trip) setpoint. ACTION STATEMENTS ACTION 1 - With the number of channels OPERABLE one less than required by the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours or be in HOT STANDBY within the next 6 hours and/or open the reactor trip breakers. ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and POWER OPERATION may proceed provided the following conditions are satisfied:

                                         .                                     a.         The inoperable channel is placed in the tripped condition within 6 hours.

lR51 The Minimum Channels OPERABLE requirement is met; however, 14e iaepauli/e en: Odditi:n:1 channel may be bypassed for up to 4 hours lR51 for surveillance testinh1pku o bunabr S ecification 4.3.1.1.1.

c. Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL and the Power Range, Neutron Flux high trip reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours.

~

d. The QUADRANT POWER TILT RATIO, as indicated by the remaining three detectors is verified consistent with the normalized symmetric power distribution obtained by using the movable incore detectors in the four pairs of symmetric thimble locations at least once per 12 hours when THERMAL POWER is greater than 75% of RATED THERMAL POWER, t

September 17, 1986 : SEQUOYAH - UNIT 1 3/4 3-5 Amendment No. 47 I

i TABLE 3.3-1 (Continued)- 'I ACTION 3 - With the number of channels- OPERABLE one less than required by the Minimum Channels:0PERABLE requirement and with the THERMAL POWER level,

a. Below the P-6-(Block of Source Range Reactor Trip) setpoint,

' vestore the inoperable channel to OPERABLE _ status prior to - ' increasing THERMAL POWER above the P-6 Setpoint.

b. Above the P-6 (Block of Source Range Reactor Trip) setpoint, but below 5% of RATED THERMAL POWER, restore the. inoperable channel to OPERABLE status prior to increasing THERMAL POWER-above 5% of RATED THERMAL POWER..

c. Above 5% of. RATED THERMAL POWER, POWER.0PERATION may ' continue.

d. Above 10% of RATED THERMAL POWER, the provisions of-Specification 3.0.3 are not applicable.

ACTION 4 - With the number of channels OPERABLE one'less than required by the Minimum Channels OPERABLE-requirement and with the THERMAL; POWER level: a. Below the P-6 '(Block of Source Range Reactor Trip)'setpoint, restore the inoperable channel _ to OPERABLE status prior to' increasing THERMAL POWER above the-P-6 Setpoint. b. Above the P-6 (Block of Source Range Reacto'r Trip)'setpoint, operation may continue. ACTION 5 - With the number of channels OPERABLE one less than required by-the Minimum Channels OPERABLE requirement, verify compliance-with the-SHUTDOWN MARGIN _ requirements of. Specification 3.1.1.1 or 3.1.1.2, as applicable, within 1 hour and at least.once per. 12 hours thereafter-ACTION 6 - i With the number of OPERABLE channels one less than the Total L Number of Channels,: STARTUP and/or POWER OPERATION 'may proceed provided the following conditions are satisfied:

'=

a. l The' inoperable channel is-placed in the. tripped condition 1 within 6 hours.

      - =-                                                                                                    . l R!

4 %**'I4 The Minimum Channels- OPERABLE requirement is met; however, l

n: :dditi:n:T channel may be bypassed for up to 4 hours for surveillance testin pjrSgijication4.3.1.1.1' } R$1 l; ACTION 7 -
"ith the - ber of OPEPA*LE _htnnel: en: 100: then the Tet:1 ,

L p 'u! :ber cf Char"e!:, ST^"TL'P ':nd/or P0MEP OPEP^TIO" ::; ;ir::::d

l nt9 perfer eace of the acxt re';uir:d C"^"'!EL PJ"CTIO"AL TZT

                                -previded the %eper:ble chaaa:! it ple :d 4- the tripp d ::nditi:n l                              --within 6 h ur .                                                                        i 1

l R5 l > j I September 17. 1986. SEQUOYAH - UNIT 1 3/4 3-6 Amendment No. 47

7 1 8 t w TABLE 3.3-1 (Continued)-

ACTION 8 - With less than the Minimum Number of Channels OPERABLE, declare.

the interlock inoperable.and verify that all affected channels of the functions listed below are OPERABLE or. apply the appro- O priate ACTION statement (s) for:these functions. Functions to be evaluated are: l

a. Source Range Reactor Trip

b. Reactor Trip Low Reactor Coolant Loop Flow (2' loop.').

Undervoltage. ! Underfrequency. Pressurizer Low Pressure-Pressurizer High Level '5

c. Reactor Trip t Low Reactor Coolant Loop Flow (l'1oop) ,

                     .d. Reactor Trip
                                     . Intermediate Range ~                                                     -

Low Pow'er Range Source Range

e. Reactor Trip .

Turbine Trip - ACTION 9 - 0:1:t:0 ~S ~- =- 0:ict:d

ACTION 10 - { fo.wd 'N" ACTION 11 - C 1:t:d ,). ACTION 12 - With the number of channels OPERABLE one less-than required by the Minimum Channels OPERABLE requirement, be in at least HOT: STANDBY within 6 hours; however, one channel may-be bypas' sed 1for up-to 2 hours for surveillance testing per Specification 4 3.1.1.1 provided the other channel is OPERABLE. R58 I H l i l

~1

l l-March 16, 1987.

l 1' SEQUOYAH - UNIT 1 . 3/4 3-7 . Amendment No. 54

                                 -w-             a  4       -   -      --<s--     x----.--__----   ..--.c     _     -- - . - - - - . - - - _ _ . - - -

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

i Tued "F Tille 3.3~/. L l-ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are-satisfied:

a. . The inoperable channel.is'placed in the tripped.

condition within 6 hourJ.

b. For the affected protection set','the Trip Time Delay

for one affected steam generator (Ts).is adjusted to match the Trip Time Delay'for multiple affected steam generators (Tg) .within 't hours.

c. The Minimum Channels' OPERABLE requirement is met; however, the. inoperable channel may be bypassed for up to '/ hours for surveillance testing of o/4,.

             ,                          channels per Specification 4.3.1.1.1.

ACTION 10 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION.mayL proceed provided that within' 6 hourg for the affected. protection set, the Trip Time Delays (Ts'and Tg) threshold power level for. zero seconds time' delay'is adjusted to 0 % RTP. l ACTION 11 - With the number of OPERABLE channels'one less than the' Total Number of Channels,. STARTUP and/or POWER OPERATION may proceed provided that within 6 hours, for the affected

l

' protection set, the Steam Generator Water Level - Low Low (EAM) channels trip setpoint is adjusted to the same value as Steam Generator Water Level - Low Low ~(Adverse). i i

                                      -      ,-<._...-_c.

TABLE 3.3-2 REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES , SE 7 Fur 4CTIONAL UNIT RESPONSE TIME . E 1. Manual Reactor Trip NOT APPLICABLE s 2. Power Range, Neutron Flux $ 0.5 seconds *

3. Power Range, Neutron Flux, High Positive Rate NOT APPLICABLE

4. Power Range, Neutron Flux, High Negative Rate 5 0.5 seconds *

5. Intermediate Range, Neutron Flux NOT APPLICABLE-

{ 6. Source Range,' Neutron Flux NOT APPLICABLE B70/Ey/, s.o . { 7. Overtemperature Delta T 5 +-9 seconds *

8. Overpower Delta T .

                                                                                                   "0T ,"#PLIC"0LC d 8,O e f
              ~9.      Pressurizer Pressure--Low                                                   $ 2.0 seconds

10. Pressurizer Pressure--High.- $ 2.0 seconds-

              -11. Pressurizer blater Level--High                                             'NOT APPLICABLE

12. Loss of Flow : Single Loop (Above P-8)- 5 1.0 seconds s

Neutron detectors are exempt from response time. testing. Response' time of the neutron flux signal

r. portion of the channel shall-be measured from detector output or input of first electronic component.in. channel. -

l l I --

                                                                    -s_.   . , ._                   . ~ _ . .  ~.   . - - . . .,m    . ~ . . _     . _.n

                                                                                                                       .                                                   m m                                                                                TABLE 3.3-2 (Continued)

E g REACTOR TRIP SYSTEM INSTRUMENTATION RESPONSE TIMES 9z i FUNCTIONAL UNIT RESPONSE TIME E q 13. Loss of Flew Two Loops

    -                      (Above P-7 and below P-8)                                                            $ 1.0 seconds

14. Main Steam Generator Water Level-- gAmh Low-Low Inse f "6 ~-9 3 2.0 seccr.d; A44./

15. Stcom/recdwotcr risw Mi;;ctch cod .

Lcu Stcc= C;ncretor t'otcr Lc.c? "

                                                                                                                 .0T A""LICAOLE   ##7JJ
                 - 16. Undervoltage-Reactor Coolant Pumps                                                   5 1.2 seconds

17. Underfrequency-Reactor Coolant Pumps 1 0.6 seconds e.. .

 ~D               18. Turbine Trip
 .w 0                       A.       . Low Fluid Oil Pressure                                                    NOT APPLICABLE B.      ' Turbine Stop Valve                                                         NOT APPLICABLE

19. Safety Injection Input from ESF NOT APPLICABLE

         . g 20.          Reactor Trip Breakers                                                                NOT APPLICABLE
  .W       p.
 .[         C' 21.        Automatic Trip Logic                                                                NOT APPLICABLE u.to                                                                                                                           R16 S Ut 22.              Reactor Trip System Interlocks-                                                     NOT APPLICABLE.
    ?5
  . . - Co
        .IO w

fa O ., Inc & 4 -r 4 6 a. ,.s 6 1 , Eoy4 ~ vp<aus p la A, <aks4,'

                                       .soh/.14&yyLA., aD.,ds, W 4L k ,/ca a,.. 74;, .

n//;J -A aspo,ua -/;,,,e >1ewa.y A, 77/EgnML MWEd in nws of so2 ATA

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

I l .

                                  .Tn.nd "G "   ta A le 3 , 3 - 2, .                                                   !

W [ fuu k., / Wn,f Reuema %. l N HM J4., fanar.4< it4L , bul -- hw-low 0 ( 1 A. A'a loop s T 4 & O sud (1) 1

(P450% erP; /> Sox R7P.) a .s % , A .,. 4 m 4 , . .s .zo n t ) A CYe / = = gw* ow 64rer . j EAn1) C. Coda.hn.,f P e << 4.;t.o.raca&b' . , (EAm) b O 1 L I V V

j

TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS a E CHANNEL MODES IN WHICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE E FUNCTIONAL UNIT CllECK CALIBRATION TEST REQUIRED

1. Manual Reactor Trip N.A. N.A. S/U(1) and R(9) 1, 2, and * -lR58

2. Power ~ Range, Neutron Flux S D(2), M(3) Q 1, 2 and Q(6)

3. . Power Range, Neutron Flux, N.A. R(6)- 1, 2

                                                                                                        .Q High Positive Rate

4. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Negative Rate-w 5. Intermediate Range, S R(6) S/U(1) 1, 2, and

1 Neutron Flux T 6. Source Range, Neutron Flux S(7) R(6) M and S/U(1) 2, 3, 4, 5, and
R31

7. Overtemperature Delta T S R )(Q " T 1, 2

8. Overpower Delta T S R .M'Q 1, 2

9. Pressurizer. Pressure--Low S R Q 1, 2 10.- Pressurizer Pressure--High S- R Q 1, 2

           - 11. Pressurizer Water Level--High              5                         R               Q                   1, 2-

12. Loss of Flow - Single Loop-- S R Q 1

. 13. Loss of Flow'- Two Loops S' R N.A. 1 if k g 14. Main St am Generator Water. S 9 Q '1, 2 R58

r R Level .nw-Low _ ,

A. L &a L u n te /- . .s At' R I,2. l LA"' * _ Law. Low W asa)

 . $ .I          B. SLwa h ara     "'S l**l-~                                                 *
  ~,,

Low. Low (EAm) .S A G h2-

c. RC.s b y o r S R' O 42.

D. Co.,fi.,-a 1 %- (EAm) 3 A O hA

                                                                                 ,          .                        ,             s                 - . , , , ,

l i-TABLE 4.3-1 (Continued) g; REACTOR TRIP SYSTEM INSlRUMENTATION SURVEILLANCE REQUIREMENTS

      @                                                                                                      CHANNEL               MODES IN WHICil Q

CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE

FUNCTIONAL UNIT CHECK CALIBRATION TEST Dde4<of REQUIRED i

15. Steam /r cdwater Flow tiismatch or,d S 1, 2 iM fJ Q -

g Lc.. Stcc.. Ccr.cratcr k'ater Lcvci U 16. Undervoltage - Reactor Coolant N.A. R ,M'Q -1

      -          Pumps

17. Underfrequency - Reactor' Coolant N.A. R JVQ I Pumps

18. Turbine Trip was. Tws

                                                                                                                                -e A.                      Low Fluid Oil Pressure              N.A.               N.A.           S/U(1)                     1 B.                     Turbine Stop Valve Closure           N.A.               N.A.           S/U(1)-                    1

19. Safety Injection Input from ESF N.A. N.A. -M(4).R 1, 2 g 20. Reactor Trip Breaker N.A. N.A. M(5) and S/U(1) 1, 2, and *'

[ 21. Automatic Trip Logic N.A. N.A. M(5) 1, 2, and

O

      "   22. Reactor Trip System Interlocks A.                     Intermediate Range-                   N.A.               R             -G/U(0) 4                   2, and

Neutron Flux, P-6

8. Power Range Neutrori N.A. -R- A'. A S/U(0) N A. 1 Flux, P .

C. Power Range Neutron- N. A.' R S/U(e) 4 1 Flux, P-8 . . D .- Power Range Neutron N.A. R -S/U(9) 4 1, 2 , Flux, P-10 a E. Turbine Impulse Chamber N.A. R S/U(0) 4 1 Pressure, P-13 g@F F. ~ Power Range Neutron g g8- Flux, P-9 it. A. R 4/U(0) 4 1 18 G. Reactor Trip, P-4 N.A. N.A. 1/U(S) R 1, 2, and

u* l R67

23 Reactor Trip Bypass Breaker N.A. N.A. M(10)R(11) 1, 2, and

R58

[ J e _ _ _ _ _ _ _ - _ - _ _ - _ _

                                                                        ,                     f          v      , s4~- .( w      ,      y        ,u. 4,,,,  e-  -

yv, -,-yo-4

TABLE 4.3-1 (Continued)- NOTATION With the reactor trip system breakers closed and the control rod drive system capable of rod withdrawal. 11 (1) - If not performed in previous 4 days. (2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjustchannel if absolute difference greater thn 2 percent. (3) - Compare incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED R118 THERHAL POWER. Recalibrate if the' absolute difference-greater than or equal to 3 percent, alsM (4) - M:nu! ESF function:! input ch :k'r; ry 28 : r.th , (5) - Each train or logic channel shall be tested at least every 62 days R58 ! on a STAGGERED TEST BASIS. The test shall independently verify the- ! OPERABILITY of the undervoltage and automatic shunt trip circuits. R118 - (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION. I i (7) - Below P-6 (Block of Source Range Reactor Trip) setpoint. 4 (8) -

          -syster Leg'44/Orly,      e2:b t:rtup er her required "ith the r:::ter trip breakers ! : d end th ::r.tre! 7:d driv Cy t= ::p;bk ;f                       ;
          -red "ithdraf:1 " not perf rmed      4, pr ;icu: 02 d y;.

i (9) - The CHANNEL FUNCTIONAL TEST shall independently verify the operability of the undervoltage and' shunt trip circuits:for the manual reactor lR1 trip function. (10) - Local manual shunt trip prior to placing breaker in service. Each train shall be tested at_least every 62 days on a STAGGERED TEST BASIS, Rll8 l 1 (11) - Automatic and manual undervoltage trip. ! R58 i i SEQUOYAH - UNIT 1 3/4 3-13 Amendment No. 54, 114 May 5, 1989

TABLE 3.3-3 m. E C ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION O

               '                                                                                               MINIMUM TOTAL NO.        CHANNELS                 CHANNELS                          APPLICABLE E

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

             ~           1.      SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION

a. Manual Initiation ~2 1 2 1, 2, 3,.4 20

b. Automatic Actuation 2 1 2 1,2,3,4 15 Logic

c. ' Containment' 3 2 2 1,2,3 36*/7 #

w s Pressure-High " ' " n : ;_^ w d. Pressurizer 3 2 2 1, 2, 3# 16* 17

Pressure - Low leW

e. D@. f fcrcntial ' 1, 2, 3 Prc;;urc S:troen-Sica;- Linc; 'iigh N'# .Ecur L cp: 3/stca: line 2/stca: linc 2/stca; linc 15*

Operating any stca; iinc R45 f Stect Flcw ir Tuc ## 1, 2, 3 u@ a =1 Stcom Linc;-tiigh

         *C O EE                         Four. Loops                     2/stcoa line     1/steae line             1/ steam lina                                      IW
         -hi*

Opcrating any 2 ;tec:

                                                                                     -?:nc; Yf                                                                                                                                                                             R45 Gt.

2-U e

TABLE 3.3-3 (Continued) m jo ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM INSTRUMENTATION 2,

=

MINIMUM , TOTAL NO. CHANNELS CHANNELS APPLICABLE E FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE ~ MODES ACTION Z i

       -   COINCIDENT WITH EITHER
                        ' ,g--tcu-Leu                                                                            1, 2,       3,,

T Yew A0 [SS_dSSS3 ITavg! 00E avg0 "7 avg C07

                       "''"""                                       Iccp:                3.!cep R45 OR, COINCIDENT WITH                                           .q/sf    A., I,,
                                                                                             .2/,4,, h,,.                        tt

f. Steam Line Pressure-Low 3/Jf'*-' l' ,,y, f/ u Ae 1, 2, 3 17* wog ye;

{ Four Lcep: 1 prc :urc/ 2 prc: urc; I prc;;urc .W Operctin;; !ccp cny locp: cny 3'!cep: R45' 5 2. CONTAINMENT SPRAY

a. Manual 2 1* -2 1,2,3,4' 20 l R67

b. Automatic Actuation 2. . 1 2- 1,2,3,4 15 Logic

c. Containment Pressure-- 4 2 3 1, 2, 3 . 18

                      .High-High                                                                                                                                          i

e I> E 3. CONTAINMENT ISOLATION o ai .

s :) ES a. Phase "A" Isolation j

~~

1). Manual .2- 1 2 1,2,3,4 20

   ~z i'     P                2)-    From Safety' Injection ~     2-                 1                    2L           I1,2,'3,'4                   15 LG xs                        Automatic Actuation
   ~,~                        Logic

[. I E *Two switches must be operated simultaneously for actuation. R67

                                                                 .                        #-              ~ . .    ,   -c .-       ..E,,,-.y..                 ,       .-

m TABLE 3.3-3 (Continued) m o C ENGINEERED SAFETY FEATURE' ACTUATION SYSTEM' INSTRUMENTATION-O z

           ,                                                                                                        MINIMUM TOTAL NO.            CHANNELS            CHANNELS              APPLICABLE E

FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION

         -       4. STEAM LINE ISOLATION

a. Manual 1/ steam line 1/ steam line 1/ operating 1, 2, 3 - 25 steam line

b. A :omatic -- 2 1 2 . 1,2,3 23-Aci. 3 tion Logic

c. Containment Pressure--- 4 2 3 1,2,3- 18 Hi lh-High w

2 d. Steer cler 4- 7"^ ' ' w Stes: Line - "igh H Four Lccp; 2/:tes: ?in 1/:t" : 1in 16* Operating any'2 tec: . lI/tec=1in R45 SL 0 ECIt: CIDE lT WIT:l EITl:CR T _ _ f m. ._ E mo 1 0 1

                                 ' a vg c _.'.:", _"_",                  ,y         ,,___     ,y                   ,1          _ _ . .
                                                                                                                                         ^* ' ' ~

se, n_"_'" !? " ' ' a v g' ' """ ;_ lava ""'_ . . .Ci.vve-layo ""' ^" mys....3 .vvy R45 nn f'n T Elt* T nE* kf T 18 7 7t1 m>

   .g . g
                ). NAa 2 iSe"Phessure--

to, '3/slw he #/# '" * ' " e.y .r/c. /<*e 2/'I"'"I*' 1, 2, 3 17* ER u,ros m '. -[ a .F;ur L:cp: 1 prc: uft/ 2 pre: ure: '1 pre::ur: 15* =- -- R Operating iccp any_ loop ar.y 3 1;;p:

  ~."=
                                                                                                                                                  "                   R45 3               G.ky Wbf wmhoes bewere                             3l1hmbne                   J'    Mbe ln       2 fat ==sb e            -]      ll a~                   .gs             yy                                                  s y </~ A-e
                                                                                                     .-                    ,--         y               , , , , ,   ,       .y

                    ,,                                                                                 TABLE 3.3-3 (Continued)

EO . Ej ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION-5

MINIMUM .. SE TOTAL NO. CllANNELS CilANNELS APPLICABLE

                    ^j    FUtlCTIOilAL UNIT'                                          0F CilANNELS                  TO TRIP                 OPERABLE            MODES        ACTION

5. TURBINE TRIP &-

FEEDWATER ISOLATION

a. Steam Generator 3/ loop 2/ loop in 2/ loop in #'

1, 2, 3 JWrf/7 (hk,oj78ff) I Water Level-- any oper- each oper-High-High ating loop ating loop

b. Automatic Actuation 2 1 2 1,2,3 23 R67

                   ,,                      Logic D      6.
                   ,,           AUXILIARY FEEDWATER I

U$ a. Manual Initiation 2 1 2 1,2,3 24

b. Automatic Actuation 2 1 2 1,2,3 23 Logic-

c. 'Hain 5tm. Gen. Water Level-Low-Low

                                          ..    .es
                                                      .. .~u. a. .
                                                                                                      .2huerl "If' I

o{[ ,g,jp7, Driven Pu=p; 3/st=. gen. 2,'st . g-.. 2/st=. gen. 1, 2, 3 .1S

gg

                                                                                                                   ...,...c.

es

               'Z'                       ii. . Start'Turbinc-                            ..                             .

4 O[. Dr ven o::p i 3/st= .gcn. 2/sta. gca.

2'sta. gcn 1, 2, 3 15*

             ~ ~ - pr                                                                                              any 2 :t . gen G

Sib

                 ~
                              'd.        S.I.

Start flotor-Driven

                                                                                                                                     ~
                 $                    1 Pumps ar.d~ Turbine
                                       ' Driven l Pump                                See 1 above'(all S.'I. initiating functions and requirements)

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

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

as #6 - - A- - ,% 4 - ,.m- a..m. a 4 a w- r-l

                                                                                                           'l l

l l e , Q? g N. N W Y $ $ S L

             <i                   n               e             d         g 4y                    ss              +             ss       #                               :

r

                                  ..~             .e'                                                    t
     ' s *l .

o N

         % 44 411                     4hr v i              i 4 3 L

i m L9 m wqq 4 Dli 4e NI

       'P H                                           5
                                 .c'                                                                     i a                                             .

b-w t i

                                * *0           4=9g          .

{s* d ' w! E sR =% h d d' vg g

      =
                                       %d 45       N D\g*

et s n -

l % 9 i i m. 4 4 x4 4 14 L - - 1 s \) C 9 - I

            *T                  n              n               w        w                                   l id eser uw,
                               '4 n8, i               j s           :irj                    - o                       .                               .

9 d ~!iits3 L7 dh J ny #

                                                                         .;*~s                             1
              ,             .e                 a                a       ,                                    1 l              N       d                                                                                     !

l

a -.a ..,a. u. u.so..a.-<a , -sc -.---- 4*"-' ' E D

             +                   w 4                      m               =

y n h n- l F 4j # ev' cf et

        @y                       +         +                      +               o                             :

9 19 .= r% ! ev. I . 5 .4

   ,,   4' <

N 4.t f t bMN y ,$ h9 d( [0i t O N nl li M M ,, ?

               .                  s         .

N d etj ,f "i g y D - Pi 5n ,h4n. e s .s e e d,g= d I d d

        %n
                           +

N, > w 3 l 1 s < 4 e I

                   ,   -    k' y 3 3     3 f                                ,
                                                                               *x                                  i ddai 4it               49                      h 1             5 1't      .e. >e                                 t q
           ,)

jjid]$ . s OS S N.N

l XI TABLE 3.3-3 (Continued) l iE ' ENGINEERED SAFETY' FEATURE ACTUATION SYSTEM INSTRUMENTATION SE EE MINIMUM c: TOTAL NO. CHANNELS CllANNELS APPLICABLE {} FUNCTIONAt UNIT OF CilANNELS TO TRIP OPERABLE MODES ACTION

                   **7. LOSS OF POWER

a. 6.9 kv Shutdown Board

                                    --Loss of Voltage

1. Start Diesel 2/ shutdown 1 loss of 2/ shutdown 1, 2, 3, 4 20*

Generators board voltage on board any shutdown us board N

2. Load Shedding 2/ shutdown 1/ shutdown 2/ shutdown. 1,2,3,4 20*

i' board board board-C' b. 6.9 kv Shutdown Board Degraded Voltage

1. Voltage Sensors 3/shutdownL 2/ shutdown 2/ shutdown 1, 2, 3, 4 20*

board board board-

2. Diesel Generator '2/ shutdown '1/ shutdown 1/ shutdown 1,2,3,4 20*

Start and Load board board- board Shedding Timer

3. SI/ Degraded 2/ shutdown 1/ shutdown 1/ shutdown 1, 2, 3, 4 -20*

    ,si                               Voltage Enable        board          -board-              -board s@7                               Timer Bit ag
                 " "n ' t     " is tec! ic2! pediricatice is te Sc J:p?c=catce              ring the :tcrtup rc: w ing inc c e rerse v ~;

e e: tage e- fe? ?ct.ing cc;p?ction of the ,nudification, whichever is earlicr.

   . .?

cs O e . O

TABLE 3.3-3 (Continued) M jg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION S E

                    ,                                                                                              MINIMUM
                                                                                                                                                                 ~

c: TOTAL NO. CHANNELS CHANNELS APPLICABLE g5 FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE . MODES ACTION w sa 8. ENGINEERED SAFETY FEATURE

                                  -ACTUATION SYSTEM INTERLOCKS

a. Pressurizer Pressure - 3 2 ? 1,.2, 3 22a i Not P-11 New S'b a JJ .

b. T I' E' 3 avg - P-I2 4 2 3 22h

c. Steam Generator 3/ loop 2/ loop 3/ loop 1, 2 22c os Level P-14' any-loop s .

u) 9. AUTOMATIC SWITCHOVER TO CONTAINi ENT SUMP

a. RWST Level - Low 4 2 3 1,2,3,4 18 R67 COINCIDENT'WITH Containment Sump Level - High '4 2 3 1,2,3,4 18 AND

                                      . Safety Injection-              L(See1aboveforSafetyInjectionRequirements)

b. ' Automatic Actuation 2L 1- 2 1, 2, 3, 4 15 R67 g? jf Logic 2 E'

8 ' ??

i. .....:. _ _ _ _ _ . _

                                         . , _ .            - . . .  .- - i._ ' , . s ; -
                                                                                .,        .. c           . . - -           1       . . , . _      _ , , ._.       _ , .; .

Q ~$nd obmeds/h Jkle./ ohn A// dd may be llon.la./ 4*Aw

                  */]plo, . M /y .27-j< d e., n      fin he A< awe ~hw b no/ J/~M TABLE 3.3-3 (Continued)

SO Y TABLE NOTATION N Trip function may be bypassed in this MODE below P-11 (Pressurizer Pressure i ggBlock of Safety Injection) setpoint. FP g-+u.yTrip 'uncthn ::y be byp:;;;d in thh "005 5:hw P-12 (Tavg Shek of hfety n_. __s __.__,_.

                     .,vu,  ,,,.rv.....

The channel (s) associated with the protective functions derived from the out of service Reactor Coolant Loop shall be placed in the tripped mode. The provisions of Specification 3.0.4 are not applicable. ACTION STATEMENTS ACTION 15 - With the number of OPERABLE Channels one less than the Total I Number of Channels, be in at least HOT STANDBY within#" ours 3 and in COLD SHUTDOWN-within the following 30 hours; however, i one channel may be bypassed for up to " hours for surveillance R67 FP testing per Specification 4.3.2.1.1 p o ided the other channel is OPERABLE. ACTION 16 - A4k/ ! Uith the aueber of OPEP^nLE Ch:rneh en: h;; th n th: TetU

                             " ember of Ch:nne h , Oper:t h n ::y pre:acd until perfer;;n:; cf gopoi,                      the aert -eauired CP^""E! rUMCTIO"^'. TEET, pr vitd the in :per:th channel 4: puced 4n the tripped cendit99 <fggr 3 heur ACTION 17 -         uith a chenael errecitted "ith an eperating Sep 4neperebh ,                     -
                             -estere the ia^aer=h'e ek-aae' t O o r o ^ D '_ r :t:t"- "'th4- 2                    i y*4",(heursa* be 4a et hest MOT SHUT 00Y" "ith!- the fol he ng 12         i heurs; heueve , one cbrane! erectieted "!th~en eper: ting Mep                        l FP                       ==y be bypeseed fe- up te 2 heur: fer curvei'un : te: ting per gp eg$<$;3gjg7 a,3,3,3     3,
                                                                                                                .i ACTION 18 -                                                                                           ;!

With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable ! channelisplacedinthebypassedconditiong"dtheMinimum Channels OPERABLE requirement is tron:tr:t._ ithi ' heur; one additional channel >may be bypassed for up to X hours for^ surveillance testing per Specification 4.3.2.1.1. Y ACTION 19 - With less than the Minimum Channels OPERABLE, operation may - continue provided the containment ventilation isolation valves are maintained closed. q ACTION 20 - With the number of OPERABLE Channels one less than the Total. , Number of Channels, restore the inoperable channel to OPERABLE -I status within 48 hours or be in at least HOT STANDBY within the l next 6 hours and in COLD SHUTDOWN within the following 30 hours. I i a SEQUOYAH - UNIT 1 3/4 3-22 Amendment No. 63 1 December 31, 1987 ) l l

In.ree f "['

               .                  Dih 3.3-3 ACTIOV /7 - With the number of OPERABLE Channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within d hourJ.

b. The Minimum Channels OPERABLE requirements is met; however, f4e /adpM/a channel may be bypassed for up to V hours for surveillance testinggper Specification 4.3.2.1.1.

                                                           & oNer s Aa .,,ls e

e

TABLE 3.3-3 (Continued) ACTION 21 - With less than the Minimum Number of' Channels OPERABLE, declare ; the associated auxiliary feedwater pump inoperable, and comply R133 with the ACTION requirements of Specification 3.7.1.2. ACTION 22 - With less than the Minimum Number of Channels OPERABLE, declare the interlock inoperable and verify that all affected channels of R16 the functions listed below are OPERABLE or apply the appropriate ACTION statement (s) for those functions. Functions to be evaluated are:

a. Safety Injection Pressurizer Pressure ',"{ g 7,*,

                                                                                . j,g 'l" L y

b. Se'ety hject'^r u<ms e . . .. i w e s ..

                         -StN unos I L M5'I:5i5ti [                Mtv J/8 e . . . .. i. w. . r , m ,

i

c. Turbine Trip Steam Generator Level High-High .

Feedwater Isolation l i Steam Generator Level High-High

ACTION 23 - 1 With the number of OPERAbd channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours 4 and in at least HOT SHUTDOWN within the following 6 hours; however, one channel may be bypassed for up to 2 hours R67 . for surveillance testing per Specification 4.3.2.1. ACTION 24 - With the number of OPERABLE channels one less than the Total ) Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY : within 6 hours and in at least HOT. SHUTDOWN within the following 6 hours.

ACTION 25 -

With the number of OPERABLE channels one less than the Total , Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve inoperable R16 ' and take the ACTION required by Specification 3.7.1.5. , AcTzov 3t , Acrrow si Im,, ) "3 " 6AMl770 Acrtov at ) SEQUOYAH - UNIT 1 3/4 3-23 Amendment No. 12, 63, 129 November 28, 1989 s

Tasic 13*3 INSERT "Y ACTION 36 - With the n' umber of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:

a. The inoperable channel is placed in the tripped condition within 4 hours.

1 j

b. For the affected protection set, the Trip Time Delay for one affected steam generator (T S

                                                                               ) is adjusted to match the Trip Time Delay for multiple affected steam generators (Tg) within 4 hours.

i

c. The Minimum Channels OPERABLE requirement is mett however, the inoperable channel may be bypassed for up to 9 hours for surveillance testing of one channels per Specification 4.3.1.~1.1.

ACTION 37 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 4 hours, for the affected-protection set, the Trip Time Delays (Ts and Tg) threshold power level for zero seconds time delay is adjusted to 0 % RTP. , ACTION 38 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 4 hour 4 for the affected protection set, the Steam Generator Water Level Low Low (EAM) channels trip'setpoint is adjusted to the same value as Steam Generator Water Level - Low Low (Adverse). e

m m TABLE 3.3-4 o g ENGIfiEERED SAFETY FEAIURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS

FUTICT10f1AL UllIT TRIP SETPOIrli ALLOWABLE VALUES C . E

        ~             I. SAFETY IllJECTION, TURBINE TRIP Af1D
       ~                       FEEDWATER ISOLATION

a. flanual Initiation Not Applicable Not Applicable

b. Automatic Actust. ion Logic flot Applicable Not Applicable 16

c. Containment Pressure--High 5 1.54 psig 5 4-7 psig gg/. Il d.

usg.g = _-s Pressurizer Pressure--Low > 1870 psig > M psig

      .{                    e. Differ:nti;! ' c;;;rc 1 ?^^ p;f                                                               1 ? 1 p- i y                               NTeVe:dr Sic = Lincr!!!g',

2 f. t;;; r ? = i- T=- S tc= L i t ,-- ' ^ f = tion & fincd a; : ^ fu,etiers i fi.cd a3 u:.a. c.. : __ : 2_ _. . .r.t , . _. ._, _ . _ , , _ . _ _ . ._ __ _ _ _____> _ , , _ _ _ _ . ._ _

                                       'J    *V   "%              " ' ' ' '
                                                                                      'gy              Lvw          avus-w .i s .       n up but c a yvi su                                  avssvwa.
                                                                                                                                                                                                                                                       .___.___4__.

n 4.ap s u s a s a pe v i vu s a nj c._..,_.. -i: . .. o _ ,. , . . . _ _ b ' uw ~;

                                                                                                                    ,,_t_...___10 '" om__       f f u ? ' _,L                             10
                                                                                                                                                                   ,, i c=

vs . . ._- - .. e svw un b vrs s i e vv w a ru L viv L_.. "_T Of _'_??Otc =,_'__? __ m _ , , , __ = , , mis t e s s u s vv un ru Lvv jL6 ! cad ,d th;7 37 t^,;; a op incr ;3 fag avuu v i vw M nt- ;;ing ' ncarly i; ; linc;rly to ; op cai.u;;-,d op corrc;p;nding to ? ?'." ing to 111.5% of fu!! sics; ce..,f ', u', .,_' _ sic;= f ! = ;t f?;u at fu?! ? d aw1 x vsa v, .

                                                                                                                                   ,,nor t,

v . r - e,nor avg u avg m ssv e S A S~'c Amm -- Zw

                                                                                                                   > 600 psi pressure (gsteamline  N./,)/
                                                                                                                                                                                          > 404 psi steam In,ne I                                                                                                                                                                                      pressure Ne4 /)

9 b e+ s T = _ . _ _ e_. -e-- *' +. c __ -=s.cw m- *-e+* r e++ w+'N- #- w_ ,_____m__. .2_ _ _- ___._n _.n_1_arus.,m_ um-_ _ _ , _ _ . _ _ _ . . . __s____ ___. . . _ _ . _ ___m __s_ _.

TABLE 3.3-4 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRittENTATION TRIP SETPOINTS a f FUf4CTIOilAL UNIT TRIP SE1 POINT, All0WABLE VALUES E

       ~
                - 2.         CONTAINMENT SPRAY
       --e
       ~                     a.       Manual Initiation                                     Not Applicable                              Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable 3.9

c. Containment Pressure--High-High 5 2.81 psig i +:-97 psig

3. C0f4TAINNENT ISOLATION

a. Phase "A" Isolation -

e,, 6j/, 21 1 1. Manual Not Applicable Not Applicable

2. From Safety Injection Not Applicable Not Applicable Automatic Actuation logic

b. Phase "B" Isolation

1. Manual Not Applicable Not Applicable

2. Automatic Actuation Logic Not Applicable Not Applicable

                                                                                                                                            .'t. 9

3. Containment Pressure--liigh-High -< 2.81 psig -< M psig a?

kI c. Containment Ventilation Isolation - U$

1. !!anual Not Applicable Not Applicable R16 gM 2. Automatic Isolation logic Not Applicable Not Applicable co -

t2 F-) s t 2-_ e._c:-rc__'_- _.-z 1 -__r _ -

                                                   -._u.twaa.       _~cx+ _- m -- ~' --~m-             --  _-~-.w - - .x -
                                                                                                                         - .a-_na--_.. m_-     +,wsw. ..m_-a..-.--_a_.,w---_. . _ _ -      --x_     _.._ _--

m TABLE 3.3-4 (Continued) m . , o C ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUNENTATION TRIP SETPOINTS o 2 r

                ,              FUNCTIONAL UNIT                                                                                              TRIP SETPOINT                                         ALLOWABLE VALUES c

z 3. Containment Gas Monitor -3 5 8.5 x 10 pCi/cc $ 8.5 x 10'3 pCi/cc

               -4 Radioactivity-High w

4. Containment Purge Air Exhaust 5 8.5 x 10

                                                                                                                                                                    -3 pCi/cc                                             -3 pCi/cc Monitor Radioactivity-High                                                                                                          5 8.5 x 10

'. 5. Containment Particulate -5 -5 ' ' 5 1.5 x 10 pCi/cc 5 1.5 x 10 pCi/cc Activity-High

4. STEAM LINE ISOLATION

a. Manual Not Applicable Not Applicable y

b. Automatic Actuation Logic Not Applicable Not licable co c. 2.

m Containment Pressure--High-High -< 2.81 psig < psig

                                                                                                                                                                                                -                                                                  i

d. St: = Fic in Tu; St = linc;- : A'funti n definc :: : A f;xti= def xd =

                                               "ip; Coincider.t with 7                                                  --

Tailsu;; A 4 cerie;;;;d ?;i!;-;; A 4 ;;. .c;xdi g i Or St =,Lin: Prc;;;rc' Y =:.= L= ing to 7" of f;il :t;;; to "" of f;?! st = f ?=

                                                                                                                                          -f ? = Mt. ;; . N;-d '"J%                              kt ;;; 0" : : '"" ? ::d m.=.-
                                                                                                                                           ? x = d i t ;.. ; 4                                   x tt;;. ; 4 inc ;;;ing N'"' S S                                                                                        i" ;;i"9 Iio"il7 to o                                 II""il7t;o4 Co "e"2 4 ;. . ap;..;'iq to                                   i g i; ?11:5% of f:P :te =

l

                                                                                                                                          -11"" of f;il-:t: = f1;u                               f! = t f;11 1::d l            Rg-                                                                                                                              t full 1::d.
            %                                                                                                                                                                                                                                                      i l            Eg                                                                                                                             r        s ce q                                                                                                                              -avg - - ~ -

r a

                                                                                                                                                                                                              .ee~       ' ' -
            )

rz el. Sb C= P=u-c. - le- > 600 psig steam Iine pressure (44 I)

                                                                                                                                                                                                > +89.psig steam                                                   ;

y Tine pressure (Af f. l) ' e,A4p.);n .dm U~ fon=re tale- Hp4 i S xv,opw; (Abk 2) fE107.0f ri (MNAT

            *m                5. TURBINE TRIP AND FEEDWATER ISOLATION O'                                         -

et% st.72,

a. Steam Generator Water level-- .< -765 of narrow range - < iP65 of narrow range High-High Instrument span each steam Instrument span each steam

                          .                                                                                                               generator                                             generator

b. Automatic Actuation Logic M.A. M.A. f ,

 ..-,n.__--                 _         --,,._ma  --
                                                         ,.t     ----sm_,__.__._.m_..____.__,,___._____m_-_.
                                                                     .                                                              __ m     .m   ,____.m_-~mm_,__s              . v....,,-.m-     ,s-.w.-     .---.%,+-          ,-.-ww__,y-.. _,,_--my,-,,

o, TABLE 3.3-4 (Continued) E gg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS 5 z e FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VALUES . C ({ 6. AUXILIARY FEEDWATER

a. Ifanual Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable ,777,

c. liain Steae Generator Water Level-10w-low -

                                                                    ?*"ef--            = 27    ' I "' ef      ar--r  :~;-

1***4 ,,T....._. g , .1 : ... ___ u T__.__

                                                                                               ...1.-.-.
                                                                                                             . ____ ___u
                                                                                                  '.': ':2 L'_:C_ ~~ ~
                                                               .c.     ,..L,l ;[ . ~~"

{{ d. S. I. See 1 above (all SI Setpoints) 4' e. Station Blackout 0 volts with a 5.0 second 0 volts with a 5.0 1 1.0 second E$ time delay time delay

f. Trip of Main Feedwater N.A. N.A.

Pumps

g. Auxiliary Feedwater Suction > 2 psig (motor driven pump) > 1 psig (motor driven pump)

Pressure-tow [13.9psig(turbinedriven [12psig(turbinedriven K98 pump) pump) E h. At iliary Feedwater Suction 4 seconds (motor driven 4 seconds 1 0.4 seconds N -@$' Tran>fer Time Delays pump) -(motor driven pump) Er S 5.5 seconds (turbine driven 5.5 seconds 1 0.55 seconds 33 pump) (turbine driven pump) R133 , .o 00 - b Ca - - aa - 4

k m \

4 m i 4 M .

q k N 4 ! 4 k n 6 ' d? I l

                                                                                                            @           e i ,&                                                          9

v4

?'

Ik vT

f; ' t 4 il gd M '

TS Ns M k p f hI

             $                                 h vi m

mf

                                              's                     w                                       o
        .m 0a  W N

D

                                             % 4(

4.1 4 '

4 VI 4 1i A '5 C yy d y, j

-)

v , l j Y 9 ! 7 . v o

          '                                             D                                D                  4           5 4

i

          ,,                                 a s        t                               t                   q                      ;

a ,( N e 3 s ' it ,- I y

                                                       "                                   i 1   g                              *.t o   -

v .s

                                                       *-         v, a                  e 4?

s, y, a y, q k

                                              ,              7 E                                                       Q         :

3d

                                              .t w     44 i.

g4 d g, M... s, I

                                                                                                                       ~4 1                       6"         .

g5 4si di ins 'y L4 *1 1. N1

Iq N@g, e @< !

s 9 vi S 4 4 ,, 1!u 1jw l 4 1 1, y1 gs . N. .h = s di N ). i *I g Y.I ylN"$

             }I      j .j
                     .?

d U N{ d.E N )* "8 B {'4 ' g tk"h -

4 a

l l l I } I ( ) ! w II & l kk kI l 3 s's

                                                 =         h N4*s i

M f \s M '

                                 %k                     %          N f                                       i 4                      c          44 S              E       vr        A             5 k

N

b i

                                    ;                               b                                        i E                               t v                            >                                g                                        1 M                             ke                              t c 1e 4                               4%

4, ss b g , sM k)% 4 .h *t NI 9 k

                                 -4                     %         e4 g

d M> A\ *4 c y;

44 4

, b S - 7 y' g

                  ,             4g at                                 'i am                               t                                   -

4"e d d 4 14 i x W s^

                   $           au1 a , % .;                        .

4 q , s fs? i Id '

           ?                 %a 4            4 x I4 a         4(                                        !
          <y      ut          -

Y), 9 9 , nN N o LE ad'

TABLE 3.3-4 (Continued) m 8 ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRtIMENTATION TRIP SETPOINTS

7 FurlCfl0flAL UNIT TRIP SETPOINT ALLOWABLE VALUES c-z

              /7. LOSS OF POWER O           a.      6.9 kv Shutdown Board Undervoltage Loss of Voltage

1. Start of Diesel Generators .

R33

a. Nominal Voltage Setpoint 4860 tolts 4860 volts +97.2 volts

b. Relay Response Time for - 0 volts with a 1.5 second 0 volts with a 1.5 +0.5 Loss of Voltage time delay second time delay -

2. Load Shedding

a. Nominal Voltage Setpoint 4860 volts 4860 volts + 97.2 volts

b. Relay Response Time for 0 volts with a 5.0 second 0 volts with a 5.0 +1.0 t'

a Loss of Voltage time delay second time delay - w b. 6.9 kv Shutdown Board-Degraded

        /,                  Voltage

1. Voltage Sensors 6560 volts 6560 volts + 33 volts

2. Diesel Generator Start and Load Shed Timer 300 seconds 300 seconds + 30 seconds

3. SI/ Degraded Voltage Logic Enable Timer 10 seconds 10 seconds + 0.5 seconds

8. ENGINEERED SAFETY FEATURE ACTUATION SYSTEli INTERLOCKS

a. Pressurizer Pressure 197f.2 442/

    .h                      !!anual Block of Safety injection P-11           $ 1970 psig-                               i 490a psig

e. g

  , ?e 35.;
    ,        - ca r-    r.. : . ,x....;n , sp,,cg:at;... ;; te t.c 5 p3c_nt_e ;t               t%     t;_ty r;n =g ; te 3e           g._,:_3
    ?@

J m t e.g r ' f e!!c::ing cc-p?rti - c f t h =0d i f4 ' -- - '- ---- - ---' -- k ed O.* N eu.

                                                        +.m r---r- % 3   y   w-     ,  w-w_, +  w       .      +-  ,&.e       _-   m   A   e------L -a--------------a

1 I i

               ,                                                                                        n                                                 ,

F= F% i

                                                                              @                                                   e m_                        -                         x m
                                                                                                        >                                                4 l
                                                                                              -          e o        -

3 m .o N* i

y [. w l

                                                                                        .N
                                                                                       .O       >

x- 2 7 , c

s k M l

m g 3 w n 4 : g .o 5 5o H t( I m W

            >                                                                           '       =        a z                                                                                 **               d                  u                  I
                                                                                              'm. a
      -     W                                                                          =                                                               .i E     g                       u ev        e t'                             ,

4

            <                                           .u.

bm $ I. 3 a)

                                                                                       +1
                                                                                               +1      .o.
                                                                                                        .              .4                                i 1

I s, g wi c u ,3 e o e a . N .-. m.

            ,                                                                          n      o         v.

e .c. i x

      =                             we                   n,                            -      m         p    z            e                   t          i r                                                                                                 ~                  \

4 g x 5 ;

     -                                                                                                 a                                                  !
     >-                                                                                                 0            -
     <                                                                                                 w.

j

a I 4'= 1 W . - Ys 3 v M

o - g b ~ u 9 CO W 'O >-

                                                                                      .o i

m s. s 8 m * .hs > O - g

e w s x - > e - . r e w a }, s U >

o gx % > C - 1 v m - > c' .o 4 v > w g- u ,5 m m e m

                                                                                      *.     .8       -

t c> e m* 5 a.

                                   --                   $..3 8

W3 w) h o = w m

     ~
           .M el                   m             ,m, m

o m m v

.c.

z 3-$ - $ = - m u < < M w Il! e 1! E S B, .o ,

     .-                     J
                                '             ->.                 s
    $            $          Ji                ..

St { , j i SS

5 4 6 as

1 , w ,e_ n , x o u N u- ' uo 5y 0 o d'l5 " '5 o s g

h. !

o m

                 =..m.

as o.o o , . ., . . .g . m g 4-w as >. o e 3 .%

                                              * --             e.            e                                                              4
    "            -c           .3e,             " "            w u.                             "            o e-                                                    ,
    $            $3           j i--                                                                         -          M=- ; s.                   &

5 "~ .:!L iio '"o i; g = w s3 ,= g .8 ,, yhd 3 g >. m > -m a g N -. i. -

                                           ---      i:
                                                              . , .s. s. Sa        i2             v     v                    W w            wu q       L,,"              c)--
                                                   - >         e y--

J e < , 6 g-

                                                                                                     'e                          t4             4
                 $odjb            l,! . 44 i:lllllI            E E .c o ;; ;-

t:5ll ~$E

u

                                                                                                           ;;                                   6-eo a vm - .!

ew .o,c ! iy, ,_ x . 2 wz y; o . e u m. a a g b.1' 41

                            ,        -       .x v,            eo                 --.
          =      m-                                                                  $ 51! e d              o                                         -i de                        m                  36-            a

y a

zw w ac 5,q Eeo=mvvem e

                                                                                                                     .s
          =      em                                                               _

U o Z >=

                 ~^
                                                                            ,E b                                                         M fs             #

a Y ~d e,i 3 .f M 4 i n & . SEQUOYAH - UNIT 1 3/4 3 Amendment No. Lt. 63 i December 31, 1987- .

I l TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONDS

3. Pressurizer Pressure-Low R5 l

a. Safety Injection (ECCS) 1 32.0(1)/28.0(7)

b. Reactor Trip (from SI) 1 3.0 ;

c. Feedwater Isolation < 8.0(2) ;

d. Containment Isolation-Phase "A"(3) 18.0(8)

e. Containment Ventilation Isolation 5.5(8)(13) gj ;

f. Auxiliary Feedwater Pumps < 60(11)

R81

g. Essential Raw Cooling Water System 1 65.0(8)/75.0(9)

h. Emergency Gas Treatment System < 28.0(8) l 1

n s ,t,k___a:_,l< dad l 4, n___...__ e....._ e.___ . : ____ue_g

v. .c <..a ,.. ...,m.....

                                                ,_:_             __ r s .e.r .r.e ,s                                                  n(7),ee n(l)             R59     l
                                                                                                                           ; .e n..              , . . .

L. /t___ D..=._.__

                                .. ..            T..J.r s'.-- <*
                                                   '                           f T s, 9 A
                                                                                                                                                                            ]

1

                            .e ._ . J *....
                                           , _ _...T ,_ 1. ., . .4.. .      _
                                                                                                                                  .O.h .
               .a .        r._._..,4.-_.._._..
                           .         ..          . . ...           3. .<___
                                                                          . . . .nu..- . . .. . _ n a.n (3 )             ~
                                                                                                                           ; ..  ,o.n.(8),,o.n.(9)
               ..          r.,.u.__-.__..
                                   ....            . . . u. . _ ._. . n. . .o. 4. _ ,...
                                                                                          , _ 3. .a. :. ._ ._.             um          a__i:__      .t,+-
  $Vg                                                                               ...                                       . . .    ..rr..         ..           R81 ,    {

i.... n. , . .- _,.

                                               . . e ._ ._.#.. .a.. .._._.

n..__,

                                                                                 ....y.,
                                                                                                                            , en(11) cM                                                                                                                          -

r , , _ _ + 4. ., 3 e. ., . r. .,._34 .y- v. . .a.. ._ .e...__

y. . . . . . . . . .. v . .y...m ;.v..n(8)rwe er v i...v n( o. ) I

u. . e_______.. n.,

                           . _ .              ...,   ...        T._._.o._.__..
                                                                     . _ . .               .e,....__
                                                                                                  . . . . ,               ; ..o.n(9) e l

5. St ::i._ ,"lew 'r_ Tac Stes: Lin : - "i-h Cci cident

_ _ , , ~ avg

               ..          e ..
                           .    ., ,.         " ' _ : _' _' . < _ _ r e r. .n e ,s

_ . ,m...s..... , . e

                                                                                                                                 .n.n        (7),,n, .n(l)   -

R59 ' l t

v. n..._...__ vi T_:_

i,.y r s. _.n _ g .e .j ts . e a l i g.v

              ..           e..__.A_.~.__ . . . ... T , _1.      ....: . .                                                 ,in,,,n        .
                                                                                                                          ~
                                                                                                                                 .n ..n(8),_n n(9)
              ..           r__.._,____.                    ,,_,
                           . . . . . . . . . . . . . ....a....,___n,t___              , . . . .

n,n(3) o

c. 00nt:in nt ":ntil:tien I: 1 ti n ":t '.pplic:b1:

               ..          u. . , n. . , ._,. e.._ _.-.       4 .a. _ ._ &,__.        _r.                                   . e.n(ll)                               'R81
              ,,          r. . ., ._._. .. 4. ., 3 n ,.
                                                     ...-       n. _ ._ ,. 4 _   , _-u.a. _ _.. .,...-

e....__ e, n(8),,,n(9)

                                                                                                                          ;      v,..               ,, ..                  >

h. Ste:: Line I: 1 tion ~

                                                                                                                                 '^ ^

4.. v

           .:             r _ _ _ _ _ _ m,,
                          - . . , . . . .            e. . . v _   ..._   . o. . _. . _    e,...__

an v. n(9) ,

5. Wee *Nn Jhs.,ba hawa h/a ~

A:sf~ a.~ .Hr.- Li,,e .n o M o , $ 9.0 SEQUOYAH - UNIT 1 3/4 3-30 Amendment No. 55, 77, 106 March 13. 1989

TABLE 3.3-5 (Continued) ENGINEERED SAFETY FEATURES RESPONSE TIMES j INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECONOS

6. St:= 9 = 8- Te : r L' :: us;g gs C;in;iint .dth Steam Line Pressure-Low M a. Safety Injection (ECCS) 1 28.0(7)/28.0(1)

b. Reactor Trip (from SI) < 3.0

c. Feedwater Isolation < 8.0(2)- ,

d. Containment Isolation-Phase "A"(3) [18.0(8)/28.0(9)

e. Containment Ventilation Isolation Not Applicable

f. Auxiliary Feedwater Pumps < 60(11) I

g. Essential Raw Cooling Water System 65.0(8)/75.0(9)

h. Steam Line Isolation < 8.0

1. Emergency Gas Treatment System k38.0(9)

7. Containment Pressure--Hich-High *

a. Containment Spray < 208(9)

b. Containment Isolation-Phase "Bu(12) '

[65(8)/75(9) R86

c. Steam Line Isolation < 7.0

d. Containment Air Return Fan i540.0and1660 j 8. Steam Generator Water Level--Hich-High l a. Turbine Trip < 2.5 ;

b. Feedwater Isolation '

[11.0(2) l

9. Main Steam Generator Water Level -

Low-Low

a. Motor-driven Auxiliary < 60. 0 (IV)770 I Feedwater Pumps (4)

b. Turbine-J :ven Auxiliary g m

                                                                      < 60.0
                                                                      ~                                        i FeedwaterPumps(5)(11)                                                                       ]

l I l l I 1 September 9. 1988 SEQUOYAH - UNIT l' 3/4 3-31 Amendment No. 55,-59', 53,J7. 82

TABLE 3.3-5 (Continued) TABLE NOTATION (10) The response time for loss of voltage is measured from the time voltage RBI is lost until the time full voltage is restored by the diesel. The response tirne for degraded voltage is measured from the time the load shedding signal is generated, either from the degraded voltage or the SI enable timer, to the time full voltage is restored by the diesel. The response time of the timers is covered by the requirements on their setpoints. (11) The provisions of Specification 4.0.4 are not applicable for entry into RBI MODE 3 for the turbine-driven Auxiliary Feedwater Pump. (12) The following valves are exceptions to the response times shown in the Table and will have the values listed in seconds for the initiating signals and the functioli indicated: Valves: R86' 40 Response FCV-67-89, times: 7. b , -90,75-10gg)/85gg) Valve: FCV-70-141 ' Response times: 7. b , 70(8)/80(9) (13) Containment purge valves only. Containment radiation monitor valves have a response time of 6.5 seconds or less. o (I'l) A us ed inc44 'Op 7;T . Alys. Acpi o ,sa Les are W 7ro in s44 d A.,v-,s, 5,/. .u y,o.. ,,,4.d.., a 4.,,,4 sold s44 pro /edio., uk d, s,.d ot4*4 , An'ses (y 4 s.,d h,e Ld,,y p, p) 744 nAs t4 ay,~< li ,ao muu.y .,4 THEpm44 PwsR ja, exem of .503 ATA SEQUOYAH Utill 1 3/4 3-33a Amendment Nn. 29. 77, 82,1M :

                                                                > larch 13, l89

TABLE 4.3-2 M S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION Q SURVEILLANCE REQUIREMENTS 5 c CHANNEL MODES IN WHICH 5 CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED w

1. SAFETY INJECTION AND FEEDWATER ISOLATION

a. Manual Initiation N.A. N.A. R 1, 2, 3, 4 R51

b. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3, 4

c. Containment Pressure-High S R Q 1,2,3

             $                 d. Pressurizer Pressure--Low                                                                                                                  S                 R                             Q                              1,2,3

_ at.lJ

e. Oi';mrc.tial Pressura 5 2 Q 1, 2, 3

'~

E Sctucca Sic = Linc:;--fiigh Me"'_ ,_ E 8

f. Stc= Fl= in Tuc Sic = 5 R Q 1, 2, 3 Linc;- ."igh Cciacidcat with

_PA__ f *_. T. a vg_ _" 'f" _

                                                     ' " " " ' ~
                                                       ._ f _ . -
                                                                   ' ~ ' - ' ' ~

Y 54. Le*,,e Pressure--Low -S g Q 1, 2, 3 +

2. CONTAINMENT SPRAY

a. llanual Initiation N.A. N.A. R 1, 2, 3, 4 R51

r. b. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3, 4 Re EE e c. Containment Pressure--liigh-liigh S R Q 1,2,3

         - $u
           ~                                                                                                                                                                                           .
         . 4 eue 3.
  <.~v,. '
                    ..+-..e_   ,_i.      . , _ .            v                                                                                  m                       su.w v'

e e -- ,---ae= w- ermm e..--+--m *. , -s, ____r.___.m._ .-__..ua-.-_.--_______-.m.m. .--._m .. - - -

i l l TABLE 4.3-2 (Continued) b EilGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION SURVEILLAtlCE REQUIREMENIS CHANNEL MODES IN MIICH CilAt!?iEL CHANf4EL FUNCTIONAL SURVEILLANCE E FUflCTI0flAL I! flit CIIECK CALIBRATIOff TEST REQUIRED

                                    '4) Containment Purge Air                                     5                R             H      1, 2, 3, 4 Exhaust Monitor Radio-activity-High

5) Containment Particulate S R M 1, 2, 3, 4 Activity-High

4. STEAll LINE ISOLATION

a. !!anual fl. A. fl. A. R 1, 2, 3 RSI R b. Automatic Actuation Logic H.A. N.A. M(1) 1, 2, 3 Y c. Containment Pressure-- S R Q 1, 2, 3 M liigh-liigh

d. S* _ - Fi= - T= Stc= 5 Q 1, 2, 3 , , , _ _ ___

                                    , :..- -m_ _ u. . ,u.
                                                  .       a....e..s_..
                                                                  . m.     .- . : , u. . _

g

                                                ' " " ' " " " ' ~ " " ' ' ' ' ' '

avg c/ Sle=~ L,*m Pressure--Lovs 3 g g 4g =-

e. AGral; Jim L* km R 4-Hy 3 A a 3

5. TURBINE TRIP AND FEEDWATER ISOLATI0ll c>

E4 a. . Steam Generator Water 5 R Q 1, 2, 3 IE Level--liigh-High

        ?R d
        "A v

b. Automatic Actuation Logic fl. A. N.A. M(1) 1, 2, 3 R67

       -s
        ~

6. AUXILIARY FEEDWATER Gu a. flanual N.A. N.A. . R 1, 2, 3

       -$7                                                                                                                                                          RSI

b. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 L.h___- %._._ .

y, TABLE 4.3-2 (Continued) E

gg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g; 50RVEILLANCE REQUIREMENIS x

e c: CHANNEL MODES IN WHICH 35 CllANNEL CHANNEL FUNCTIONAL SURVEILLANCE [[ FUNCTIONAL UNIT CHECK CALIBRATION . TEST REQUIRED

c. Main Steam Generator Water S 1, 2, 3 , __ ,

Level-Low-Low JWserf ", L , Q fM/YTD

d. S.I. See 1 above (all SI surveillance requirements)

e. Station Blackout N.A. R N.A. 1, 2, 3

f. Trip of Main Feedwater N.A. N.A. R 1, 2 Ri Pumps a

i' g. Auxiliary Feedwater Suction N.A. R M 1, 2, 3 R$ Pressure-Low

h. Auxiliary Feedwater Suction N.A. R N.A. 1, 2, 3 gi33 Transfer Time Delays

                   //.         LOSS OF POWER

a. 6.9 kv Shutdown Board -

Loss of Voltage

1. Start Diesel Generators S R tt 1, 2, 3, 4

   . y L'                                 2.       Load Shedding.                                                      S                                      R                                                                                  N.A.               1, 2, 3, 4                       R33 2$

8- R C O g

o$ ~ A r tn r r . T i. i c e ,-s.nic31 Spectr:c3 tie- te s:e igi r -..te.s 3 t tg e t 3. tg 5,,i im:ng (N h,a er.ri t g 9..t 3,7 _- ,s. r,_ i i r

                                                     . r .g , emp i ,, t ! ~. ,, r tie ~s:ric3t:,..,.j.5cg..e                                                                               3                       , i:e
    $I$

f

       %)        *
 ,         - ~ -               .      . .              _ . _ _ _              _ _ _ _ _ _ _ _

             -o4    .a 4 -_4o_4       _- , _ -      a aa.

t l

    *)
    'L *j                                                                   l k
         .                       ,,             .,         n   n            ,
     'g                          NN ss             #

4 N'

N\ l

       -l N

i e t I4 0 0 e 9 ; N, 4"I ; D M , 9 % s N sd jd * *

v .

2e . N t k Nr ,

                                '              5                            !
$      d                                                  9   w i             a N

s '; e Q 4 Q 9 ! l i t 17 T k d ! in 43 4

s MtN tV W i e

A

                                                              %            :)

j., g

       $                        4                         s5                 I 5                    ~1         M d.           4.e?g
                       \             k                    D      W
                       ) fA                    Y ~k       N e          .

4 . u . l

TABLE 4.3-2 (Continued) M 8 EilGIflEERED SAFETY FEATURE' ACTUATION SYSTEM INSTRUMENTATION S SURVEILLANCE REQUIREMENIS E CilANNEL MODES IN WilICil e CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE y .FUf1CT10t1AL Ut11T CHECK CALIBRATION TEST REQUIRED

                      /is.      6.9 kv Shutdown Board -

Degraded Voltage

1. Voltage sensors S R M 1, 2, 3, 4 g33

2. Diesel Generators N.A. R N.A. 1, 2, 3, 4 Start and Load Shedding Timer

3. SI/ Degraded Voltage N.A. R N.A. 1, 2, 3, 4 t'

s-Logic Timer Y 8. ENGINEERED SAFETY FEATURE t; ACTUATION SYSTEM INTERLOCKS n,

a. Pressurizer Pressure, N.A. R(2) N.A. 1,2,3 P-11 RSI Mem' .57 0 b. ',yp,A/')Pf

                                                                   "^
                                                                                                          . . . m".                 ,,m',

c. Ste m Generator N.A. R(2) N.A. 1, 2 Leve , P-14 UF 9. AUTOMATIC SWITCHOVER TO lll 3 CONTAINMENT SU!!P

 ' "z %

43 a. RSWT Level - Low S R !! 1, 2, 3, 4 , ,/* COINCIDENT Willi

  ??  '

Containment Sump Level - liigl i S R H 1,2,3,4 AND {t Safety injection (See 1 above for all Safety Injection Surveillance Requirements) fj. b. Automatic Actuation Logic H.A. N.A. M(1) 1, 2, 3, 4 R67

      ,$       *nn7p:      ing; t,;;9_gc37 Sp7 ig 3tygg te g 5 7 7_e_tme 3t tg ,g;7t 7 ye7 t=5_; gu 7m g r _ i i m; ont mp
                          -e-   fe!!c N g cc7 ?ctier of the medi'icatien, d ic N ecr is ear'ier.

3 Y

i I i

3/4.3 INSTRUMENTATION ! BASES 3/4. 3.1 and 3/4. 3. 2 PROTECTIVE AND ENGINEERED SAFETY FEATURES (ESF) i INSTRUMENTATION t The OPERABILITY of the protective and ESF instrumentation systems and j interlocks ensure that 1) the associated ESF action and/or reactor trip will ' be initiated when the parameter monitored by each channel or combination ; thereof reaches its setpoint, 2) the specified coincidence logic is maintained, i

3) sufficient redundancy is maintained to permit a channel to be out of service -

        -for testing or maintenance, and 4) sufficient system functional capability is                             ;

available for protective and ESF purposes from diverse parameters. The OPERABILITY of these systems is required to provide the overall , reliability, redundancy and diversity assumed available in the facility design ' for the protection and mitigation of accident and transient conditions. The '; integrated operation of each of these systems is consistent with the assumptions used in the accident analyses, t The Engineered Safety Features System interlocks perform the functions I indicated below on increasing the required parameter, consistent with the

setpoints listed in Table 3.3-4:

P-11 Defeats the manual block of safety injection actuation on low pressurizer pressure. P-12 0;f;;t; th; ;;n;;i ti;;k ;f ;;f;;y 'nj;;ti;n ;;;;;;i;n :n 'igh ;;;;; lin, fie- .ue is. te;; line p.;...re. ' de) P-14 Trip of all feedwater pumps, turbine trip, closure of feedwater F g isolation valves and inhibits feedwater control valve modulation . On decreasing the required parameter the opposite function is performed at reset setpoints wi... .- .-. 7...- .. .. .. -.... ...... P 12 netie; :;n;;; tie;k ef ;;f;;y inj;;ti;n ;;t;;ti;n :n high ;;;. line fi;w end lew ;;;.; line pr;;;;r;. 2;;;;; ;t;;; lin; i;;1;tien ,

n 'igh :t;;: '!;w. ^ff;;;; ;t;;; d" ti;;h; (i.;, pr;v;nt; >

pr;;;t.r; ti;;k ef th; net;d f.n;ti;n). The surveillance requirements soetified for these systems ensure'that the ' overall system functional capability s maintained comparable to the original design stanoards. The periodic surveillance tests performed at the minimum 1 I frequencies are sufficient to demonstrate this capability. 1 i SEQUOYAH - UNIT 1 B 3/4 3-1

1

l DEFINITIONS CHANNEL FUNCTIONAL TEST

1. 6 A CHANNEL FUNCTIONAL TEST shall be:

a. Analog channels - the injection of a simulated signal into the channel as close to the sensor as practicable to verify OPERABILITY including alarm and/or trip functions. !

b. Bistable channels - the injection of a simulated signal into the l alarm and/or trip functions,

c. sensor

                                ;/. / sto k verify   OPERABILITY
                                            ,/s - Ai        'e n of a r..~ 4including %*/ ryJ in6     %Jose
                                                                                                  *~/ w 14 seni ,t,.p.t 4            roun                      i  net   owMrar,y
                                                  /* / j'f/ d /'""rada    os p*edudla                                ,

CONTAINMENT INTEGRITY daO id A d a/

1. 7 CONTAINMENT INTEGRITY shall er.ist when: lR6: },

a. All penetrations required to be closed during accident conditions I are either:

1

1) Capable of being closed by an OPERABLE containment automatic isolation valve system, or

2) Closed by manual valves, blind flanges, or deactivated auto-( -

b. matic valves secured in their closed positions, except as provided in Table 3.6-2 of Specification 3.6.3.~ All equipment hatches are closed and sealed,

c. Each air lock is OPERABLE pursuant to Specification 3.6.1.3,

d. The containment leakage rates are within the limits of Specification 3.6.1.2, and s

e. The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

CONTROLLED LEAKAGE j

1. 8 CONTROLLED LEAKAGE shall be that seal water flow supplied to the reactor %63 coolant pump seals.

CORE ALTERATION

1. 9 CORE ALTERATION shall be the movement or manipulation of any component. R6 within the reactor pressure vessel with the vessel head removed and fuel in i the vessel. Suspension of CORE ALTERATION shall not preclude completion of q

movement of a component to a safe conservative position. { SEQUOYAH - UNIT 2 1-2 Amendment No.63-May 18, 1988 ; {c.

TABLE 2.2-1 M y-< REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

                  @     FUNCTIONAL UNIT                            TRIP SETPOINT ALLOWABLE VALUES c-      1. Manual Reactor Trip                    Not Applicable                        Not Applicable 5
                 -*                                                                                                             2 2 4 7.

2. Power Range, Neutron Flux Low Setpoint $ 25% of RATED Low Setpoint $-064 of RATED N

THERMAL POWER THERMAL POWER is t.4 % High Setpoint 1 109% of RATED High Setpoint H05 of RATED THERMAL POWER THERMAL POWER

3. Power. Range, Neutron Flux, < 5% of RATED THERMAL POWER with < 6.3% of RATED THERMAL POWER High Positive Rate ""

a time constant 1 2 seconds with a time constant 1 2 seconds

4. Power Range, Neutron Flux, 1 5% of RATED THERMAL POWER with 1 6.3% of RATED THERMAL POWER g3 m High Negative Rate a time constant > 2 seconds with a time constant > 2 seconds: 5. Intermediate Range, Neutron i 25% of RATED THERMAL POWER $ 30% of RATED THERMAL POWER Flux 5

5 1.3 x 10 5counts per second

6. Source Range, Neutron Flux i 10 counts per second

7. Overtemperature AT See Note 1 See Note 3

8. Overpower AT See Note 2 See Note JY i _

I N V.F.

9. Pressurizer Pressure--Low 'l 1970 psig > 4969 psig 13 w.a.

10. Pressurizer Pressure--High 5 2385 psig $ M 95 psig 9.t.7 %

5l y 11. Pressurizer Water Level--High 1 92% of. instrument span $ M of instrument span yf , ES G%

, oa ja g: 12. Loss of Flow 1 90% of design flow per loop

1 995 of design flow ll, per loop *

              ~%
             $z
              *o i                       " Design flow is 91,400 gpa per loop.

9 q

        -                     ,                        m    . . . . ,        ..       -      - - - - , -    , , - - - , . . .

             .                                                                TABLE 2.2-1 (Continued) m                                                 .

o 8

            -<                                         REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS                                                                          .

I

FUNCTIONAL UNIT TRIP SETPOINT

'

g/,,,, _n

                                                                                                           ,a.+ ~A "

ALLOWABLE VALUES 1 E 13. Steam Generator Water s leer gy 237 77;; < etnn;;t s in er n;, , L 7;;;; :=t7x;;t g7 p Level--Low-Low Tp---e rk ste- -;: ;reter Ep: er' eter g-.; rate-ro

14. c* D LI</"--+--n-

                                                                 a*        < " - * - - ' ' - '                      =-'<-n                  - * - - ' ' - - *
                        -Mir rt:5 :M L e Ste =                   3'TED T"E""_"1 """                      i cid: ;       8^TES "";Ti "'" cei ci d; .;

C::: rater W te- L:: 1 eith t r g.;r:te e:ter !;;;! eith ste = g .;; t- este- !:::T. 1 25*' ef 2 7. - r;ng; :--t -

                                                                                                                       ;   2 ef r:r-;; r: ;; im t m

:nt spe --e r k ste- ;:n:reter

! x xt 7: -e r ' ete- g n : ter 2

15. Undervoltage-Reactor R76

                                                                 > 5022 volts each bus                                 > 4739 volts each bus                               -

Coolant Pumps l

16. Underfrequency-Reactor > 56 Hz each bus > 55.9 Hz each bus

e Coolant Pumps

17. Turbine Trip A. Low Trip System > 45 psig > 43 psig Pressure B. Turbine Stop Valve > 1% open > 1% open Closure

18. Safety Injection Input Not Applicable Not Applicable from ESF

19. Intermediate Range Neutron > 1 x 10 10 amps > 6 x 10 31 ,,p3 v, Flux, P-6,. Enable Block

_.3,

       *g               Source Range Reactor Trip
                                                         .                                                                s2.4
   - @ o-          20. Power Range Neutron Flux                 < 10% of RATED                                        < -HE of RATED THERMAL POWER
    - *!,               (not P-10) Input to Low                 THERMAL POWER
                                                                                                                      ~                                                                        '

u <+ . Power Reactor Trips z Block P-7.

     ." O .

g.n N O n n N

i A b w I tt i % v 9' v t

                           +         r>                                 t
                     $)              (                                                             9         9          i 1-              ,                                 )                           w

Y vT D

            ,          {   '

f v a i l 1 (t . M sh' is 14

          -4              vf       M
  .m      l               s        "4

qd
3 T c

     '    g           0 $.         4 .1 M      '

e 4; A% v j

    %     N          %%                               vt                                            vf       vf N                                                                                                                .1 v                                                                                                                    ,

n . v b I 1 e m 3 1 Y bk

                                                                      %al s                         i e         t 3        ss "%          %)                  5-                                           .,       9
   .M d

9 s s b"' af

                                   .4                 g

( nf a S-

                                                                                                   '4        3           I l    I     g         N 0 .4          M!                 c                Mi                           P        L'        I E

4 VI A\ % vi y; J l b l l t L O 1 T 3*h , 4 T v u** s y* a\ w x-n d 1 g $% {g 3

                                                       &              {g                             -       "Y

6e s e d ,4 ~i 'm > di >

11 sv La 9w v1*N 't i

I4 @ar *f J

          ) l1               5 4: Nh                                  4                             'T g!'              '

li'! l 43l J 5 42 4 f~i 'J, j gj j e " 3 g 3 xl, n i i

       .e      w       -~    e  a.+ m.-se. -,. men    4  e4em     & 4 ,

Y

I b y i e t ; 1 kk kY 2 Ns %4 !'

   %                 *i          5                  *s Nk                             M f

ma 4 r ,p 4 c *4 4 vt r5 7 :

't                                                                                    !
,                                                                                        6 v

R

'                    %                              b                                 :'

I 1 7 > 3 y ?e t e ' l a i Ik Ik R 1 i t .% 11 ; 9 nt w % ( =t o i R ;I Mq Q-Q s

                   +p     -

G w at s , h i h 4 7 m

          ,        4g                              q
          '; S      o '!4                          b e44,                              4 4,       .

4"4 1ld 1 9I f t h^ M a , s .! : s

fs t 4 44 ia 4 e s .:. 4e 9N i 4 ud Y 11 4 %'* s o E i Q 4 l

                                                                            ..n.-   _

4

                                                                                                                                                                                                                                                                                                                         ~
              ,                                                                                        TABLE 2.2-1 (Continued)

E g REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS s x FUNCTIONAL UAIT TRIP SETPOINT ALLOWABLE VALUES g 12.Y % q 21. Turbine Impulse Chamber Pressure - < 10% Turbine Impulse <-tiKr Turbine Impulse

              ,            (P-13) Input to Low Power Reactor Trips                                             Pressure Equivalent                                               Pressure Equivaient Block P-7 Vf%                                                                    SnV%                                                                                                                                .

22. Power Range Neutron Flux - (P-8) Low <-95% of RATED < -3fM of RATED Reactor Coolant Loop Flow, and THERMAL POWER . THERMAL POWER Reactor Trip l' 1.6%

23. Power Range Neutron Flux - (P-10) - > 10% of RATED > M of RATED -

Enable block of Source, Intermediate, THERMAL POWER THERMAL POWER and Power Range (low setpoint) reactor Trips ' 4 24. Reactor Trip P-4 Not Applicable Not Applicable ! (2.v %

25. Power Range Neutron Flux - (P-9) - < 50% of RATED <-5tN of RATED Blocks Reactor Trip for Turbine THERMAL POWER THERMAL POWER Trip Below 50% Rated Power (b1+1pl M0TATION NOTE 1: Overtemperature AT ( M < AT, {K3 -K 3)[T( -T'] + K3 ( W ) - f y(al))

G+T5 l* TrS 3 2(1+IfA y+1 54 g> i

      '<                          where:  8+Tfh p                         3 = Lag compensator on measured AT i
       "[   y                                                       1                                                                                                                                                                                                                                                        ,
    ~

Id.Q <d//e.- g 1 3, 7 = Time constants utilized in the ' q --_,: x m - for AT, t = 2 secs 7j f 3 3 a s.

       *y                                     af,                       = Indicated aT at RATED THERMAL POWER y                                   K 1
                                                                        < 1.15 R21 g                                   K 2
                                                                        = 0.011
         -=     ,

4 1

                        ,                                                                                                                                            TABLE 2.2-1 (Continued)

REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS

E NOTATION (Continued) g NOTE 1: (Continued) j Z 1+rSq ! 3

                                                                                                                      =   The function generated by the lead-lag controller for                                                                                                                      T,,,  dynamic compensation p                                                                                    y&T t                                 =   Time constants utilized in the lead-lag controller for                                                                                                                       T,,,, t = 33 secs.,

1 1 r 4 secs. T = ~ Average temperature *F

                                                                             ~
                                                                                    -                                 = 1 7e , - z* - - = :;- ad T,,9 ra

co .,

_. 11.-_. . . . .- . ._.

                                                                                                                                            . . . . . . .       : i. a. _a :_ . u...

_-.,.-...> r.,,g i,__,-__..._.,.,_,,_,.,_ . _ , _ . _ _ . . T' 5 578.2*F (Nominal T ,9 at RATED THERMAL POWER) ; K 3

                                                                                                                      =   0.00055                                                                                                                                                                                                     R 21

! t P = Pressurizer pressure, psig P' = 2235 psig (Nominal RCS operating pressure) , S = Laplace transform operator (sec-I) ! and _3 f (AI) is'a function of the indicated difference between top and bottom detectors ! g of the power-range nuclear ion chambers; with gains to be selected based on measured n , instrument response during plant startup tests such that: ! hT M pi (i) for -- between - 29 percent and + 5 percent f (aI) = 0 (where q and . are rc RATEDTHERMALPOWERinthetopandbotkomhalvesoftheborere ctively, R 21

                  -w.xE2 to and gt*9b                        is o al                                                           in m cen of                                                                   D MN N).

CD 94 W H ' ,

            -_______-___ ________ - __._-_ _ _______-_- __.- -__ .~.                                                                          . .                    . - - _ . . - _ _ - - _ _ _ _ _ . _ _ _ - _ _ - _ _ _ _ = _ _ _ - _ . _ _ _ _ - - _ _ _ . - . _ - - . _ - _ . _ . _ .                     -            . . - _ _ - _ _ - - . . . - - -

I m , TABLE 2.2-1 (Continued) ,

           @~                                               REACTOR TRIP SYSTEM INSTRUMENTATION TRIP SETPOINTS SE E                        '

NOTATION (Continued)

E

           -'i NOTE 1:       (Continued) m (ii) for each percent that the magnitude of (q                          -

exceeds -29 percent, the AT trip set-point shall be automatically reduced by If50 hr) cent of its value at RATED THERMAL POWER. R21 (iii)foreachpercentthatthemagnitudeof(q!86

                                                                                                                    ) exceeds +5 percent, the AT trip set-point shall be automatically reduced by O                                  rcent of its value at RATED THERMAL POWER.

I 71s (u.+w_hQ-K I NOTE 2: Overpower AT.( M S,AT, (K4 I '} I 7 i+ T,5 i+7h

                                                 # 'Ib 5

I*' '45

                                                                                                                                            -"6 U

[( * *45

                                                                                                                                                                                           - T9 - f 2I0III 3    '

Where: = as defined in Note 1 5 1 ty = as defined in Note 1 AT, = as defined in Note 1

K 4 5 1.W R104 yg K/ = 0.02/'F for increasing average temperature and 0 for decreasing average fJ temperature

     *I
  ,  Y&                                         '6 i -   _g                                      y.           =
     ="                                                       The function generated by the rate-lag controller for T,yg dynamic compensation wg O

l

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

w _ TABLE 2.2-1 (Continued) E g REACTOR TRIP SYSTEM INSTRUMENTATION ~ TRIP SETPOINTS 5lE Z -

NOTATION (Continued)

E Q NOTE 2: (CoM $nued) T = Y3 Time constant. utilized in the rate-lag controller for Tavg, TY y= 10 secs. I = .. a n --,4 4. in- . - i i .v g 5 ~-- - " ' "

                                                                                                     , ,a s.e. a..
                                                                                                    . . . .          4. , na__ a - i.

4 K = E ro 6 0.0011 for T > T" and K6 = 0 for.T 1 T 0 o.

                                                                           -T               =      as-defined in Note 1 T"              =      Indicated T-avg at RATED THERMAL POWER (Calibration temperature for AT-instrumentation, 5 578.2*F)

S l= as defined in Note:1 f 2f0I) '= 0 for all AI NOTE 3: ' The channel's maximum trip setpoint shall not exceed its computied trip point by more than

           ,           y.                                  fpercentx AT rpan.

e cn /,9 {M , ll "w NOTE 4: The e 4 ,,,,aj's ino.2 da,, - 1. s m ,ana.,

                                                                                                      ; ,,,, daf *,,,aaa-/ph:n.

W A // oro [ e n ec 4 daf -f,jo fL sc/pbi.,f- Jy

                   . g: g                                         .

jVOTE S:.'<.I,,x,} '"6" N_ $ . . , _ , , , s - e,2= -..w-- -w- - - - - - - ^ ^ " - * * * " - ' ' " - " ' ' " ' '

      ,r,u,.s. -             ,e-h,,,21.- s.A a      <a4: s_m--m   L ~

w (j

l i

                                                                                 .             )

1 w

e &. et 1 L dd : , 11 j T 'V a I

                                                         +Y h.,

i

tt5 . S k f s

l

                 )+ Q*     0                     e i,          s. .

3 7, A V -T.EN i m 5 s s - *Q = y C ( v O ; 4 y .t -9 [ s a ss e' , r, 'fv i I ? T, 1 8 vts 4 R n I . g .I h Q

                        +Y                         %k II Y4 b                                               s y, y g ssm
                 ]s s                                                 <

a i 2

                 +4 e$ ,$

k k,rgf3 Y [ ,, 4 s 4 >

I i LIMITING SAFETY SYSTEM SETTINGS BASES i Intermediate and Source Range, Nuclear Flux (Continued) Range Channels will initiate a reactor trip at a current level proportional to approximately 25 percent of RATED THERMAL POWER unless manually blocked when P-10 becomes active. No credit was taken for operation of the trips associ-ated with either the Intermediate or Source Range Channels in the accident analyses; however, their functional capability at the specified trip settings is required by this specification to enhance the overall reliability of the Reactor Protection System. Overtemperature AT 1/ermo-4 M D "#/W" 3 The Overt eratureIeltaTtripprovidescoreprotectiontopreventDNB for all com ' ations of pressure, power, coolant temperature, and axial power ! distribu n, provided that the transient is slow with respect to ;j-in: transit elays from the core to the temperature detectors (about Meconds),J i and pressure is within the range between the High and Low Pressure reactor trips. This setpoint includes corrections for axial power distribution, i changes in density and heat capacity of water with temperature and dynamic compensation for pipin; thy: free th: : r: to the he; t: ;:r:tur: tt::t:rn With normal the core s etyallimit power distribution, this reactor trip limit is always-below as shown in Figure 2.1-1. If axial peaks are greater than nuci de 'gn, as indicated by the difference between top and bottom power range detectors, the reactor trip is automatically reduced according to the no ions 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 set point modification , because the P-8 setpoint and associated trip will prevent DNB during 3 loop

                   -operation ~ exclusive of the Overtemperature deita T setpoint. n r:: he;

eper: tier eben th:

                                                ' h:; a-S ::t;ci-t h p:-h:ibh :fter :::tth; th: 9,                 i i
                    -K2, and K3 %;et: to th: 0 :rt::;;r:ter: d: h: T ch:::: h : d r: h S; th: a-S                  ~

tpe!"t te it: 3 he; =ht. b thh ::t Of 0;;r:the, th: P-0 ht:th:k end'"tria functien: :: : "i;;h M:etrer thx tr!; et the reted pent h=1.

l4roupo<f 'lAr<moall Overpower A V ,8 "N '' '"y w/ ATD ' reyws Nims o4/*y $~ K a#* D ,! Th e.g., Overpower delta T reactor trip provides assurance of fuel integrity, ! melting, under all possible overpower conditions, limits the required. ! range High or Overtemperature delta T protection, and provides a backup to the i den ty an Flux trip. The setpoint includes corrections for changes in eutron for pipin;,dthy:heat #re? capacity of water

                                                   +" 00r:        withhe; to th:    temperature.

ter;;r:ter:and dynamic compensation 6 tecter:. Ne credit n: tcher fer e;eratier of thh' tM; i- th: :::i t at :::1y:::; 5:= v:r, it; functien:1 ::pabi4ty et th: :pecified tM ; ::tt h; h :;eir d by th h

                    - ; :i#icatier t Orh:n:: th: :: r:11 r:1 htility of th: P-o~ Ma                                h miNy:b:ter   Pr:t::thn of m,;w siteSyste~ .r         rbe Tas beds
                                                        -/<;o powds pmLA Onap/me                          n TWs  wirfwm' repod.J k WM9- 911Q, 'Kendu- Gn. Ayon. 4 fuessiva .L, 4 1%            Alma.                                                             A k J rrS EQUOYAH - UNIT 2                                B 2-4

biban .2, O ' At.,bs O<<< Lpa <. /-<< a .,,/ pa<po-- D<//. -D bY$Y heb ~ kj Cf W.fr/ l,, de home' . yam vs .sood 0N#feWe ** A r 'Yro)dij rep <a.ra.di da soo% ATP nA,a ca maasand A_y da p/a.,f A, eu.4: h p. 7he <ormahers ea<.4 /cy 'r s Y +<'pv 4 4 ocL/ opamd'sj' co-dNoni exieisj s/ d 4na of mensu /> -

      -4        .4 u:.y -M' 4;, 4 adid +4 egai- 6 t /!st po-a-wa&//a,,.< v .a.uw us is -/,4 .ocat,4 1 anf<.<as. ; v4a afdu ,cai        t,,  AU Ay a T a , - Ar d a ,4 .ranaa / Adr.r, e.m. m ea.,a ,c d A u /oop /h a. 3na4 M., d~ / ~4.<31n                                     1
      %         * / .< h t fg a i , - - a / -. . p n . . M :, / ,, 6 . 4 .,< h6          ,
                                                                                                 )

pada.,l<. ud,4 du hp Aw., a << so/ eyaa.rt,/ 4 4ya i w N e<4 Avf, toda/ ,u-- red:,4; A.,/>,, 4ha. , jusda.,4 may occur, mut/y i., .<ma// a./ .yas /,, / cop ,

      .y a t A       s T va /s u.        Au m /a      A 4.-,; a 4 ,    o ,e'   1 4 /s.p              l;
      .yeud sT ska .<h a W de md wda,, pa &.,y
      .                                        ru.att-&a., s,,a/ J- .dd,                         1
     .rua
        .<&A a/sco<awk.,p   .<    Oa.a,A,9
                                           , po . af.<4d48~ .,o f ,, w                  A,              ,

xenon o< o/4< 6 ,,.,ia.d eo., a . ,), ' i I i iI i i I

6 0

LIMITING SAFETY SYSTEM SETTINGS i BASES 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 i 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 i (2485 psig). The Low Pressure' trip provides protection by tripping the reactor in the event of a . loss of rear: tor coolant pressure. Pressurizer Water Level The Pressurizer High Water level trip ensures protection against Reactor Coolant System overpressurization by limiting the water level ~to a volume i 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. -j 1 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 p [

                                                                                                   ~
                                                           ~ 907 Above 11 percent of RATED THERMAL POWER anlautomatic reactor trip will occur if the flow in any two loops drop bel        89% of nominal full: loop flow.
     ' Above;?E% (P-8) Of "TED T9E""AL POWEP., a omatic reactor-trip will occur if-
   . the low in any single loop drops'below           of nominal' full loop flow. This 1

ter trip will prevent the minimum value of the DNBR.from going below 1.30 uring 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. Yith th: 07:rt per:ter: ! delt: T trip ::t pei-t :dje:ted t: th: v:h: :;;;ifhdfer2 h:p:; rthn, i the b-9 trip :t 75% "'TED THER"^L POMEP "*" pr::: t th: :' '?r v:h: Of th: ; DMSP, fr::

     -trintient;    withg:h;
                          ? l b:hu ep: 4-1.30 derh; n:rt:1 Op:r:th :1 tr:::htt: : d ::th'p;t:d      j Oper: tier 4de P 8 b,Lluly
                                                                                               .\

l i a SEQUOYAH - UNIT 2 B 2-5 l . j l

LIMITING SAFETY SYSTEM SETTINGS BASES i Steam Generator Water Level h d '0" Mh: Ste:r C :: :ter Wter L:::1 L:u-L:u trip pr vid:: ::r: pr:te:tter b y

                                   -preventing eperatier '!th th: :t::: g::;r:t:r ::t:r h :1 b:10;                       th ei-i-"

v l' : r:;;te:d f r :d:g;;t: 5::t 7;;;v 1 ::p::ity. Th; :p;;ifhd ::tp:*-t

                                   -pre"ide: 2110etn: th:t th:r; will b; ;;ffi: kat .;;t:r ing:nt:ry fr th: :t ::

00ntreters fet tster ty:t--et the t h: Of trip t: :ll:a f:r :t rting d:hy; Of th: :::i'itry e+---/e- a ..+ - e u ut,-.a .a i... c.--- c.-- ..-- u.+- i-m-1 N T Low Water leeam/Feedwater Flow Mismatch in coincidence with a Steam G trip is not used in the transient and accident ator_ i yses but is included in Table trip settings and ther-1 to enhance ensure the functional capabilit the overall reliabi the specified a Protection System. This tr tof the Reactor Low-Low trip. s redundant to t eam Generator Water Level The Steam /Feedwate l low Mi ch portion of this trip is-activated when the steam flow exceeds feedwater_ flow by greater than or 1 equal to 1.5 x 106 lbs/ hour. Th eam Gen the trip is activated when tor. Low Water level portion of _ water level drops low 24 percent, as indicated , by the narrow range in mer.t. These trip values in in excess of norm e sufficient allowance perating values to preclude spurious s but will initiate a r or trip before the steam generators are dry. efore, the require pu pacity and starting time' requirements of the auxiliary ater are reduced

                                    -y;t ; :nd :t ::

and the resulting thermal transient on the Reactor Coo g:n:r:ter; i: mini-i::d. Undervoltage and Underfrequency - Reactor Coolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump bus trips ; provide reactor core protection against DNB as a result of loss of voltage ~ or underfrequency to more than one reactor coolant pump. The specified set ! points { point isassure reached. a reactor trip signal is generated before the low flow trip set Time delays are incorporated in the underfrequency and q undervoltage power transients. trips to prevent spurious reactor trips from momentary electrical- ' for a signal to reach the reactor trip breakers following the simultaneousFor i trip of two or more reactor coolant pump bus circuit breakers shall not exceed 1.2 seconds. For underfrequency, the delay is set so that the time required BR for a signal to reach the reactor trip breakers.after the underfrequency trip set point is reached shall not exceed 0.6 seconds. SEQUOYAH - UNIT 2 B 2-6 Revised 08/i8/87

Sed;en .2.0 Bn> ~~Q g ; T 1 f D. >~~

+ + ss +t s, . s' 9 'gs '
~~h g i'i 4m i N% $ )t g. 4 b, 4 I i~~
~~.m 't L4 .f} }4 yR*'1. 4, , ,w m~~
~~ti .e-~~
~~.c* *p.~~
~~sn 4 9 is 4t e . U s:g~~
~~'t q~~
~~lq Q-~~
~~;M d .5 4 'd ,D , d et j# $pf 5n,wh4 p~~
~~s < e A~~
~~=k--u4 hs.~~
~~9- D~~
~~$) 4 4-N !s a4 5a 4 a! 'l 2~~
s ) i ? ; t- : i .j l dd : 2 JN,E ? w 1.E m vi 4 E rsy y6 - s 'I ' Mts3f y Q{bQ. D eT ' v y ({ 94L 9s k d$ E i M d 4 i 1 R x i a !
~~. ~.~~
TABLE 3.3-1 (Continued)' TABLE NOTATION n With the reactor trip system breakers in the closed position, the-control rod drive system capable of rod withdrawal, and fuel in the reactor vessel. The_-channel (s) associated with the. protective functions derived from the- ' out of service Reactor Coolant Loop shall be placed'in the tripped condition. The provisions of Specification 3.0.4 are not applicable. N High voltage to detector may_be de-energized above the P-6 (Block of Source Range Reactor Trip) setpoint.' ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less.than required by the Minimum Channels OPERABLE: requirement, restore the inoperable i. channel to OPERABLE status within 48 hours or be in HOT STANDBY-within the next 6' hours and/or open the reactor trip breakers, [ j ACTION 2 - With the number of OPERABLE channels one less than the Total i Number of Channels, STARTUP and/or POWER OPERATION may proceed j provided the following conditions are satisfied: '
a. The inoperable channel is placed in the tripped condition-within 6 hours. l

b. The Minimum Channels OPERABLE requirement is met;1however, lj l,,p =: :dditienet channel may be bypassed for up to 4 hours i
~~for surveillance testjng per Spegification 4.3.1.1.1.~~
~~c. Either THERMAL: POWER is restricted to less than'or equal -i to 75%,of RATED THERMAL POWER and the Power Range, Neutron ~~~
Flux trip setpoint is reduced to less'than or.e 85% of. RATED THERMAL POWER within 4 hours; the or, qual to i QUADRANT POWER TILT RATIO is: monitored at least once per 12 hours, d. The QUADRANT POWER TILT RATIO, as. indicated by the remaining three detectors, is verified consistent with the normalized symmetric power distribution obtained by using-the movable j incore detectors in the four pairs of symmetric thimble. i locations at least once per 12. hours when THERMAL POWER is greater than 75% of RATED THERMAL POWER. ; i September 17, 1986 . SEQUOYAH - UNIT'2 3/4 3 '. Amendment No. 39
- - - . - . . . . .-- . - . -. . - . . . ~ .- 'l TABLE 3.3-1 (Continued)
~~ACTION 3 - With the' number of channels OPERABLE one less than required by the Minimum Channels 0PERABLE requirement and with the THERMAL POWER level:~~
~~a. Below the P-6;(Block of Source Range Reactor Trip) setpoint, restore the inoperable channel to OPERABLE status prior to ;~~
~~increasing THERMAL POWER above the P-6 Setpoint'.~~
b. Above the P-6-(Block of Source Range Reactor; Trip) setpoint, but below 5% of-RATED' THERMAL POWER, restore the inoperable channel to OPERABLE status-prior'to increasing. THERMAL .
~~POWER above 5% of RATED THERMAL POWER. ;~~
~~c. Above' 5% of RATED THERMAL POWER, POWER OPERATION _may :~~
~~continue.~~
~~d. Above 10% of RATED THERMAL POWER,-the predisions of Specification'3.0.3 are not applicable. >~~
ACTION 4 - With the number of OPERABLE channels one less tsn requ' ired by the Minimum Channels OPERABLE requirement and.with the THERMAL POWER level: a. Below the P-6 (Block'of Source' Range Reactor Trip) setpoint,;
~~restore the inoperable channel:to' OPERABLE status; prior to increasing THERMAL POWER above the P-6 Setpoint.~~
b.
~~- Above the P-6-(Block of Source Range Reactor Trip)-setpoint, operation may. continue. .~~
ACTION 5 - With the number of:0PERABLE- channels one'less than required by. the Minimum Channels.0PERABLE requirement,! verify 1 compliance with the SHUTDOWN MARGIN requirements of' Specification 3.1.1.1 or 3.1.1.2, as applicable, within I hour and at least'once per 12 hours thereafter. l
' .l ACTION 6 - With the number of OPERABLE channels one less.than the Total Number of Channels, STARTUP and/or POWER-OPERATION may proceed provided the following conditions'are satisfied:- 1 1
~~a. The inoperable channel is placed in; the tripped condition within 6 hours..~~
l R .. l fle Ayed < The en Minimum _ Channels OPERABLE requirement is met; however, dditier,;i channel may be. bypassed for up to 4 hours- l R r surveillance testingrgpegification p 4.3.1.1.1.-
~~. AC1:0N / -~~
~~th th:Of-"-5:r Mr-5:r Chent:15, Of' OPERABLE STe"TL'? :channel:~ene d/07 **E" OPER^.T!?M lete thin the Tetal~~
~~y precett l enti' p:rfere:nce Of the n:xt re;uired CP^""El FLHCT!0F^L TEST~~
~~-pr0vided the iaeper:ble Chea"01 it pl:00d #"~~
~~the tri"?td tenditier "ith4" S h^"rt. .~~
; l R' --September 17,1986
~~SEQUOYAH - UNIT 2 3/4 3-6 Amendment.No. 39 l~~
~~, i TABLE 3.3-l' (Continued) i~~
~
e ACTION 8 - With'less than'the Minimum Number of Channels OPERABLE,-declare ' the interlock inoperable and verify that all affected channels ! o f.the-functions listed below are OPERABLE-or apply the'appro-priate ACTION statement (s) for those functions. Functions to : 1 be evaluated are: '
~~a. -Source Range Reactor Trip.~~
q
~~b. Reactor Trip i Low Reactor. Coolant Loop Flow (2 loops)~~
~~,Undervoltage=~~
~~Underfrequency R99 :' Pressurizer Low Pressure Pressurizer-'High Level~~
~~c. Reactor Trip. '~~
~~Low. Reactor-Coolant Loop Flow-(1 loop) h~~
~~d. Reactor Trip Intermediate Range Low Power Range s Source: Range -~~
1
~~e. Reactor Trip l Turbine Trip R104' 4 l~~
ACTION 9 - h ht:0 i ACTION 10 - Celete E'" + "F . ACTION 11-Oel;.t;G J ACTION 12 - With the number of OPERABLE channels one less than required by' 1 the Minimum Channels OPERABLE. requirement, be in at least HOT-STANDBY within 6 hours; however, one channel may be bypassed for up to 2 hours for surveillance testing per Specification 4.3.1.1.1 provided the other channel'is OPERABLE.. R46 l
+
~~SEQUOYAH - UNIT 2 ' 3/4 3-7 Amendment No. _46, 99 104 May 5.-1989~~
~~.In.ted "f " Tille 3.3-/~~
~~ACTION 9 - With the number of OPERABLE channels one less than the Total Number of Channels. 3TARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:~~
a. The inoperable channel is placed in the tripped condition within 6 hours.-

b. For the affected protection set, the Trip-Time Delay for one affected steam generator (TS ) is ' adjusted to-match the Trip Time Delay for multiple'affected steam generators (Tg) within 8f hours.

c. The Minimum Channels OPERABLE requirement is met:
~~however, the inoperable channel may be. bypassed for up to V hours for surveillance testing of oh channels per Specification 4.3.1.1.1.~~
~~. i i~~
ACTION 10 - With the number of OPERABLE channels one less than the Total { Humber of Channels,-STARTUP and/or POWER OPERATION may - proceed provided that within 6 hourg for the affected ' protection set, the Trip Time Delays (Tsand Tg ) ' threshold power level for zero. seconds time delay'is adjusted to 0 % RTP. I s ACTION 11 - With the number of OPERABLE channels.one less than the Total Humber of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 6 hours, for the affected protection set, the Steam Generator Vater Level - Low Low (EAM) channels trip setpoint is adjusted to the same value as Steam Generator 'later Level - Low-Low (Adverse).
I Alll t 1.3-2 vi E HI:ACIOR IHil' SYSil M If3'1RilHf fil Allull Ri sl*0tl51 IIMLS 8
. 5;i 2 FUNCTIONAL llNil RLSI'ON5L 11ME g 1. Manual Re.. tur Irip Not Applicable Z
g 2. Power Rasiije, Neutron Flux i 0.5. seconds *
~~3. Power Range,. Neutron Flux.~~
~~High Positive Rate Not Applicable~~
~~4. Power Range, Neutron Flux, :~~
~~liigh Negative Rate 3 0.5 seconds"~~
~~5. Intermediate Range, Neutron Flux Not Appiicable~~
$ 6. Sour'ce Hange, Neutron F lux . Not Applicable w -
F. O ' - 4 . 7. Overtemperature AI . 1 4 s ecosids ** '
~~8. Overpower - Al -~~
~~":t.";ffic.Lic 4 8.0 .suo,,d.,~~
~~9. Pressurizer Pressure--tow -~~
~~2.0 seconds~~
10. Pressurizer Pressure--liigh 1 2.0 seconds-

11. Pressurizer Water ' Level--liigli Not Applicable 1
Neutron detectors re exempt frons response time testing. Response time of the. neutron flux sign.el portion of. the chasinel stiall be meastereal f rom nietector utilptil. or inspist. of f irst electrosaic t.osapusiesit isi <.li.isiniel. m g,- ., . .. " - _ _ . , .- y & +- - - - 1,%.J g--g- %.g-y ,m,. , g .,y%+,.6g,_, . .
~~,,f _ _ __ , _ _ _ , _ _ _ _ , , , . _ _ , ,.,___,,,,,.m.~~
._ . . _ . .._. . , = - - - - - - - --
I AlH L 3.3-2 (Contiouseil) M klAC10R IRIP SYSilM IN51RifMINIAIION HISP0 fist IIHLS o . -- 8 FUNC110NAL'bNil Hl5PON51~ llML h 12. Loss of Flow - Single Loop
~$ (Above P-8). <-1.0 second --e N Loss of Flow _lwo Loops -13.
~~(Above P-7 and below P-8) < l.0 second~~
~~14. Steam Generator Water level--Low-low~~
~~.T u. + G'['; ?.a . ; r t ..;d ., '~~
~~, . &/,L/ .~~
~~15. Sic;;/f tcitt::- F: "i;.utd -!~~
Lee Ste= Cc;;cratcr " ate: it;c! "e t Ap;r ' ic A ! , 16. Undervoltage-Reactor Coolant Pumps 1 I.2 seconds
$ 17. Underfrequency-Reactor Coolant Pamg>s < 0.6 seconds .j w .
4 18. Turbine Trip . ,. o- A. 1ow Iluid Oil Pressure. Not Appiicable B. Turbine Stop Valve- Not Appiicatale
19. 7 Safety Injection Input from ESF Not Applicable

20. Reactor. Trip Breakers Not Applicable ,

21. Automatic trip Logic Not Applicable

22. Reactor Trip System Interlocks Not Applicable J
~~Y $ . h $ . iaLA if. 1.w-dlm fyI.-21 ),e<<n j ,4A.- e4.:,,4~~
. sow.iAA fr., Ad.., a.%. 6, smW' sebdae A. 'a <<.: '7:bi '
~~n4Ji 14 <<p- 4 auwia,, A THEM'U.~~
~~}kucER u, eaaan of 105 ATP. . _ _ . - _ . ~ - . _ _ _ . _ _ , _ _ _ _ _ _ _ _ _ _ _ . _ _ . _ _ _ . _ _ _ _ _ _ _ _ _~~
~~'l Iw f "G " .'~~
~~tallel3.3-2 l s f~~
~~;fune-donal Un.'f ' Astpome ~7?ma /M #7aar 34., ' fana,.4 ' W L l.cv;I - - k w L a w ^~~
l A . A C.$ L oop s T 4 & O w 4 (i) . ( P 4 5 D % R TP; P > s o x 2 7P) l u B. J/a , Ca.,.~ L sa 6 4 .2.0.1eo -4 :(a) - 1 Len/--do .L (Al/an. , ' EAm) C, Ion 4,:, / p m- - 4 .7.o .r,< , 4 (EAm) i 5 d i l 1 e
~~,l F .i: -l~~
~~TABLE 4.3-1 REACTOR TRIP SYSTEM INSTRUMENIATION SURVEILLANCE REQUIREMENTS SE~~
' CHANNEL MODES FOR MIICH CHANNEL CHANNEL- FUNCTIONAL SURVEILLANCE IS E FUNCTIONAL UNIT' CllECK CALIBRATION TEST REQUIRED m 1. Manual Reactor Trip 'N.A. R46 N.A. S/U(1)and R(9) 1, 2, and *
~~2. Power Range, Neutron Flux 5 D(2), M(3) Q 1, 2 and Q(6)~~
~~- 3. Power Range, Neutron Flux, N.A. R(6) Q 1, 2 High Positive Rate~~
~~4. Power Range, Neutron Flux, N.A. .R(6) Q 1, 2 g High' Negative Rate~~
=
~~w 5. Intermediate Range, S R(6) -S/U(1) 1, 2, and~~
~~g Neutron Flux~~
6. Source Range, Ne'utron Flux S(7) R(6) M and S/U(1) 2, 3, 4, 5, and *

7. Overtemperature AT S. R XQ 1, 2

8. Overpower AT S R. JPQ 1, 2

9. Pressurizer Pressure--Low S R Q 1, 2

10. Pressurizer Pressure--High S R- Q 1, 2 - R16

r >
~~W5 11. Pressurizer Water ~ Level--High- S R Q 1, 2 ' ?r E~~
12. Loss of Flow -~ Single Loop 5 R. -Q 1 7

13. Loss of Flow - Two Loops .S R N.A. 1 l
" R16 . 14. : Steam Generator Water Level-- S R Q 1, 2 Low-Low -
~~g .sw 6,a,.L.6% la.et- J K -Q *>~~
~~. Lo-lw (Abw< t, 2.~~
~~g . 1 00a. /werea k < Wa h> $d n f~~ .$ $ 0 Il~~
~~. Lew-low WAA)~~
~~C gts. LW sT~ j g: Q ~~~
~~. - - _ _ = -.~~
TABLE 4.3-1 (Continued) REACTOR TRIP SYSTEM INSTRUMENTATION SURVEll. LANCE REQUIREMENTS o CilANNEL MODES FOR WilICil 5 . CHANNEL CilANNEL FUNCTIONAL SURVEILLANCE IS-7 FUNCTIONAL UNIT gg CHECK CALIBRATION TEST REQUIRED e
~~15. Stec-/Fec&cter r! w "ic etch end S o Q 1, ? R16~~
~~-Lee Stec- Centreter "eter Leve!-~~
~~ro 16. Undervoltage - Reactor Coolant N.A. Pumps R Jt'Q 1 ,~~
17. Underfrequency --Reactor Coolant N.A. R ,W Q _ 1 Pumps

18. Turbine Trip-A. Low Fluid Oil Pressure N.A. N.A. S/U(1). 1 B. Turbine'Stop Valve Closure N.A. N.A. S/U(1) 1 R
s
19. Safety Injection Input.from ESF N.A. N.A. -M(4tR 1, 2 w 20. Reactor Trip Breaker N.A. N.A. M(5) and S/U(1) 1, 2,-and
M 21. Automatic Trip Logic N.A. N.A. M(5) 1, 2,.and *
~~22. Reactor Trip System Interlocks A. Intermediate Range .N.A R -5/U (3) 4 2, and~~
Neutron Flux, P-6 B. Power Range' Neutron N.A. -R-- M A. S/U (0) M A. 1 Flux,.P-7 C. Power Range; Neutron N.A. 'R S/U (31 # 1 I
~~Flux, P-8~~
~~0. Power Range Neutron N.A. R S/U (U) ( 1, 2 Flux, P-10~~
~~?8 @$" E. Turbine Impulse Chamber N.A. R- $/U (C)- R 1~~
g. 8- Pressure,'P-13 F. Power Range Neutron 4 "@ Flux, P-9 u N.A. R S/U (S) 4 1
~~.~ F.~~
~~G. Reactor Trip, P-4 N.A. N.A. 4/'! (9)- R - 1, 2, and~~
~~RSS~~
~~~ ~ 23. -Reactor Trip Bypass Breaker N.A. N.A. M(10)R(11) 1, 2, and~~
~~R46~~
~~$ ?~~
~~N. _2.__-._ - ._ _ _ -_ _ _ _ _ _ _ . _ _ _ 1~~
~~-1 i~~
i TABLE 4.3-1 (Continued) j NOTATION-With the reactor trip system breakers closed and th'e control rod drive system capable of rod withdrawal. 31 (1) - If not performed in previous # days, j (2) - Heat balance only, above 15% of RATED THERMAL POWER. Adjust channel l .if absolute difference greater than 2 percent. .i l= (3) - Compare incore to excore AXIAL FLUX DIFFERENCE;above 15% of RATED R10] THERMAL POWER. Recalibrate if the absolute difference greater than or equal to 3 percent. 1 DsisL/ (4) - "enua! ESF functien:1 i nput ch::k every 19 m:ntP (5) - Each train or logic channel shall be tested at least every.62 days on a STAGGERED TEST BASIS. The test shall . independently verify the OPERABILITY of the undervoltage and automatic shunt trip R10 circuits. (6) - Neutron detectors may be excluded from CHANNEL CALIBRATION. 1 l (7) - Below P-6 (Block of Source Range Reactor Trip) setpoint. , (8) - LD;l:/.MOnly, : 5 t:rtup er when r: quired with the rc :t:r trip '
~~' systep breaker: cle :d :nd th: contr:1 red dr'c: ;ytter ::p bl: of l red Lithdrrf:1 if n t perfert:d " previcu; 92 d:yt.~~
l
~~. (9) -~~
The CHANNEL FUNCTIONAL TEST shall independently' verify the R104 $ l operability of the undervoltage and shunt trip circuits for the manual reactor trip function. R46-t (10) - Local manual shunt trip prior to placing breaker .in service.- l Each train shall be tested at least every.62 days on a STAGGERED TEST BASIS. (11)- .R104 f Automatic and manual undervoltage trip. f l SEQUOYAH - UNIT 2 3/4 3-13 Amendment No. 46', :104
~~~May-5, 1989~~
~~TABLE 3.3-3 un rn~~
~~@o ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 5~~
~~x - e MINIMUM TOTAL NO. CHANNELS CHANNELS APPLICABLE E~~
- FUNCTIONA_L UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION -4 to 1. SAFETY' INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION
a. Manual Initiation '2 1 2 1, 2, 3,_4 20

b. Automatic Actuation 2 1 2 1,2,3,4 15 Logic

c. Containment-Pressure-High 3 2 2 1, 2, 3 X /7 #
s
~~d. Pressurizer- 3 2 2 #~~
~~T Pressure', Low 1, 2, 3# ' X /7~~
~~$ Clr/cAc/~~
~~e. Si"erent?:1 ' ' '~~
~~c..._~~
- ww - s ___ _ u : _u. . - , 11 .y F: r L::;; 3/;te = 14ne--2/stca;-- linc 2/5tcar. lin: MA- ^perating any etec: ?inc j 10 3 D un Be -
~~em a" B~~
~~.. I.~~
~~M et , W.@ W h.~~
~~.___,m_._--._ d~"~~
sr. TABLE 3.3-3 (Continued) v. m ENGINEERED SAFETY' FEATURE ACTUATION SYSTEM INSTRUMENTATION
-2 o ' MINIMUM sZ TOTAL NO. CHANNELS CHANNELS APPLICABLE FUNCTIONAL UNIT 0F CHANNELS TO TRIP e OPERABLE MODES ACTION C
z '. Ste = r! .- i : Tu i., ', $#8 m Q - Ste = Line "fp. ru - F r L::p: 2/ste : 1inc 1/ sits;; 1ia: 1/ sits: 1in 16* Operating any.2 :teer lines R33'
~~-GGINGInro.v. ui. v. o. . .e.., v. o. .m .s T - Leu-L u -~~
88 , avg i , 2 ' R, Feer Le:p: 1T avg /1: 0p 2T ny l'T c r.y t n_r_ _. -..3 _ . : __ avg avg IS* , i m_ o_ m , i____ u.s ,. .e .. v. a cn A. D.. , R33 l.'n. T. .h.. ar T. neu. r ut.v. u.
~~-8 .Steam Line Pressure-Low 3 -- j,~ gjj,_ ffy y4 g. -#~~
~~7, ,g ,4 - Ila. 1,2,3* 17 8~~
-Ferr Leep 1 prectere/ 2 p-e::ere: 1 p- : er oper: ting -16^
1==p :=y ?::;:: . ==y ? ?==p: i H33 3*to Be co m N i < Q. (D BB to tt 3 iD a r+ t1
~~- 2~~
~~. ' O .V~~
~~.-W @CD W pt t. .-~~
i
+-
~~M 6 m~~
' ' " -- --.._--____------t-' - _ - '" '1*'" Y ' ' ' ' #Y & ' ' ' ' ' " ' ' ' * '"4- --V.' *'---------.-___Q. -
L'.x------.----x--m ' - m'uus,k'- i %- r M au. 2
u, TABLE 3.3-3 (Continued) ig ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 8 gg MINIMUM
~~=c .~~
TOTAL NO. CHANNELS CHANNELS _ APPLICABLE i FUNCTIONAL UNIT OF CHANNELS- 10-TRIP - OPERABLE MODES ACTION EE
~~-< c. Containment Ventilation~~
~~[[ Isolation .!~~
1) Manual 2 1 2 1,2,3,4- 19

2) Automatic Isolation 2 1 2 1,:2,'3, 4 15 Logic
~~,s.~~
3) ' Containment Gas 2 1 1 1,2,3,4 19 Monitor Radioactivity-High 4). Containment Purge 2 1 1 1.2,3,4 19 Air Exhaust Monitor.
{;' Radioactivity-High us -5) Containment Particu- 2 1 1- .1, Z, 3, 4 19 J. late Activity;High' o>
4. STEAM LINE ISOLATION

a. Manual- -1/ steam line' 1/ steam line 1/ operating 1,2,3 25 steam line-

b. ' Automatic- 2 'l 2 1,2,3 ~23 1 Actuation Logic-

c. Containment Pressure-- 4- 2 '3 1, 2, . 3 '18 High-High a e . _ __ r i __. - r. _ i , , e 3> *
~~- ,g _ca_ t . -_. u.. . L..~~
Ea Feer.tre;: - ?/:te:e ':n 1/: tear 'ine 1/: tear ':nc Is* (j! Operating 2ny 2 stear ng ,:_..... s
-g. R33 mm ,* '_ . . - . ___ms 4 & w ay: ._.,:m Mw r - " ^-w-'~ - - - - - A------ -
~~TABLE 3.3-3 (Continued)~~
~~'EO ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION 5~~
x ' MINIMUM e . TOTAL NO. CilANHELS CHANNELS 5 ~ APPLICABLE FUNCTIONAL UNIT OF CHANNELS TO TRIP w - OPERABLE MODES ACTION
^*
COINCIDENT MITF EIT"ER - T -- avg rm.L~e-Lee .i_, 1, 2, ? ir r i -- or ,_. ir ic*
- n,;;;J.i. .'.",' ^ ' avg' ' '~r i_; avg ' ;,;gyg ',_. , .~ v -.-v. ., ~
~~OR YITu d.COINCIDEPT~~
. Steam Line Pressure- 2 ,8 g Low 3 /5 4 4'. ~**J',j- 2/>4 #~ 1, 2, 3 17 R
Feur Leep: 1 prc;;urt/ 2 prc;;urc; i prc;;urc IE* T Opcrating 1;;p any 1ccp; any 3 ?cep;
~Y G* W(th j ) *** ' " " .3l3/a kn 2/1 I" . ' ' 2lAla-Ike .) . ;y5
5. -TURBINE TRIP &.- 7 R33 FEEDWATER ISOLATION

a. Steam Generator 3/ loop 2/ loop in Water Level-- any oper-
~~-2/ loop in 1, 2, 3 -- EF/7~~
each oper-High-High 'ating loop ating loop y b. Automatic Actuation 2 .1 g:g . Logic 2 1,2,3 23 RSS e
~~o. U" *3 6. AUXILIARY FEEDWATER~~
~~=E r~~
~~2 a. - Manual Initiation 2 1 2 1,2,3 24~~
. ,0 w b. . Automatic: Actuation- 2 1 2 1,2,3 =w 23 w- logic-m-
~~e L---.- __ c - .m __ _ - . - - _ .~~
. _ _ .. ~. , _ - . . - - - - - - ~ . . , , , - . , - , - - - ~ _ - . . . _ - , , ,a.-- . - . . . . _ ..
, TABLE 3.3-3-(Continued) 19 ENGINEERED SAFETY FEATURE ACTUATION SYSTEN INSTRUMENTATION 8 -c MINIMUM EE TOTAL NO. CilANNELS CHANNELS APPLICABLE . FUNCTIONAL UNIT OF CHANNELS TO IRIP OPERABLE MODES' ACTION EE c. Main Stm. Gen. Water il Level-tow-Low w ; .e s., . ,.. u.....,,,_ :Ensul " fl" Driver Perp 3/ tr ger 2/:t ger 2/str ger 1, 2, ? 15*
any stm gen. ii Start Turbir.c- R29 Driver Perp 3/ste ger 2/str ger 2/ste ger 1, 2, 3 IE* any 2 sim. gen.
~~d. S.I.~~
Start Motor-Driven Pumps and-Turbine Driven Pump See 1 above (all S.I. initiating _ functions and requirements) u, e. Station Blackout 1 Start Motor-Driven to _ Pump associated- 2/ shutdown' 1/ shutdown 2/ shutdown e' with the shutdown ~ board board board' 1, 2, 3 c' 20 board and Turbine Driven Pump
f. Trip of Main Feedwater Pumps Start Motor-Driven Pumps and Turbine Driven Pump. 1/ pump 1/ pump 1/ pump' 1, 2 20*
~~z 3, g. Auxiliary.Feedwater 3~~
=g ' Suction Pressure-Low 3/ pump- 2/ptap- ' 3/ pun.p 1, 2, 3 ' 21* @& h. . Auxiliary Feedwater Ril6 %@ Suction ~ Transfer s, " Time Delays- ." $F
~~1. Motor-Driven~~
- Pump 1/ pump- 1/ pump _ 1/ pump- 1, 2,'3 21*
~~EQ! '~~
2. Turbine-Driven Pump 2/ pump 1/ pump- 2/ptanp 1, 2,.3 21*
~~%04 * - _ _ m_ _ . . . _ _ . _ _ _ _ = . _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ,___i~~
~~t' M W k W Y $ $ S 4, n % + n~~
~~& 5' $ 3' $~~
~~9 .e A 3~~
~~'I '%~~
~~4e Qg iD 4 i' i N% =~~
~~,{ i -~~
~~h M- M~~
, 1, 4, .- - 4, mm N .=% . t*
i
~~. p QS 9; .?~~
~~e s Q4t e d' st . t-~~
~~.t ls - - . -Q l JN 4 b4 Y' D , cf W.~~
~~h.h{ 5,h,4, !- 3~~
~~% 1 i i- m G G S-at: %~~
~~qsjsf C~~
%,4 w > ; -4 % 'I ystr> >
~~i, s ens u!! rj~~
~~." s 1 .~~
~~y jl:4 94 4 d: 9-E d% : 5 y t I am a '~~
~~s. ,rs3 es y , 9 .~~
I
~~.e e a e , .~~
~~d i~~
~
~~,...w . m .m..4 % a j ~!~~
t
~~' i e~~
~~_N w -% k w ~ %~~
~~% O D D '*n y~~
~~g,, m m m n~~
@V *
~~d s'~~
~~k ss I o s~~
~~~n .t ?~~
~~y j -p j -t v .e a -~~
~~,u (As 'n : m 9 n 1, w 4 *v, d~~
~~m M .e-~~
~~.c I-1 e e ,~~
~~i I' 2jh t. M 4N A -\~~
~~=5 het) p{ R, ,b,4 .~~
~~o 4 .+ 4 , ' d $ k ('4!~~
~~= -~~
~~h s, , ,Y ,.~~
~~.h >~~
a < n i c I 3 3' . ae>s) s< < JI' g,.#.= M *A'4 r t - o s, . g rN tN g %
~~~ .O s~~
~~rs3 r s. x gG~~
~~4 4~~
N
,. a M 0 ,-%s - .* m. ~ - - e e
~~t-TABLE 3.3-3 (Continued) Q ENGINEERED SAFETY FEATURE~~
~~ACTUATION SYSTEM INSTRUMENTATION O~~
4
, MINIMUM c TOTAL NO. CilAfiNELS C11ANfiELS APPLICABLE 5 FUNCTIONAL UNIT. OF CIIANNELS TO TRIP OPERABLE- MODES ACTION -e .
~~N #.7 LOSS OF POWER~~
~~a. 6.9 kv Shutdown Board~~
~~--Loss of Voltage ,~~
~~I l I~~
1. Start Diesel- 2/ shutdown 1 loss of 2/ shutdown 1, 2, 3, 4' 20*
~~Generators board voltage on board any shutdown board~~
2. Load Shedding 2/ shutdown 1/ shutdown 2/ shutdown ~1, 2, 3, 4 20*
~~y s board board board w b. 6.9 kv Shutdown Board~~
~~$ Degraded Voltage RI8~~
1. Voltage Sensors- 3/ shut.down 2/shutev+f 2/ shutdown 1, 2, 3, 4 20*
~~board board board~~
2. Diesel Generator 2/ shutdown 1/ shutdown 1/ shutdown 1, 2, 3, 4 20*
Start and Load -board board board Shedding Timer g 3. SI/ Degraded 2/ shutdown 1/ shutdown 1/shuttfown 1: 2,3,4 20* y> - =y Voltage Enable board board board g Timer
.m e g 5 WTE: This techaica? 07^Ci#icati"^ it 10 50 N: ' --' -' ' O """; !!'O 5 t 'e t"" f a l f ~ ~ "g t '= l e t *
~~f m? ! !!u;-~~
g g- ert:gr. 5
TABLE 3.3-3 (Continued) ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION a SE .
~~, MINIMUM c:~~
TOTAL NO. CHANNELS CilANNELS APPLICABLE sg FUNCTIONAL UNIT OF CHANNELS TO TRIP OPERABLE MODES ACTION w n> 8. ENGINFERED SAFETY FEATURE ACTUATION SYSTEM INTERLOCKS
a. Pressurizer Pressure - 3 2 2 1,2,3 22a Not P-11 Da LLJ

b. T.,,g D 1__9 1 9 1
~~. 1. ,~~
~~1 m, 1~~
~~. 7 ]p.~~
~~c. Steam Generator 3/ loop 2/ loop 3/Toop 1, 2 22c os Level P-14 any loop os 9. AUTOMATIC SWITCHOVER TO h~~
~~a CONTAINMENT SUMP~~
a. RWST Level - Low 4 2 3 1,2,3,4 18 f55 COINCIDENT WITH Containment Sump Level - High 4 2 3 1,2,3,4 18 AND Safety Injection (See I above for Safety Injection Requirements) e, b. Automatic Actuation 2 1 .2 1,2,3,4 15 g33
*lI Logic' Eii ba s, . -..m.:.__ -
Tr.) Anf.sn es6mLIf 64LJ okn AlI cd m5y is Al.u W be /c~ P-ll sda, Jdh Djolr.~ w J/<*m h%s Au,.m - l e w r nel Absid TABLE 3.3-3 (Continued) I TABLE NOTATION ! Trip function may be bypassed in this MODE below P-11 (Pressurizer Pressure gg' Block of Safety Injection) setpoint. i
~~'; fun:t hn ::j 5: tjp::::d in thh "00E L:h. '1: (T avg Chd ;f hf:t,,~~
~~gggInjett' n) ::t;;i S j The channel (s) associated with the protective functions derived frcm the~~
~~,out of service Reactor Coolant Loop shall be placed in the tripped mode.~~
The provisions of Specification 3.0.4 are not applicable. l 1 1 ACTION STATEMENTS ACTION 15 - With the number of OPERABLE Channels one less an the Total ! Number of Channels, be in HOT STANDBY withing hours and in 1 COL:, SHUTDOWN within the following 30 hours; however, one ! channel may be bypassed for up to hours for surveillance R$$ l testing per Specification 4.3.2.1.1 provided the other channel is OPERABLE. E
-)
~~DelaU i ACTION 16 - -With the~~
~~-"-'.:r o,,f OPEP^.9LE Ch:rn: h n: h:: th:r th: Tet:1 um-w.. ., ru 1~~
~~__ ....4. .......,4 m. 43 m._,....,_ m< ,~~
'3-3b:5 rE[iEE[CHNkUUk'Ci50$ITE55,3bvidib'555 i :;;rdh :h:rn:1 'h ph::d '- th: trip;:d ::ndit hr -1 5:;r. "d+- . ACTION 17 - -With : ch:nnd :::::hted .;ith :r :;;r: tin; S:;ni :;;rdh, ! . retter: th: 4- perdh charn:1 t: OPEP^.BLE :t:te: cith'- 2 heur: " 9 "f # 4 a f'-cr' b:
rev:r, at h::t MOT SML'TOOW" "ith'- th: f:! hu'n; 12 heur:; n channel ::: : ht:d 'ith er ;;r: tin; M:; ::3 5: byp::::d Sped #kat*0 f^r up t: 2 heur: fer curv:i' h n:: t::t h; p:r 1.2.2.1.1. R' ACTION 18 - With the number of OPERABLE Channels one less than the Total Number of Channels, operation may proceed provided the inoperable channel is placed in the bypassed condit and the. Minimum ChannelsOPERABLErequirementisd----gdwith' rih;;;; .- y one additional channel may be bypassed for up to hours for R: surveillance testing per Specification 4.3.2.1.1. ACTION 19 - With less than the Minimum Channels OPERABLE, operation may continue provided the containment ventilation isolation valves - are maintained closed. ' ACTION 20 - With the number of OPERABLE Channels one less than the Total
~~' Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or be in at least HOT STANDBY within the ,~~
next 6 hours and in COLD SHUTDOWN within the following 30 hours. SEQUOYAH - UNIT 2 3/4 3-22 Amendment No. 55 December 31, 1987
Iruer f [' l 7';//< 3.3 -3 l
~~/fCTIOM /7 - With the number of OPERABLE Channels one less than the Total ,~~
~~Number of Channels, STARTUP and/or POWER OPERATION may proceed ' provided the following conditions are satisfied:~~
~~a. The inoperable channel is placed in the tripped condition within d hourJ. )~~
'l
~~b. The Minimum Channels OPERABLE requirements is met; however, i f4e /sepvwee/a channel may be bypassed for up to V hours i for surveillance testing gper. Specification 4.3.2.1.1.~~
~~& oNer s An .,,ls i~~
I l I i l e 5 4 4 i i s 1 1 l 4
4
~~. . TABLE 3.3-3 (Continued)~~
ACTION 21 - With less than the Minimum Number of Channels OPERABLE, declare the associated auxiliary feedwater pump inoperable, and comply with the ACTION requirements of R116 i Specification 3.7.1.2. ACTION 22 With less than the Minimum Number of Channels OPERABLE, declare l the interlock inoperable and verify that all affected channels of the functions listed below are OPERABLE or apply the I appropriate ACTION statement (s) for those functions. Functions to be evaluated are:
~~a. Safety Injection g y,,, %,,,,~~
~~Pressurizer Pressure y,s;,, ,rg t.. S,, ,, ,, 4 f, {~~
~~b. Safety I-jecti^^ IM.L/ i~~
~~. Mig ASteer Li e r i mm, -~~
~~Steer u4,# c+ L4 e I ^'eti^^ jV,s g2g j m.. e.4 ... r$ ~_- 4~~
~~c. Turbine Trip Steam Generator Level High-High {~~
~~Feedwater Isolation Steam Generator Level High-High ; ACTION 23 -~~
~~. With the number of OPERABLE channels one less than the Total Number of Channels, be in at least HOT STANDBY within 6 hours :~~
and in at least HOT SHUTDOWN within the'following 6 hours; however, one channel may be bypassed for up to 2. hours for surveillance testing per Specification 4.3.2.l.1. RSS-ACTION 24 - J With the number of OPERABLE channels one less than the Total i Number of Channels, restore the inoperable channel to OPERABLE ; status within 48 hours or be in at least HOT STANDBY within- ; 6 hours and in at least HOT SHUTDOWN within the following ' 6 hours.
)
ACTION 25 - 1 With the number of OPERABLE channels one less than the Total ' Number of Channels, restore the inoperable channel to OPERABLE status within 48 hours or declare the associated valve . inoperable and take the ACTION required by Specification 3.7.1.5. e 1 ACTroW 36 D gutw n .Tme f ~3 " gggy7p Acrzov at j SEQUOYAH - UNIT 2 3/4 3-23 i Amendment No. 55, 116 November 28, 1989
G Ta6le 3.3-3 INSERT 'T ACTION 36 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
a. The inoperable channel is placed in the tripped condition within 6 hours.

b. For the affected protection set, the Trip Time Delay for one affected steam generator (T S
~~) is adjusted to match the Trip Time Delay for multiple affected steam generators (Tg) within y hours,~~
~~c. The Minimum Channels OPERABLE requirement is mett however, the inoperable channel may be bypassed far up to 4 hours for surveillance testing of oNur channels per Specification 4.3.1.~1.1.~~
ACTION 37 - With the number of OPERABLE channels one less than the Total Number of Channels STARTUP and/or POWER OPERATION may proceed provided that within 4 hours,for the affected protection set, the Trip Time Delays (Ts and Tg)' threshold power level for zero seconds time delay is adjusted to 0 % RTP. . ACTION 38 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided that within 4 hours j for the affected protection set, the Steam Generator Water Level Low Low (EAM) channels trip setpoint is l adjusted to the same value as Steam Generator Water Level - Low Low (Adverse). e
=
~~TABLE 3.3-4~~
~~.m,,~~
~~EO ENGINEERED SAFETY FEATURE ACTUAT10rl SYSTEM INSIRtffENTATION TRIP SETPOINTS x~~
~~. FUNCTIONAL UNIT TRIP SETPOINT ALLOWABLE VAtllES C~~
5
~~1. SAFETY INJECTION, TURBINE TRIP AND FEEDWATER ISOLATION~~
~
a. Manual Initiation Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable Not Applicable I. 6

c. Containment Pressure--High 11 54 psig 11-7 psig I64.1

d. Pressurizer Pressure--Low 21870 psig 1 W psig h Mad

e. recture -dan ; :i --!!? p I
~~-O!'.fer--tir!~~
~~o,,...___ c._ _a :_____u:_t w+-ww- < sw- u,,,w., ay:yi N~~
* ' 7= Ste;_ tin;3__ . Ster r7e : ,1, n;;t5;7 g ,57,;3 ;; . , n;_. t 5 ,
g,: =g n_ T m 8'?? Ceix!&ct eit5- Tavg-- L - L = Tc? ?c ::: *
^p cer c p ..f Tc!! =: * ,...'r, -
~~ng A _- we~~
~~~.__e:-~~
~~c n .w,,.,,..n. . n m a -u v v v- -> rss- sv g ,'_' v,, u= su=~~
~~,,,'g gny su- >sv- < w w wm u rw - ---~~
~~_7 9 w g ,, g _ _ _ _ _ ny ,,,g ,9n, y~~
~~-- _ __ . s. .s s - .,- gg g~~
g,,,, T m. _ _ _a aL__ _ a__ a a_ . _ . _ _ ' _ _ _ _ _ . ;_._ evuu ew t b e rb r a u pg1 b a v'. s y ET ugs y T TL 3 L J 3 I'J I nv enj a c Inw. I I .. s == 1 u 4, n s I ; -. . s . f , 9 m Am. s. ... _.o ..f.
^; ce-resp- '~g'-1"" :;. te !!! 'Y * '- st '
~~af f.m I I r_ 0 . u- f 1, 3,8 f1*.. ,~~
.I f .. I R I, ..I f , .1 1 t, ,s , a T . t. t no t- I t sp*'t ~ avg - 'avti f Sle L: . Pan <. -- Lv.e LG00 psig ste.sm 1ine ^5W,-g'2' l iine pressure (Wol.1) iires,pwI>comu. e (Wol. I) s t__L-----
~~1 TABLE 3.3-4 (Continued) a ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION TRIP SETPOINTS~~
$ FUNCTIONAL UNIT TRIP SETPOINT All0WABLE VALUES EE 2. CONTAINMENT SPRAY n> a. Manual Initiation Not Applicable Not Applicable
b. Automatic Actuation Logic Not Applicable Not Applicable

c. Containment Pressure--High-High 12.81 psig 2.9 1,2-97 psig

3. CONTAINMENT ISOLATION

a. Phase "A" Isolation

1. Manual Not Applicable w Not Applicable E>
'" 2. From Safety Injection Not Applicable Automatic Actuation logic Not Applicable
b. Phase "B" Isolation

1. Manue.2 Not Applicable Not Applicable

2. Automatic Actuation Logic Not Applicable Not Applicable

3. Con ta i nmen t . P res s u re--H i gh-li i gh 1.9 1 2 81 psig 13 psig

c. Containment Ventilation isolation
~~, 1. Manual Not Applicable Not Applicoble~~
2. Automatic Isolation Logic Not Applicable Not Applicable w % 4 v- -- c- > w c v =- , ___.___._m -_.n. -:ta _s.._ _ _ .a w-e.,._-_ .. m,.,-_ _ . . - _ _ . . - _ , m.
m TABLE 3.3-4 (Continued) p ENGINEEREb SAFETY FEATURE ACTUATION SYSTEM INSTRt#1ENTATION TRIP SET C o FUNCTIONAL UNIT
-< TRIP SETPOINT All0WABLE VALUES z <8.5 x 10 -3 pCi/cc , 3. Containment Gas Monitor- -3 pCi/cc '
~~c Radioactivity-High -~~
~<8.5 x 10 5-s 4. Containment Purge Air Exhaust -3 pCi/cc -3 m 18.5 x 10 18.5 x 10 pCi/cc Monitor Radioactivity-High -5 pCi/cc <1.5 x 10 -5 pCi/cc
5. Containment Particulate Activity-High 11 5 x 10 -

4. STEAM LINE ISOLATION

a. Manual Not Applicable Not Applicable

b. Automatic Actuation Logic Not Applicable w Not Applicable l w c. Containment Pressure--High-High 2.9 a -<2.81 psig _e-97
< psig w "3 ;
StOZ El= I" I= StOZ ' #700-- ' " f"."" t i Or ^ f IOd !! '^ #'*-tya ,4-f 4 4 ugg cejej e t jth y Te!! = : a ap cerre:gg= _s g, ' Or Ste r tir P- cture ayg_t~_ty
~~7e!ime: .a A- re.~~
~~7, ,.~a ; $~~
~~,_; ge ,,, = y y=y i ge =~~
~~te ,,, = 7 ,u , , s g ? _ ,~~~
'! ~ M t:n - '" :-' 2'" bet . u - '* d 2'" !:: ' ! M M thr : ^p W ther : ^p i rrecti ;
i
^
rrecti ; 'ircr!y te : 'i r r!y ' - = *p er--- -~+ cerrer--~* 8 ; te -Q5'# cf ful' ;tes:- 11'" of fe!' ster '?= '! = at * ' ir " at fe?' 'c?* T stAnor T st3cor i ofa ), Sko- f: Arn.,.< ~ leu ~**9~~'~ ' B V9 ~~ >
>600 psig steam >580 psig steam '
~~E$ Tine pressure (M4 d y n :s 5.~~
~~e. Way L Jke L.'.u Aann Raf -//y{ g ,,,, , ,. (,gx 2y Tine pressure (M4 0~~
$ ,o7, g,,; (ys 2) i e TUR8INE TRIP AND FEEDWATER ISOLATION 81% . El.2 %
Uz a. Steam Generator Water level-- <M7r of narrow range j en High-High Instrument span each stear,i
<7t2 of narrow range instrument span each steam enn generator generator N
~~b. Automatic Actuation Logic N.A. N.A.~~
~~R55 l~~
~~,,e .e w ,w+~~
w ~ w. . , . . _ ., .,y. , , , _ _ ,. . _ , _ _ , . _ _ _ _ . _ _ _ _ _ _ _ _ _ _ _ _ . , . , _ , _ _ _ , . ., _ _ _ . , y_ , , , , . .__,__,y _ m_,m_. _ , ._ , , , , _ _ , _ , _ _ _ _ , , _ _ _ _ , , , _ _ _ _ _ _ . _ , , _ , _ , ,
TABLE 3.3-4 (Continued) N S ENGINEERED SAFETY FEATURE ACTUATI0ff SYSTEM Ill5TRIRTFNTATION TRIP SETPOINIS 5! f FUNCTIONAL UNIT TRII' SETP0lNT All0WA15LE VALUES E 6. AUXILIARY FEEDWATER Z m a. Manual Not Applicable Not Applicable
b. Automatic Actuation Logic Not Applicable Not Applicabic

c. Main Steam Generator fox f Water Level-low-low '
~~'er 3 7._ 7 ~;r s m e r - ,- _ - >. 7, Testr --t sp esc 5 Tstr= mt:~~
~~erb-s t e r ;;e-~ > ter tie 7 r-et -~~
~~d. 5.1. See 1 above (all SI Setpoints)~~
$ e. Station Blackout 0 volts with a 5.0 second 0 volts with a 5.0 1 1.0 second y time delay time delay $ f. Trip of Main Feedwater N.A. N.A.
~~Pumps~~
~~g. Auxiliary Feedwater Suction 12 psig (motor driven ptmp) i I psig'(motor driven pimqi)~~
~~Pressure-Low 1 13.9 psig (turbine driven 1 12 (turbine driven pimp) lR84 (ptunp)~~
h. Auxiliary Feedwater Suction ' 4 second, (n.otor <! riven pump) 4 seconel, !O.4 seconels (=>t or Transfer Time Delays ifriven ptm.p) 22 y5 5.5 seco uis (tun bine airiven ptimp) 5.5 set.onil . !O.55 secon I, E l l #>
~~7 ,K ( tin hine it. ivi n pimgi) N,S "E~~
~~;;;*z M>M o, =~~
MF* 0) U e . e _ . ._ . _ , _ - .~ , , _ . , _ _ _ _ _ _ . - -_. _,____________ .___ _ .________ -._ _ __ _ _ . _ . . _ _ . _ _ _ . _ _ . _ _ _ _ _ . _ _ . _ . _ _ _ _
n+. -.. . .x....a+..s _...a - - - . . . -
~~...a - . ,-~~
~~k %~~
~~4 T i 7 .~~
M N i h 4 7 . ;D k n l
~~!;3~ E ; w' e i~~
~~. m .: : "r ty vQ j~~
{ 'T 4 ' ik I M gh 1% N g( 4
~~s hw M ng k k nt 3~~
~~1 m h i 9 g ( 8 4.. 1 4, s 4 - w~~
~~>i d~~
at c w dw ! s i l l l I , , n 9 l d h h ,i 0 !
~~- a, t 't 34 1, e p ;~~
~~F i 1(~~
~~- e ,~~
~~%44 s e %"~~
i
~~, N5 "@ \ 1 d S ,v '* l ? [a .~~
~~b "' s s ee f~~
.4 A aj $ 3 i 1s - *4 c Vt $ de h" lJ g
~~k> -~~
~~4. w 41 w y, .~~
)
t . g n E 4U 1 T 3 b h w is ,$ ;- '* g s i V' ) 14 %
~~8) -~~
Ti
~~3. , 24 ;~~
ce t i gt t ! 1 ha gQ M1ie s d ,,4 13 2 v v i b
*g4 '* i I4 4
~~si 9w 1 64 1 @4 1t~~
~ % 4 i s uiNe u, 't d 4 i k' " * ~ ' ), xb i j] q & 3 g g l w s - .? l 4
~~t, .~~
~~- - .r ,~~
~~l b h - 1 I i l ) l W ht k ! k gk kI~~
* %4 w
Ns*i h a i s 4 N f \s M my w b f l a r? e 44 i w r5 9 ' g i 4
~~<g 1 b ik j .i T~~
~~m m~~
~~k. ,~~
t e r 3 Q s 4 13 11 ! 9 at .( =t ; ? 4a'y s oa LT A\ s G - yf , Qt Ag % .3 a l; 4
~
~~aa q3 3~~
~~.t v 1 6"0 q4d 4d, 4 -~~
~~t 1{d 1 Ij W 6^$ s u,gI 4 - id~~
~
~~i Is1 ia 4 et s%'*40 4e < 4 F % ti t 114 $ ny - td~~
~
m TABLE 3.3-4 (Continued) E g ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRtRIENTATION TRIP SETPOINTS s 7 FUNCTIONAL UNIT TRIP SETPOINI ALLOWABLE VAlllES E /.7 LOSS OF POWER
~~-4 m a. 6.9 kv Shutdown Board Undervoltage loss of Voltage~~
1. Start of Diesel Generators nts
~~.a. Nominal Voltage Setpoint 4860 volts 4860 volts 197.2 volts~~
b. Relay Response Time for 0 volts with a 1.5 second 0 volts with 4. 1.5 10.5 Loss of Voltage time delay second time delay

2. Load Shedding

a. Nominal Voltage Setpoint 4860 volts 4860 volts 197.2 volts w b. Relay Response Time for D 0 volts with a 5.0 second 0 volts with a 5.0 il.0 Loss of Voltage time delay second time' delay
~~. Y a,~~
b. 6.9 kv Shutdown Board-Degraded Voltage

1. Voltage Sensors 6560 volts 6560 volts t Si volts

2. Diesel Generator Start and Load Shed Timer 300 seconds 300 set.ond:, 1 30 seconds "25

3. SI/ Degraded Voltage Logic yp Enable Timer 10 seconds 10 seconds i ti.5 second:,
NR 8. ENGINEERED SAFETY FEATURE h ACTUATION SYSTEM INTERLOCKS az a. Pressurizer Pressure sf 7r.1
~? Manual Block of Safety Injection P-11 <1970 psig e 1+W& psig C kN CC ChCC Cat CC s tC bC % CmCCbCN N^dI .] kbC stC~' If b ' ' ", 7"j kl '! " M C : t ge. Rl8
~~is:~~
e
~~. . . .- - - . - - - . . . - -._ ______.__________._____s~~
TABLE 3.3-4 (Coatinned) ENGINEERED SAFETY FEATURE ACIUATIOrt SYSTEf1 Ill5TRUMElliAlloil TRIP SETrolig e f FUNCTIONAL UNIT IRIP 51IP0lly AlIOWAHIE VAlIIFS E 8. ENGINEERED SAFETY FEATURE ACTUATION Z SYSTEM INTERLOCKS (Continued) ro
~~b. ikll/.u i Drerente " ra? S!xk cf S:fety~~
~~[. Inj e tier P-!? ; 540 r ;5c*F~~
c. De/elad

"

r :! S!= E ef Safety Injectien, Ste:- Li r Iseletien, S?eck Sic--
w Sg -' 5'O r 152S*r g - w d. Steam Generator Level 4 Turbine Trip, Feedwater Isolation P-14 (See 5. above)
9. AUTOMATIC SWITCH 0VER TO g33 CONTAINMENT SUMP

a. RWST Level - Low 130" frem tant base 13ts" 1 4" 4 os tank base COINCIDENT WITH Containment Sump Level - Higli 30" above elev. 660' 30" 1 2.5" above elev. 660' AND k
~~Safety Injection -(See 1 above for all Saf ety Injection St.- points / Allows,le Valv .~~
~~b. ~ Automatic Actuation Logic fl. A. 11. A. g33 g~~
~~9+ t -~~
~ 2*
O g *
"ers s WJ der f aft lEgel 5** L f$= LOM bhsee h/eJJa***** hee! ** 's $ 5 S O .tes Y ' ~ ,,,a- x s r.,a,. - A4L .2: ~C u. &f udl'o</ in fie rs d - /g w . L ,/le d A,4 h. Jdw L.- Aa , A= f -
My b zy so .su-4 m..- i-. ---
T ABLE 3.3-5 (Continued) l i i 1 E!GINEERED SAFETY FEATURES RESPONSE TIMES l INITIATING SIGNAL AND FUNCTION RESPONSE TIME IN SECON05 i i
3. Pressuri er Pressure-Low I R47 )

a. Safety Injection (ECCS) 3 32.0(lbS.M"

b. Reactor Trip (from SI) < 3.0

c. Feedwater Isolation 8. 0 N) ,

d. Containment Isolation-Phase "A"(3) _18.0(3) R96 l

e. Containment Ventilation Isolation < 5. 5 8)(13)

f. Auxiliary Feedwater Pumps 60(11) lR68

g. Essential Raw Cooling Water System 65.0(8)/75.00)

h. Emergency Gas Treatment System <28.0(8)
~~A /. 4 4"<4erertie / 'e::ure n-t :::r ste:- L!ner u5 3 m~~
.e-m+u.. ..,., i m,4 m. . +. 4. m m. R' . t, r e. r. .e ,s a.. o . n ( 7 ) / n o. . n. (l ) . b . II - .. ~ , r m as,e+ - t,mt.+4mm o n(2) !
I.
~~. rS.- . .. N.. ...'3.... . ..~~
~~__ng.,, usn(3) $ (8),,e.n(9) _. .n. ..~~
~
~~c. COI$ i"'^nt NI.$.):$ien : l:tien~~
+
~~HOI I??)i:N):~~
~~^ux ' f ery F: d::t r 6; ?:~~
4
~~;SO(11)~~
~~R68~~
g. E;;^.ti"' b COeling W:t : Sy;t r 155.0(OI'?5.0(9) k T Emer;ency C:: re:t- nt Sy.ter -

18.0(9) .
L
.. .. .. .g w . .w .YW t 2i +h T -Ia. .I at a .s .
~~c.yg~~
~~. . . , , '...e...... ; _ { {,' , _v.... _ , r e e n, a~~
~~_ . n. . n.(7) /e n. n. (1)~~
~~. R47~~
b. "ee ter T-i; (fr:r.SI) c 5. 0 ~

c. F i::t:r ! 01 ti n 10.0(2)

a. . .. ... ....
~~rm. 4. .. ... .v,.3. ...4..._nu... uan(3) I, ' _.n n.(8) ,3 n. . n.( 9 )~~
~~e. C^-t:i-- nt "^ntil:tien !:01: tier Net ^^-11:21~~
~~^ex 'f:ry Fe^i;:ter "" ?:~~
~~4 R68 r . ,3,. 150(11) g, 4 g on g g 4 ; u... ey:te -g7,n(8),.r.,g(9) , , k Ce.n. ..- t 4. m.~~
~~. . .m. T. ,. a.1. 6 4. m. m. . 27n n E r;en:y C:: r::t :nt Sy:ter T~~
~~_ 0. M9) s , e ,. r . a . a a - . u - 8 9~~
o. .s lu., L:m .ruh% 48.0 SEQUOYAH - UNIT 2 3/4 3-30 Amendment No. 47, 68, 96 March 13. 1989
~~, 4.n. , -~~
~~l l TABLE 3.3-5 (Continued) 1 ENGINEERED SAFETY FEATURES RESPONSE TIMES i l INITIATING SIGNAL. M!D FUNCTION RESPONSE TIME IN SECONDS I~~
~~6. -St::: "1:1 '- T= St::: Lin : uie !~~
~~C:it:it zt e644 Steam Line Pressure-Low '~~
~~a. Safety Injection (ECCS) 1 28.0(7)/28.0(1)~~
~~R4 i~~
b. Reactor Trip (from SI) < 3.0 '

c. Feedwater Isolation 8.0(2) {

d. Containment Isolation-Phase "A"(3) [18.0(8)/28.0(9) j

e. Containment Ventilation Isolation Not Applicable f.
~~R6 l Auxiliary Feedwater Pumps <60(11) j~~
g. Essential Raw Cooling Water System 65.0(8)/75.0(9) '

h. Steam Line Isolation < 8.0

1. Emergency Gas Treatment System '
~~[38.0(9)~~
7. Containment Pressure--High-High R$1

a. Containment Spray < 208 59)
~
~~b. Containment Isolation-Phase "B"(12) R73~~
~~$ 65(8)/75(9) )~~
c. Steam Line Isolation '$ 7.0 ,

d. Containment Air Return Fan > 540. 0 and . < 660 R5 i

8. Steam Generator Water Levol--High-High- ;

a. Turbine Trip < 2.5

b. Feedwater Isolation [11.0(2)

9. Main Steam Generator Water Level -
~~Low-Low~~
~~a. Motor-driven Auxiliary ~~~
~~<60.0(14) :~~
~~Feedwater PumpsC4) g~~
~~b. ' Turbine-driven Auxiliary < 60.0~~
~~- t FeedwaterPumps(5)(11)~~
September 9, 1988 SEQUOYAH - UNIT 2 3/4 3-31 Amendment No. EE, K,584 73-
) TABLE 3,3-5 (Continued) TABLE NOTATION (10) The response time for loss of voltage is measured from the time voltage is lost until the time full voltage is restored by the diesel. The R68 response time for degraded voltage is measured from the time the load shedding signal is generated, either from the degraded voltaae or the SI enable timer, to the time full voltage is restored by the diesel. The response time of the timers is covered by the requirements on their setpoints. (11) The provisions of Specification 4.0.4 are not applicable for entry into R68 MODE 3 for the turbine-driven Auxiliary Feedwater Pump. (12) The following valves are exceptions to the response times shown in the u ble and will have the values listed in seconds for the initiating si1nals and the fenction indicated: R73 Valves: 10 kesponse FCV-67-89, times: 7.b, -90,75-10h)/85{g) Valve: FCV-70-141 Response times: 7.b, 70(8)/80(9) (13) Containment purge valves only. Containment radiation monitor valves have a response time of 6.5 seconds or less. Ri D* 000 nol iruk4 'Gy ~% Aloys. Aeyo.,a };ma irode/ ine / A -fle -/wm.u, E*j4 2I poeai jaf a />n tof6,Q sold.tbd pro lu No* uit.,s h, so,/ pe L/toa, A vhu (cy f s,,o/ iali.dy p-p). TAh n/M -14 ny..>. 4 <a nuussy f 7dEhn4L POWH in caos o/ 50 % ATA 1
~~. j SEQUOYAH - UNIT 2 3/4 3-33a Amendment No. 18, 68, 73, 96 March 13. 1989 i~~
TABLE 4.3-2 M S ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRt#fENTATION S SURVEILLANCE REQDTREMENIS
c. CilANNEL MODES FOR WIIICil 5 CHANNEL CilANNEL FUNCTIONAL SilRVEILLAtlCE IS
~~-4 FUNCTIONAL UigIT CHECK =~~
~~CALIBRATION TEST REQUIRED~~
1. SAFETY INJECTION AND FEEDWATER ISOLATION

a. Manual Initiation N.A. N.A. R 1, 2, 3, 4 R39

b. Automatic Actuation. Logic N.A. N.A. M(1) 1, 2, 3, 4

c. Containment Pressure-High S R q 1, 2, 3
$ d. Pressurizer Pressure--Low 5 R q 1,2,3 w DrLid g, e. Di"creti:1 Per:t-- S "
~~3, ?, ?~~
g "etc:- Ster Lix:- "?p S* r rw 4. % ster 3 e 1, 7, 3 Li-- - ugg c 5_.5c ,t w5th Q r _ _ e_ _ i_. _. .....
_ c. . - - , i :. . m A A L;ae Pressure--Low _f f Q bh3
2. CONTAINMENT SPRAY Y a. Manual Initiation N. A. N.A. R 1, 2, 3, 4 g)9 I* b.. Automatic Actuation Logic N.A. N.A. 11( 1 ) ] , 2, 3, 4
~~[" c. Containment Pressure--High-High S R I , 2, 3 4 Q~~
?
~~w W w~~
~~% .w - .e- w-, ,.s-r- -#-, ,,,-+.3 . -,w,-- v, .. ,- . _--_ _ _ _ _ _- -----.-.-m-_._.~~
_. _ - _ _ .w_m_m._m ____.m __. __., ._._m____.,_, . _ m . _m___ ..m= _ . _ _ ___m_.,m____m _ , _.,_ ___ m.
TABLE 4.3-2 (Continued) y ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION g SURVLILLANCE REQUIREMENTS 2-CHANNEL MODES FOR MIICH CHANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS FUNCTIONAL UNIT CHECK CALIBRATION TEST REQUIRED z Z 4) Containment Purge Air 5 R M 1, 2, 3, 4 m Exhaust Monitor Radio-activity-High
5) Containment Particulate 5 R M 1,.2, 3, 4 Activity-High

4. STEA4 LINE ISOLATION

a. Manual N.A. N.A. R 1, 2, 3 l R39
{ b. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 Y c. Containment Pressure-- S R Q 1, 2, 3 ci High-High l
~~d. Ster r!r '- Tr Ste r S R Q 1,2,3~~
~~'fr:- "if Siri t i riih T -- t r L e r Steam Line e, h j"Y's~ & ,A n b4-Ajk S A 0 3~~
5. TURBINE TRIP AND FEEDWATER ISOLATION ed" a. Steam Generator Water 5 R Q 1, 2, 3

{ Level--High-High 9 53 0
e
~~b. Automatic Actuation Logic N.A. N.A. M(1) 1, 2, 3 m 6. AUXILIARY FEEDWATER 1~~
.?
~~g a. Manual N.A. N.A. R 1, 2, 3~~
= R39 l b. Automatic Actuation Logic N.A. N.A. M(1) I, 2, 3 E
1
TABLE 4.3-2 (Continuet!) E ENGINEERED SAFETY FFATUllE ACTUATION SYSTEM INSTRIXIENTATION S SURVEILLANCE REQUIREMENIS E -
CilANNEL F100ES FOR WIIICil CHANNEL CilANNEL FUNCTIONAL SURVEILLANCE IS E FUNCTIONAL UNIT CilECK CALIBRATION TEST REQUIRED
*- c. Main Steam Generator Water
~~53"+ i,Q~~
~~" S 1, 2, 3 Level-Low-Low~~
d. S.I. See 1 above (all SI surveillance requirements)

e. Station Blackout N.A. R N.A. 1, 2, 3

f. Trip of Main Feedwater N.A. N.A. R 1, 2 Pumps

g. Auxiliary Feedwater Suction N.A. R H 1,2,3 w
~~Pressure-Low~~
~~h. Auxiliary Feedwater Suction M.A. R N.A. 1, 2, 3 Y~~
~~w Transfer Time Delays~~
7. LOSS OF POWER

a. 6.9 kv Shutdown Board -
~~Loss of Voltage~~
1. Start Diesel Generators S R M 1,2,3,4

2. Load Shedding S R N.A. I, 2, 3, 4 R18 EP -*W TE: hit t + ica! r pci'ic: tie- it t e be 47 !r:- t ' '" : ~,; t '~ t ' ' "; 'e i ' '-

: ; ' '- ! ' - ' i ; ' " ' ":
5
~~@3 g '~~
ag
~"
NZ
~~.U e. <n .~~
_ - - ~ . . _ ._ . . . . . , - . . - , _ . - ~ . .,. , - - - . . - - , . . . , _ . . . -.
d,; k ,
~~+ Q @~~
~~o t@~~
~~% ss e o t~~
i
~~s. s t I4 Q q 0 e 9 et M~~
4 M A ,4 jd * *
~~w 4~~
~~99 M A N~~
~~& s Nr D 9 9 'd "~~
l 1 N i e Q 4 4 9 ! i W7 T T d' i s d,4't i$ q,N 1 4 %: a ye tE s
~~y. . ~1 r- 9x~~
~~&5 k b Q .~~
s - . - u l C _ _ _ _______ _ -_--_ __-_____-.----- ___- ----- - - - - - - - - - - - - - - - - - '
j o, TABLE 4.3-2 (Continued) 13 I gg ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUf4ENTATIDN g SURVEILLANCE REQUIREMENIS x '
CHANNEL MODES FOR WilICII c- CilANNEL CHANNEL FUNCTIONAL SURVEILLANCE IS 25 FUNCTIONAL UNIT CliECK CALIBRATION TEST REQUIRED
~~-e //b. 6.9 kv Shutdown Board -~~
~~Degraded Voltage~~
1. Voltage sensors S R M 1, 2, 3, 4 Rl8

2. Diesel Generators N.A. R N.A. 1, 2, 3, 4
~~' Start and Load Shedding Timer~~
~~3. SI/ Degraded Voltage N.A. R N.A. 1, 2, 3, 4~~
~~}{ Logic Timer '? 8. ENGINEERED SAFETY FEATURE ,~~
~~8$ ACTUATION SYSTEM INTERLOCKS~~
a. Pressurizer Pressure, N.A. R(2) N.A. 1, 2, 3 P-11

b. T I 1?
~~"^ "'s',' "^ ,,g,~~
~~c. Steam Generator N. A. R(2) N.A. I, 2 Level, P-14~~
?> ## 9. AUTOMATIC SWITCHOVER TO 8E ea CONTAINMENT SUMP "@ a. RSWT Level - Low 5 R !! I,2,3,4 M '* COINCIDENT WITH - EF Containment Sump Level - High 5 R 11 1,2,3,4 Of AND lEjs Safety Injection. (See.1 above for all Safety Injection Surveillance Requireu.ents) en w
b. Automatic Actuation Logic N.A. II. A. If(I) 1, 2, 3, 4 RSS U N^t"* "is i"Ch^iCOl spCCiYiCOtiO"; i! 10 DC igli.4.,1Cd d T!Cg ibC S !Ori" p f 0 ! N 3 *j !!' I'_ i ' '^ I" ' "'j "lt ," ' ggg m g.
~~- . - - - _- - -. -- -. - ~ - - . . .~~
3/4.3 TNSTRUMENTATION BASES i 3/4.3.1 and 3/4.3.2 REACTOR TRIP AND ENGINEERED SAFETY FEATURE ACTUATION SYSTEM INSTRUMENTATION , The OPERABILITY of the Reactor Trip and Engineered Safety Features Actuation Systems instrumentation and interlocks ensure that 1) the associated action and/or reactor trip will be initiated when the parameter monitored by , each channel or combination thereof reaches its setpoint, 2) the specified ' coincicence logic is maintained, 3) sufficient redundancy is maintained to permit a channel to be out of service for testing or maintenance, and ,
~~4) sufficient system functional capability is available from diverse parameters. !~~
The OPERABILITY of these systems is required to provide the overall reliability, redundancy and diversity assumed available in the facility design for the protection and mitigation of accident and transient conditions. The integrated operation of each of these systems is consistent with the assumptions used in the accident analyses. The surveillance requirements specified for these systems ensure that the overall system functional capability is maintained 1 comparable to the original design standards. The periodic surveillance tests i performed at the minimum frequencies are sufficient to demonstrate this . capability. . The Engineered Safety Feature Actuation System interlocks perform the functions indicated below on increasing the required parameter, consistent with the setpoints listed in Table 3.3-4: P-11 Defeats the manual block of safety injection actuation on low pressurizer pressure. ]
]
~~D-12 D: feet: the e.:nua?-b'cek c' ::f:ty #nj:: tion :: tutti:n : Si;h :t::: J~~
, 'ine ee cod 'ee etter '#-- pr:::ur:. -
P-34 Trip of all feedwater pumps, turbine trip, closure of feedwater isolation valves and inhibits feedwater control valve modulation. ) On decreasing the required parameter the opposite function is performed at reset setpoints, eith the - reptier Of D-12 :: neted 5 !:r-c.19 ro.wi. - .n.1 k1.,9 ., e.r.+u 4.4.,+4 . .,+n.+4 . . k< w .+.. 34..
~~'Ibb bbd Ibb,,.m~~
_w< ktbbb #b0 "hbbbbr0. hbbbb:kkbbbI#n0 i050tkbbOr s ....- i,,.,+. .+..., si ,&. << . ...m.... . . . . . . . . . . D . "h # h^ "Ot^d i
O
~~) . l i~~
l SEQUOYAH - UNIT 2 B 3/4 3-1 t 4 a .,.n. v. + - , - - _.--,__,-,.-c-. _ ,-- - . - - , - - - v - - . - - - - - - 3
~~-l ENCLOSURE 2 PROPOSED TECHNICAL SPECIFICATION CHANGE ,~~
I SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-27) I j DESCRIPTION AND JUSTIFICATION FOR ; REVISING SPECIFICATIONS-2.2.1,-3/4 3.1.1, and 3/4 3.2.1 P V 5 1 i i 6 1 l
l 1
-\
ENCLOSURE 2 4 Description of Chante l , Tennessee Valley Authority proposes to modify the Sequoyah Nuclear Plant l (SQN) Units 1 and 2 technical specifications (TSs) to revise the definitions section and Specifications 2.2.1, 3/4.3.1.1, 3/4.3.2.1, and ; their associated bases to reflect reactor protection system (RPS) upgrades i and enhancements to be implemented during each unit's respective Cycle 4 1 refueling outage. Specifically, the following changes are proposed:
~~A definition for a digital channel function test is added.~~
The allowabla values of Tables 2.2-1 and 3.3-4 are revised to reflect rack drift allowances associated with the Eagle 21 digital process 1 protection system. i The low-low steam generator water level entries of Tables 2.2-1, j 3.3-1, 3.3-2, 4.3-1, 3.3-3, 3.3-4, 3.2-5, and 4.3-2 are revised to ; reflect the incorporation of the environmental allowance modifier ; (EAM) and trip time delay (TTD) features. l
The steam /feedwater flow mismatch and low steam generator water level I reactor trip (Tables 2.2-1, 3.3-1, 3.3-2, and 4.3-1) are deleted to reflect the incorporation of a median signal selector (MSS) that separates the control and protection signals for steam generator water i levels. l f The overtemperature and overpower delta-T (differential temperature) entries of Tables 2.2-1 and 3.3-2 are revised to reflect the l elimination of the reactor coolant system (RCS) resistance temperature '
detector (RTD) bypass manifold (Final Safety Analysis Report (FSAR) ! Figure 5.1-1). i l A new steamline break (SLB) protection logic is implemented that results in the deletion of the high-differential pressure between steamline signals, the revision of the high-steam flow coincidence l signal so that low steamline pressure alone initiates the { corresponding engineered safety feature, and the addition of a high negative.steamline pressure rate actuation for steamline isolation (Tables 2.3-3, 3.3-4, 3.3-5, and 4.3-2). Actions 2.b and 6.b of Table 3.3-1 Actions 15, 16, 17, 18, 21, and 23 ! of Table 3.3-3, and the channel functional test intervals of Table 1 4.3-2 are revised to implement the Westinghouse Owners Group (WOG) I Technical Specification Optimisation Program (TOPS) engineered safety. ] features actuation system enhancen:ents of Westinghouse Electric l Corporation WCAP-10271, Supplement 2. Outdated footnotes and unused action statements are deleted from the reactor protection tables. i
~~. 1 \~~
l 1 1
Reason for Change The changes described above represent.a compilation of modifications and enhancements to the RPS that will improve that system's reliability and availability. The Eagle 21 upgrade will replace the Foxboro H-line analog process protection racks with digital technology equipment. Use of this digital system avoids most of the drift problems associated with the analog process instrumentation, and is the most efficient means to implement the RCS RTD bypass elimination package. The Eagle 21 system also has improved test features that include automatic surveillance testing, self-calibration of analog circuits, and self-diagnosis of system troubles. The EAM/TTD features for steam generator water level are being implemented to reduce the incidence of unnecessary feedwater-related reactor trips. The EAM/TTD concepts were developed by the WOG Trip Reduction Assessment Program (TRAP). Similarly, the deletion of the steam flow /f eed flow mismatch reactor trip, by implenenting the median signal se16ctor (MSS) feature, will also reduce the potential for unnecessary feedwater-related trips. RCS RTD bypass elimination provides benefits in three primary areast reduced radiation exposure, improved availability, and reduced maintenance. The system piping within the RCS loop compartment creates a significant source of radiation exposure during maintenance. As with all piping systems, there is also associated operability testing, periodic maintenance, and unexpected outages. The installation of fast response RTDs in the hot and cold legs of the RCS and the processing of the corresponding signals by the Eagle 21 digital system provide the necessary process signals for the RPS. The new SLB protection logic is implemented to acconinodate the removal of the steam flow signals and the comparison of steamline pressures. These ESF signals have historically been the source of inadvertent actuations. The new protection logic utilises low pressure and high negative rates of change setpoints in each steamline, without comparison between steamlines, to initiate mitigating actions for SLB scenarios. The incorporation of the WOG TOPS enhancements is done to reduce surveillance testing impacts on the plant. The enhancements provide increased maintenance and test times and less frequent surveillance.. These changes are intended to reduce the potential for advance spurious actuations. The corresponding bases revisions are an administrative change to provide consistency and clarity. Similarly, the deletion of unused action statements and outdated footnotes also provides clarity by removing extraneous text from the TSs. The addition to the definition section is also administrative in nature, making the channel functional test definition complete for the Eagle 21 RPS configuration.
Justification for Change The changes to the allowable values to Tables 2.2-1 and 3.3-4 are supported by Revisics 4 of Westinghouse WA? 11239, "Setpoint Methodology for Protection Systems." The methodology used for determining the allowable values for Eagle 21 is the same as that used for the Foxboro B-line analog process racks. The changes to the allowable values are the ! result of changes in rack calibration and measurement and equipment accuracies. The elimination of the analog rack comparator also affects the allowable values. TVA has used the rack improvements for reserve margin between the safety analysis limit and the allowable value for all channels except where the channel act.uracy improvements were used to provide additional operating margin (i.e., steam generator level trips). This allows the nominal trip setpoints to remain unchanged for most functions. s The EAM modification to the steam generator low-low level reactor trip setpoint consists of a means to detect an adverse environment inside [ containment (e.g., elevated containment pressure), which is used to enable a steam generator low-low level reactor trip setpoint containing an error , allowance to account for the ef fects of adverse environment upon level ' measuring instrumentation. The modification would permit the use of a lower steam generator low-low water level reactor trip setpoint during
normal operating conditions (i.e., a trip setpoint that does not contain the full environmental error allowance). NRC review and approval of the conceptual design of the EAM enhancement are documented in Westinghouse
.WCAP-11342-P-A (Revision 1), " Modification of the Steam Generator Low-Low Level Trip Setpoint to Reduce Feedwater Related Trips." The SQN EAM feature is consistent with that described in Westinghouse WCAP-11342-P-A (Revision 1) and was determined to be acceptable based on the TSAR '
reanalyses presented in Enclosure 4 The TTD modification is designed to impose predetermined delays upon the reactor trip and auxiliary feedwater system actuation during ' less-than-limiting steam generator water level transients. Restoration of water level to above the low-low level trip setpoint, either by manual control or control system level stabilization, during the trip delay period will avoid an unnecessary reactor trip. A series of applicable accident analyses.are used to confira the acceptability of the trip delays with respect to standard review plan criteria. These analyses are presented in Enclosure 4. NRC review and approval of the conceptual design of the TTD enhancement are documented in Westinghouse WCAP-11325-P-A (Revision 1), " Steam Generator Low Water Level Protection System Modifications to Reduce Feedwater-Related Trips." The technical justification for the deletion of steam flow /feedwater flow ' mismatch reactor trip by implementing the MSS feature is contained in Westinghouse WCAP-12417. " Median Signal Selector (MSS)." Because the justification contains proprietary information. Westinghouse WCAP-12417 and its associated affidavit and application for withholding proprietary information from public disclostre are being transmitted under separate cover letter. e
~~,,.,,,._m.. ,~.y,,,,,,y... ,.,ry. ,, .,,._ y~~
l 1 i i i The elimination of the RCS RTD bypass manifold at SQN is consistent with ! the RTD bypass manifold elimination performed at Watts Bar Nuclear Plant. l The ability to obtain narrow range RCS temperature measurements is ' i ~ maintained by installing fast response RIDS into the hot and cold leg penetrations where the bypass manifolds are currently attached. The cold ! leg temperature is utilized directly, while the three hot leg RTDs are electronically averaged by the Eagle 21 system to account for temperature l streaming. This averaged value .is used as the RCS hot leg temperature. l The overtemperature and overpower delta-T changes made are consistent with ; the revised Eagle 21 setpoint methodology (Westinghouse WCAP 11239 Revision 4), and supported by the FSAR reanalyses provided in Enclosure 4. The new SLB protection is supported by the non-loss of coolant accident ) , (LOCA) analyses that model the protection system logic that leads to j safety injection, feodwater isolation, and steamline isolation when i analyzing excessive cooldown events. The old and new SLB protection j i differ in the logic that leads to the actuation of these safety . functions. In order to support the upgrade of the SLB protection from old to new, Westinghouse analyzed and evaluated the impact of this change on the following licensing basis non-LOCA transients major rupture of a main I feedwater pipe (FSAR Section 15.4.2.2), depressurization of main steam ! system (FSAR Section 15.2.11), major rupture of a main steamline (FSAR i Section 15.4.2.1), and SLB mass / energy release inside containment (FSAR I Section 6.2.1.3.10). The FSAR reanalyses in Enclosure 4 demonstrate that I the new SLB protection logic adequately detects the postulated secondary I system faults for the initiation of protective actions. The incorporation of the RPS testing enhancements is supported by the NRC = review and approval of the justification for these changes documented in Westinghouse WCAP 10271-P-A, Supplements 1 and 2, " Evaluation of j Surveillance Frequencies and Out-of-Service Times for the Reactor Protection Instrumentation System," and " Evaluation of Surveillance i Frequencies and Out-of-Service Times for the Engineered Safety Features -
~~Actuation System." The evaluation presented in these WCAPs indicates that j RPS reliability remains high with the proposed changes. Additionally, ]~~
' plant safety is maintained or improved, plant availability is improved, and the burden to the plant caused by TS compliance is significantly reduced. ' ( TVA recognizes that the WCAP-10271 evaluations performed were based on . analog process protection systems, TVA contends that the replacement of { the Foxboro H-line equipment with digital Eagle 21 equipment, which is at " least as reliable as the analog equipment, does no: invalidate the ; conclusions drawn in WCAP-10271 and its supplements. Reliability studies have been performed on both the Eagle 21 process protection system and the Houston Lighting and Power Company (HL&P) (Docket Nos. 50-498 and 50-499) qualified display processing system l (QDPS), which has a high degree of component commonality with Eagle 21. i In addition, actual operating mean time between failure (MTBF) data from the QDPS has shown that the microprocessor-based equipment is exceeding r design goals. The following is more detailed information on these , 4 6 5 [ ~ [
i l reliability studies and actuel operating history and provides evidence that these selected channels are highly reliable. From the results of the reliability / availability assessment of the Eagle 21 and equivalent process protection systems, it has been determined , that the Eagle 21 digital system availability is equivalent to the analog ( system availability even without the incorporation of the fail-safe design F principles, redundancy, functional diversification, automatic surveillance ( testing, and self-calibration and self-diagnostic features of the Eagle 21 " process protection system. If, for example, full credit were to be given to the capabilities that exist for the Eagle 21 self-diagnostic features such as the electrically progransnable read-only memory (EPROM) checksums, random access memory (RAM) checks, math coprocessor checks, and loop cycle p time checks, they would have a definite effect to further improve accepted y system availability. Furthermore, it could be concluded that if credit ' were to be given to the redundancy, automatic surveillance testing, and L self calibration capabilities, there would also be a definite effect on improving accepted system availability. Items such as decreasing the ( impact on system accuracies because of drifting signal loops being ! corrected by the Eagle 21 self-calibration feature or decreasing system L downtime because of the automatic surveillance /self-diagnostic features ( will be minimized. With the man-machine interface (MMI) provided with ; Eagle 21, the amount of technician time required for maintenance and ( troubleshooting of the system will be greatly reduced. By using the MMI, large quantities of the available instrument mechanic and engineering time needed to perform and review the quarterly functional tests can be , eliminated. To strengthen the arguments provided in the analog versus digital reliability study, actual plant operating data was obtained from HMP QDPS, Prior to beginning the evaluation of the data, an evaluation of the design similarities between the Eagle 21 and the QDPS systems was performed. Upon completion of this review, it was determined that the Eagle 21 system is similar in design to the HMP QDPS. (Many of the ( l Eagle 21 components are identical to those used in the QPDS.) A review of the South Texas Project QDPS reliability study was conducted for NRC by the Idaho National Engineering Laboratory. No methodology errors were noted as part of the review. In this study, expected MTBFs were postulated for the digital instrumentation channels using data obtained from the PC board vendors and compared with expected analog channel MTBFs. The HMP QDPS data verified that the actual MTBF for the key ! components in the system far exceeds the postulated expected MTBF for the ' system. These results concur with the higher availability estimate of the implemented digital channel compared with a hypothetical analog ' counterpart. Additionally, the NRC staff concluded ". .
. the QDPS provides a highly reliable system for its application at the South Texas Project and that its assessed reliability is acceptable." Based on these assessments, it was determined that because of the similarity in design between the QDPS and the Eagle 21 process systems, the conclusions reached for the QDPS may be applied to the Eagle 21 system. \
I
I Taking inte account the arguments presented regarding the system design features and the QDPS digital process system data evaluated, the following conclusions can be reached:
1. The results of the availability assessment of the Eagle 21 and the equivalent analog process l protection system demonstrate that the 3 Eagle 21-digital system is at least as reliable as the present analog i' technologies.

2. If a more detailed analysis were to be performed incorporating the automatic surveillance test, redundancy, and self-test, self-calibration, and self-diagnostics features of the Eagle 21 system, it would have a definite effect to further improve predicted system availability.

3. With the incorporation of the self-test, self-calibration, self-diagnostic, and automatic surveillance testing-features, operator; i interface with the system is minimized, resulting in decreased system downtime and improved system accuracy.

4. Actual system MTBFs for a digital system far exceed the expected MTBFs l
(derived from system design data). Because of these results, it is concluded that the quarterly surveillance ; intervals and the six-hour time limit for an inoperable channel to be maintained in an untripped condition that are currently contained in the SQN TSs can be maintained.. The conclusion can also be extended to j encompass the RPS enhancements being incorporated from WOG TOPS as part of this license amendment request. The definition addition, the deletion of unreferenced action statements, .
^
and the deletion of outdated footnotes are administrative in nature and provide clarity and completeness in the TSs. Environmental Impact Evaluation The-proposed change request does not involve an unreviewed environmental question because operation of .SQN Units 1 and 2 in accordance with this ; change would not: l
~~1. Result in a significant increase in any adverse envirot.sental impact previously evaluated in the Final Environmental Statement (FES) as i'~~
modified by the Staff's testimony to the Atomic Safety and Licensing Board, supplements to the FES, environmental impact appraisals, or decisions of the Atomic Safety and Licensing Board. l
2. Result in a significant change in effluents or power levels.

3. Result in matters not previously reviewed in the licensing basis for SQN that may have a significant cavironmental impact.
N
ENCLOSURE 3 PROPOSED TECHNICAL SPECIFICATION CHANGE SEQUOYAH NUCLEAR PLANT UNITS 1 AND 2 .i DOCKET NOS. 50-327 AND 50-328 (TVA-SQN-TS-89-27) DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS
~~.i i )~~
1
~~. j.~~
~~i l l u~~
~~, i \~~
l
~~-q j~~
1 gl
i ENCLOSURE 3 Significant Hazards Evaluation TVA has evaluated the proposed TS change and has determined that it does not represent a significant hasards consideration based on criteria established in 10 CFR.50.92(c). Operation of SQN in accordance with the proposed amendment will nott (1) Involve a significant increase in the probability or consequences of an accident previously evaluated. The proposed' changes that incorporate new RPS setpoints, response times, or protection logic have been evaluated by the reanalysis.of the corresponding Chapter 15 events. These reanalyses demonstrate ]i that the overall conclusions drawn concerning the plant's ability to cope with design basic events remain unchanged. Therefore, these 4 changes:do not increase the consequences of any previously evaluated. ! accident. The proposed changes reflecting.new allowable values do.
~
g not impose any new safety analysis limits or alter the plant's ability to detect and mitigate events. As such, these changes will ' not increase the consequences of a previously analyzed event. The-proposed changes to RPS surveillance intervals and out-of-service times are supported by equipment reliability. studies that demonstrate that'RPS reliability and. availability are not reduced. .Again, because the RPS's ability to detect faulta and initiate protective action is not reduced,-the FSAR analyses. remain bounding,,and the. ! consequences of accidents previously analyzed are not increased. . f The proposed changes to Specifications.2.2.1, 3/4 3.1.1.- I and 3/4 3.2.1 are made to reflect upgrades and enhancements to be i made to the SQN Units 1 and 2 RPS during-each unit's respective Cycle 4 refueling outage. The significant' upgrade to.the RPS consists of replacing the Foxboro H-line analog process protection equipment with the Eagle 21 digital system. The use of the' digital ; f system avoids most of.the drift problems associated.with the analog. .l process instrumentation. The Eagle 21 system also'has improved' test ' features that include automatic surveillance testing, self-calibration of analog circuits, and self-diagnosis'of system troubles. 'The general revision of RPS setpoint allowable values is primarily the result of differences in rack drift and measurement and
~
4 test equipment errors between the analog and digital systems. The [ elimination of the analog rack compactor also affects-the allowable { values. The new allowable values are calculated in Revision 4 of .! WCAP 11239, "Setpoint Methodology for Protection-Systems," utilizing i' accepted setpoint methodologies. The RPS enhancements to be incorporated with the Eagle 21 upgrade are , the steam generator. level EAM and TTD trip reduction features, ' elimination of the steam flow / feed flow mismatch reactor trip through the incorporation of an MSS feature, elimination of the RCS RTD
bypass. manifold, and the incorporation of new SLB protection logic. These enhancements are represented as specific revisions to RPS functional units in the proposed change package and are supported by the reanalysis of corresponding FSAR s'ections.- The remaining changes incorporate the RPS' enhancements described.and evaluated in WCAP 10271 and its supplements. The incorporation of the enhancements completes SQN's implementation of the WOG TOPS. Other WOG TOPS RPS enhancements were incorporated into the SQN:TS by
~~-Amendments 27 and 47 for Unit-1 and Amendments 16 and 39 for Unit 2..~~
The administrative changes made to add a definition and to delete unused and outdated information will not increase the probability or consequences of previously evaluated accidents. (2) Create the possibility of a new or different kind of accident from any previously analyzed. The RPS monitors selected plant parameters and initiates protective ! action-as required. The proposed changes.to the'RPS TS reflect new setpoint allowable values, enhanced protection feature setpoints and ; logic to reduce unnecessary trips,.and enhanced-RPS testing. The-proposed changes that incorporate new RPS setpoints.: response times.. _( or protection. logic have' been evaluated by .the reanalysis' of: the. -I corresponding chapter-15 events. . These reanalyses demonstrate..that overall conclusions drawn concerning'the plant's ability to cope with ! design basis events remain' unchanged. Therefore, these changes do j not increase the consequences of any-previously' evaluated accident. The' proposed changes reflecting'new allowable values: do not- impose any new safety analysis limits.or alter the plant'sLability-to detect -j and mitigate events. As such, these changes will not increase the { consequences of a previously analyzed event. The proposed changes-to l RPS surveillance intervals and out-of-service times:are supported by 1 equipment reliability studies that demonstrate that RPS reliability l and availability is not reduced. 'Again, because the RPS's ability to '
~~-I detect faults and initiate protective action are not reduced, the '~~
~~FSAR analyses remain bounding, and the consequences,ofLaccidents-~~
~~.previously analyzed are not increasedi -Therefore,'the possibility of j how or different accidents from those previously analyzed is not created.~~
The administrative changes made to add'a definition'and to' delete unused and outdated information wi11 not increase the probability or consequences of previously evaluated accicents. (3) Involve a significant reduction in a margin-of safety. The RPS is fundamental to plant safety. The reactor trip system acts ; to limit the consequences of Condition II events;(faults of moderate ~ frequency) by, at most, a shutdown of the reactor and' turbine. This imposes a limiting boundary region to plant operation that ensures that the reactor safety limits analyzed in FSAR Chapter 15 are not E f
exceeded during Condition II events and that these events can be
~
~~accommodated without developing into more severe events.~~
Similarly, the RPS acts to limit the consequences'of Condition III events (infrequent faults) and. mitigate-Condition IV eventsL(limiting' faults). This is accomplished by_ sensing selected plant parameters-and determining whether or.not predetermined safety limits are being1 exceeded. If they are,-the system sends ~ actuation signals to-those components whose aggregate function best. serves the requirements of-the accident.
As previously discussed, the proposed'TS changes reflect RPS upgrades and enhancements that are~ supported by FSAR reanalyses,-accepted setpoint methodology, and staff reviewed reliability studies. - The supporting information. demonstrates (that'the reliability and availability of the'RPS are maintained, if not improved, and that the RPS will still perform its function 'of . sensing plant parameters to initiate protective actions to limit / mitigate faults. Therefore, the proposed changes do not de:rease the' margin of safety. i The changes made to add a definition and to delete: unused and~ outdated information are administrative in nature and therefore do ]i not decrease tbe margin of safety.- l o a i l1 l
~~. ENCLOSURE 4-SEQUOYAH NUCLEAR PLANT FINAL SAFETY ANALYSIS REPORT (FSAR) REANALYSIS ;~~
i MARKUPS FOR EAGLE 21' UPGRADE i 15.2.2 15.2.4- ' ' 15.2.7 i 15.2.8 ! 15.2.9 I 15.2.12 15.4.2.2. ti i h 1 1
~~'. j u~~
l 1
t Changes to FSAR Section 15.2.2-Uncontrolled Rod Cluster Control' Assembly - Withdrawal at Power e
~~'} .l~~
1
?
~~i i 2.~~
~~.'l -l l~~
~~j i 9 O O 6 i 1 i~~
~~.g=~~
~~w I~~
~~a. .~~
I
~~S S I~~
~~. -I . f:~~
5
~~) 4 5 -~~
~~1 _ = g~~
-a s . .. / r s' * . .z s
o
;' t, R; th R -- W " 8 =A i . g- wE !
~~g a 'd C <~~
~~- - w -! t 4~~
~~5 s -~~
~~- g ;-~~
~~y .~~
~~- [.~~~
~~og i~~
~~..a ;~~
~~C m, -!~~
"E.m f*
~~5 .z- *; u .; W .~~
~~= - = %# .)~~
~~ct 1~~
- % i 3 5 .5 - 1 g g, -t g, .,
~~I I I I I I I I 9 SW 8R S' M~~
8 g- g 8 e*
~~. m d~~
~~' d i & Kme RR, f~~
~~(1YNIHON40 (YlSd) 3 MOSS 3Hd - 'ei - w N0110VH3) XD13 N0H103N- #32180SS3Hd '~~
. I' m-9 e (*
~~P n... e e +wo -ngdre.3 ,~~
+ ' 4g <e e sc.nn.
~~,;[~~
f
?
~~4 s t-~~
-1.6 s !.1.4 REACTIVITY INSERTION RATE - 75 x 10-5 4K/SEC 1 h _ 1.2 < j B
~~g 1. <~~
~~i:v '~~
~~g. .8<~~
i D .. ! .; E .6- ! i d : [ .4 Ew
* .2 3
~~[- {~~
~~0. ui '~~
~~0 1 2 3 4 5- .6 7 '8- 9. .10' TIME (SEC) ! i~~
~~+ .i i~~
~~Y gure. 15.2. 2.-l 1 4 2- . t~~
l
~~. -i I~~
~~2600. 2500.~~
~~'1 E 2400.<~~
~~SJ 4 2300'. ! W~~
~~c. 2200.<~~
or
~~. 6J -~~
~~g 2100.< -l N'-2000, n.~~
~~-I 1900. <~~
~~0. 1. 2. 3.- 4. 5. 6. 7. 8. 9. 10. -~~
~~TIME (SEC) 'es,n,w~~
~~=l e-i j~~
~~t 1 I i e igure, b .Q .2 " u 1~~
~~. ~J~~
~~.s' 600 - -~~
~~1 i I 590 W 580 ~~~
~~'E # 570 -~~
~~x .~~
~
~~560 - - 550 l 4.0 j~~
~~} ,~~
~~I 3.5 - I~~
~~. 4 3.0 -~~
E
~~, 2. 5 - ~ ~~~
~~E 2.0 - 1.5 ' 1 l -- l l - i~~
~~t. 1 0 10 -~~
~~20 30 40 50- j TIME (SECONDS) 3 i~~
~~-Rep \ ace. Gnkneir 2 Lnhure Figure 15.2.2-2~~
~~Typical Transient Response for Uncontrolled Rod Withdrawal from Full .~~
~~Power Terminated by High N$ntron Flux Trip-l~~
~~. 1 . a s~~
~~620 ,~~
~~; a0 ,;~~
~~e 8~~
~
~~600< - M i* -- ! f -!~~
~~.{ 590 .w .~~
~~sao - ! 5 !~~
& 570 m u i E .i " 560 4 -- . 550 0 1 2 3 4 5 6 7 8_ 9 10 ,
TIME (SEC). -l i i I l F igure, 1 S.2 .2.- 2. I o9 2. < l y , __
a, W 9 I r 4
~~3. 5 -~~
~~i l1 1 ,.s . 1 2. N 3.5~~
~~1. i 0 1 2- 3 4 5 6- ._7 =. 8 9, .10-TIME (SEC) 'i 4~~
~~4 F tc3ure \ S . 2 . 2. - 2.~~
~~'2. o & 2-~~
~~o 1 .I~~
=
E
~~.g I E~~
~~y ,~~
~~- g i . 'T.. .~~
~~3 s / 5. g.-.~~
~ - . .- ~ .s < - Ee .a v* =g s:
~~8: c .. 3~~
~~= . - *W -~~
ti
~~,s t = -~~
~~tt E :: E . w' $~~
/-u l .gg / 11 e : ., - e.~
~~v .~~
~~.% i~~
~~. gg l~~
~~l- 1 I I l- ! I l- I l-o- T S~~
~~= 8 le sss 3 a~~
~~8.2 g s 8 s.~~
a= nasn g' I (7YWlHON80 -(VISd) 3HnSS3Hd ; :- j noilena) xnla Nounn uninnssna a. H C 6 kN Nk bb O bb 6
~~, . .i I~~
~~1 J~~
~~.I i~~
~~1 1 l r 'j 1 4~~
~~l t~~
~~i 1.6 - J~~
~~'E 1. 4 < REACTIVITY INSERTION RATE --3_x 10-5. AK/SEC E ,.i .~~
~~w C _ ( 1.~~
.E ~U l N u .8, 4 - ?
~~i Det i~~
=
a .b' > 6 1 5 a 4 m i w - l .
~~} . . . . O. .~~
~~0 5 10 15 20 25 30 35. 40 45~ 50- L~~
~~($EC) TIME ,,,,,,,,- 4 i a N i~~
i

l i
~~3 l l~~
Figo.ce 16.1.2. L

o. .
[ of 2 . -
L
~~-t P~~
e 3r h. k-2600.. . l i 2500.< . 5' -!
~~'E 2400.. ~ .~~
~~2300, w m u u 1..~~
~~' 2200.<~~
u . w N i g 2100. , w m w w c. 2000.' . 1900. L -
~~1100.~~
~~l 0. 5, 10. 15, 23. 25. 30. 35. 40, 45. 50.~~
~~- TIME- (SEC) ,,,,,,,,. -r i;~~
~~i l l 1 .: L~~
~
~~Fic3ure \5 .2. 2. ~5 . j I A w * , - - - - + , - , ,- < -~~
~~+ . . . m n ,~~
l Y 6 fA - m. y. la. w** i g g !
* .. . g ' ka 9
~~w . , o~~
~~"~ a. . I \. / -~~
i
~~. w '. S.~~
Ec
~
~~gg X" og R g-~~
/ 8 w ;O- .s-g**
~~t i I s w 1~~
~~.x .~~
~~y y 1 , N~~
""" . >=' E w. 1 - 6. -
~~w G. ?~~
~~.~ r 5A ]~~
~~e i bs T p - g .- L~~
- o. ;; E - - - o. - - ;;. E . '!
~~AW '~~
'bb ; *e I = ,
~~I y -g i i 4 p~~
~~; g '~~
o:' iA N W m O m 8D O CD W P O N e 49 - O. td1. --. N O.N N til e Oa g w 3 W en 6f) to @ 4D F 09- . = =
. ec aswo-en C U ) 0"J. l A
~~Repiace di4h nex+ +wo ftgures , a 9~~
~
~~620~~
~~- Q 610 <~~
~~a 8 [ 600 ; i s .~~
.5
~~590 <~~
~~e 580< w ., W e 570 4 i w '~~
\
E-
~~" $60 < ;~~
~~550 5 to 15 ~20 25 30 35- 40 45 50 TIME (SEC) I s 4 4~~
~~.' :1 Fic3um E.1.2.-4 1 of : 2_. -~~
~~4 l 4 ,~~
)
3.5 i 3.< 1l 1 tv '
~~$ 2.5 '~~
2.
~~1. 5 <~~
i
~
~~1. 0 5 10 15 20 25 30 !35 140 45 50 i I TIME' (SEC)~~
\
~~t .!~~
~~i q~~
~~Fgure 15.2.2 -4 " 2.42. I~~
~~.f'~~
'? .g y ~ -('
N
. a . . % R . s 4 , - m 7 b I . ~ .;j-n l
a
~~- f . 85 E~~
~~6 m .-~~
~~- -- y g g "~~
~~2 "$. d g - g 8e g tE~~
~~# a m WE g- p g.~~
~~t / .~~
- .e E / $ Am g - -
~~e g ;g'~~
~~= /j 4 .~~
~~u . w e~~
~~.sg W~~
E- E &E a-ea e eN e iw s3 g g.g \
~~- g p. -~~
~~2 $. 6~~
~~% t2 -~~
~~d .g g- . s 5 W W' ' i~~
~
~~5 4~~
~~.c g !_I -~~
~~A "m ..~~
*g E = M I w g '
I
~~- g.~~
E w - g l . ~ S 5 s- .
~~g. 1 .~~
~~i 1 t i i T, f~~
~~. k- k W NIW .~~
l
~~. 7 ,~~
8
~~, gg g'(- A d ! .. Yb e~~
~~_ . _ _ _ , , . _ _ _ - _ - _ _--------m----,--.-.-.,.----------- - .~~
~~m---(~~
.. .c & -:: J g ~ g <. N V' ~g m _.
~~x .,~~
~~,Y '. . /-~~
r
/ 5 E ~ -1 \ Q I , g: / ,, 'o- ~
y y ,. . 3 ..
~~$lo - ~~~
~~5 - / ..~~
~~~l .~~
~~mw E~~
~~/ .l.~~
N
-a. . , 'l i a*: EW ~~
~~ag -~~
~~~ Ee ac!~~
~~E- i g-. , : g~~
~~- ~ ::~~
~~l I s e e g- e - i lo E II ,~~
'l .E- ~E , , ~ -- ! g_ .. l
~~e. - ..~~
- ** k k i * "i 5 .
j
~~'W W l o m j g-; :g a~~
~~y ,~~
.. N .__ 'o e . -jf l y i I I
~~6n v $ @ g~~
~~~ ~~~
e J ; ; - I VENO WIWININ .- 1
4
~~/en s- 'g I -~~
n
\
A
~~, s l. .~~
~~E '\ - i l - m e Ieg- lie~~
- - g- - ~. sg 5 / w .
~~S $ / k = U =~~
=
d 3 / - g ._ 1g sj E / k s te b- A
=
~~s w at a a~~
~~-f n g .gg n~~
E j - = E b-TI-a g ng5 C - m w E
$ EE 5 & 55 - ~ C 47 E E EE E t ". -
~~a y a m me a- + C C. -~~
~~, $yg eu be o~~
~~m E E: .~~
~~g E- - 8 ,, 4 d .: - W .. 4 9 g EE - = d il~~
~~. e ~~~
~~E -- e~~
~~~ a~~
~~g. ..~~
-E .n: , .c M E E u - ~ / -
~~9 . 2 I 8~~
\ . = / 1 1 1 1 =
a
~~= = -~~
e aT BBN0 HOWINIH Sc,3 a.Cc vJ (NR rie m- IigLAr6
_ _ _ _ - - - - _ . - ..___ .__ -- ._ .. _ _ _ _ _ _ - _ __ _ _ . _ . _ - . - - .~ _. . . . F n-t
~~. . n. .. .,~~
g
. w m ., r- , .. .;i
~~(- .ag..~~
~~.. m- ..~~
l'
\ > %. -^ s 5 s W. . . ., x y , d G \ ~
~~u z ...~~
> e==. .a >- --o w \ 'T o
E u w .. . n { am , sp. a, 3 .. g \ ... - N W-
~~' emme i~~
w
- g rN ,- ... m w 5
em
~~>N ,~~
zw
~~.1 ..~~
~~2-D~~
-3 z .- ,t d > T ,, O-p .. . ~
~~w$ as - w-g* .v g W .. N. - g 'I~~
* .W .p +
~~I x f$e e m gi~~
-s- >. C E .. . . t.
~~3 g -- kl . w~~
~~-e > I"'.~~
~~g .- . . i~~
~~'W. .* . . m l~~
~~l . .. , =~~
~~, x -~~
~~l l- . . m~~
~~8. 't !~~
I ' i l l i - e 5- 5 I .
~~. t e o e o o ,~~
~~r ,~~
~~. e. m. . n.~~
~~4 l.. - - - - - l; -~~
~~'i l~~
l EENO WOWIN!W 4 I 4 4 7, w , ,w, , -r---- ,
+a ,u e. a w - ~ a a - r x-,.+s--.a<-w e - ** . - . www = - ------ --w ---e_ n- w-- -at- , . - w x- wa - ~+-<-m--y - /.
~~M 9>. ?~~
*- . w f g l $== % * ' . esa W- /
L R - =- E$
~~. T., .~~
c. /-

y j l
y
/ ; . g e-4 g - m y$
g m. . m .h- hw g ~n E ~
~~/. -~~
~~w _e.s , l 5~~
~~- ' -E * ~~~
~~m 6- -~~
~~= / g g- g g s. . . ;~~
~~E :. e"l a sgu mm v v s g o; w~~
~~= .~~
~~h .h ' "' " ( gg o," S8 tt C 218~~
~~.- + -~~
c
~~-- met -~~
~~R S b~~
~~~E- -g E- -~~
~~m 15 s~~
~
i-
~~+ =~~
s g - Nk E Y' .d i WW . w - ,
* *. 5 . ~
~~E.E MM~~
. m xo=g Eg E i S. l N-e m
W
~~-J i -~~
l' i 1 o ' i o o e "' CD. #. up f b .N N O. . pum '
~~-' h. -~~
~~puump~~
' e - ESHQ WOWINIH - , %pl oa wk nexE i
~~_. Sigure <~~
*\
~~1 l l 1~~
~~. . .. .. . .... .. ..---- - - - - - - - - - - - - - - - - - - - - - - - - - -'l~~
~~c 9 M.~~
-i . %% e n m o - g . .. ,
g
~~a. 8 - !~~
~~% '% " e n N-N s q m~~
~~a. A' >e. n- m~~
~~- % a E W .~~
~~( >: M~~
~~.D < n d~~
g
\
~~[. g,~~
~~. .- ~-~~
~~W ,~~
)
~~E I g 5~~
- 1- i =-- --' _ s . g = . . -
~~g3 g I ::~~
~~. - ~~~
~~4ar! .~~
~~;- I. . . -~~
~~m WE' gg a o m , h. 1 mg g g &~~
~
w m . a p 1 g set g e= b w C w n I" W 6 W b a g og . t- ' . , o N 5 m b-as '!
~~> > c' N' N'(.~~
~~D m C -.~~
"E ,e e ;e E . L1 ; = u u 5 'i C
~~5 5 . q m 3D .~~
- e n- N m -* %"E k k , , M l
t i I l , i . .. u i I i g. 1 e I .O-I- g I g - 4 g o ! m 8
~~e e~~
~~Eh e air .O~~
~~# # m' p M8NO WOWINIW t ,~~
l
~~-l I~~
t'
~
~~Changes to FSAR Section 15.2.4 Uncontrolled Boron Dilution .. t~~
+
~~e G s i..<l i 3~~
9
~~-1 .. ?~~
5 I g l l
u' SQN j l tripped within approximately 2.5 seconds-following the drop of a rod cluster. assembly group. The core is not adversely-affected during this ; period, since power is decreasing rapidly. ' I
~~. The most severe misalignment situations with respect to DNBR at significant power levels arise from cases in which group D is inserted to .~~
l its insertion limit ith one assembly fully withdrawn. The insertion limits'in the Technical Specifications are chosen based on a number of limiting criteria. One of'these criteria is that control bank D may be I inserted to its low-low insertion limit at full power with any one q assembly fully withdrawn without the DNBR falling below 1.30. This is ' demonstrated in Table 15.2.3-1. Multiple independent alarms, including a bank insertion limit alarm, alert the operator before the postulated conditions are approached.- .
]
DNB calculations have not been performed specifically for assemblies missing from other banks, however, power shape calculations have been -L done as required for the rod cluster control assembly ejection analysis. 1 Inspection of the power shapes shows that. the DNB and peak kW/f t I situation is less severe than the group D case discussed above assuming
a insertion limits on the other groups. j 15.2.3.3 Conclusions -1 It is shown that in all cases'of dropped single assemblies, the DNBR remains greater than 1.30 at power and, consequently, dropped single assemblies do not cause core damage.
For all cases of dropped groups, the reactor is' tripped by thel power -( j range negative neutron flux rate trip;and consequently dropped banks do -1 not cause core damage. .l l For all cases of any group inserted to its rod insertion limit with any ) single rod c?uster assembly in that group fully withdrawn, the DNBR- i remains greater than 1.30. Thus, rod misalignments do not result'in core l damage. l i 15.2.4 - Uncontrolled Boron Dilution R
~~. . 1 15.2.4.1 Identification of Causes and Accident Description - ~~~
Reactivity can be added to the core by feeding primary grade water into the Reactor Coolant System via the reactor makeup portion of the Chemical and Volume Control System. Boron dilution is a manual operation under strict administrative controls with procedures calling for a limit on 'the. rate and duration of dilution. A boric acid blend system is provided to-permit the operator to match the boron concentration of reactor coolant makeup water during normal charging to that in the Reactor Coolant System (RCS). The Chemical and Volume Control System (CVCS) is designed to 1 limit, even under various postulated failure modes, the potential rate of k s 15.2-12 COC4/Oll5F i
SQN-6 dilution to a value which, after indication through alarms and
~~instrumentation, provices the operator sufficient time to correct the~~
( e situation in a sa,fe and orderly manner. The opening of the primary <ater makeup control valve provides makeup to the RCS which can dilute the reactor coolant. Inadvertent dilution from this source can be readily terminated by closing the control valve. In order for makeup water to be added to the RCS at pressure, at least one charging 5, ump must be running in addition to a primary makeup water pump. The rate of addition of unborated makeup water to the RCS when it is not at pressure is limited by the capacity of the primary water supply . pumps. Normally, only one primary water supply pump is operating while the other is on standby. With the RCS at pressure, the maximum delivery ; rate is limited by the control valve. i The bo'ic acid from the boric acid tank is blended with primary grade water in the blender and the tcaposition is determined by the preset flew , rates of boric acid and prin.ary grade water on the control board. 1 3 ' In order to dilute, two separate operations are required:
1. The operator must switch from the automatic makeup mode to the dilute mode, and

2. The boric acid to blender flow control switch must be turned to the 'g start position. l I i Omitting either step would prevent dilution. Information on the status ;
~~of the reactor coolant makeup is continuously available to the operator l by:~~
1. Status lights on the control board to indicate CVCS operating conditions.

2. CVCS ceviations in flow from programmed levels at the. boric acid and

deminer6112ed water blender. !
~~. . 3. Pressuri;:er level and pressure would be increasing from prescribed values (at higher than planned dilution flows).~~
4 Volume control tank level deviation from programmed level, l Thus there are a number of diverse indications available to the operator 1 to indicate inadvertent or excessive dilutions. 1 15.2.4.2 Analysis of Effects and Consecuent.ej Methods of Analysis i 3 . l
-?: : Ort p W ; c' t%: phM ;,m .ig Boron dilution during l refueling. .. ..w m m , . . . . . . . . . . . m .......a startup, and power
~~(. 1 15.2-13 C004/0115r 1~~
~~, -_ _ - - , , . . . _ . . - - . _ _ ~ . , _ . _ , _ , . . . . _ _ _ . . _ , . . , . _ _ _ . _ . _ _ _ , _ , . . , . _ . _ . . _~~
f ! SON-6 operation are considered in this analysis. Table 15.2-1 contains the ' time sequence of events for this accident. Dilution Durino Refuelino ) An uncontrolled boron dilution accident cannot occur during refueling. - This accident is prevented by administrative controls which isolate the RCS from the potential source of unbotated water. Various valve combinations that are required to be locked closed during < refueling operations are specified in technical specification 3.9.1. . These valves will block the flow reths which could allow unborated makeup . r to reach the RCS, Any makeup which is reuuired during refueling will be . l borated water supplied either from the refueling water storage tank by l the low head safety injection pumps or the centrifugal charging pumps, or ; from the toric acid tanks via a boric acid transfer pump and a $ f centrifuga) charging pump. ggA ; M 1ution Durino Startuo Prior to 5 up the RCS is filled with borated (approximately 20 pm) water from the r llng water storage tank. Nuclear instrumentation is cred closely in cipation of an 6 unplanned reactivity rate of chany . Mixt the reactor coolant is > accomplished by operation of the rea lant pumps. High source range flux level and all react p alarms Q ective. In the analysis, a um dilution flow of 300 gpm limi the I capacity of t a primary water makeup pump's is considered. volume < of the r or coolant is approximately 9967 ft', which is the active voh , of the Reactor Coolant System excluding the pressurizer.
/
Dilution Followino Reactor Shutdown s l Following reactor shutdown, when in tot standby, hot shutdown, and I subsequent cold shutdown condition, and once below the P-6 interlock setpoint, in accordance with Sequoyah Nuclear Plant Units 1 and 2 - Surveillance Instruction-603, (High Flux Adjustment Af ter Shutdown), the high flux at shutdown alarm setting will be adjusted to no higher than 1/2 decade above the cour.t rate 30 minutes af ter plant shutdown. The alarm setpoint must be set or verlfled every 30 minutes for the first i 2 hours following plant trip, every 2 hours for the next 6 hours, and~ once per shif t thereaf ter until the flux level has stabilized. p,pl,c, gyh lnsery , h at Power _ Hith the unit at powe RCS at pressure, the ' rate is / limited by the capacity of the p , wat up pumps which supply the charging pump. A conservat gh v r the expected boron 7 I . concentration ower and a conservative . ilution flow g rate s used.
~~/p ; \~~
~~1S.2-14 y CO 4/Oll5F , i~~
~~$0W-6 15.2.4.3 conclusions~~
(
~~.- For dilution during refueling: , Dilution during refueling cannot occur due to admiristrative controls (see Section 15.2.a.2).~~
The operator has prompt and definite indication of any boron dilution from the audible count rate instrumentation. High count rate is alarmed key lgcg,- in the reactor containment and the control room. In addition, a high #h source range flus level is alarmed in the control room. The count rate increase is proportional to the subtritical multiplication factor. . lnsed b r dilution during startup: For lution during startup, there is adequate time (-52 minutes) f m Ig - transie t initiation for the operator to recognize the high. count att I signal a terminate manually the source of dilution flow. The operator alerted to the uncontrolled reactivity inser ion during startilp via the ncressing count rate on the Source Range guclear i Instrumentation, ecorders on the control board contindusly provide a tire history of the uclear flux level. This increase'in flux level is very slow, based *on t reactivity insertion rate for the startup case, it takes apprortmately minutes for the flux level to increase by a factor of 2. This is ade ate time for the operator to recognize from-
/
~~therecorderstheneedforaqton. Thus there would still be 48 minutes l l ( for the operator to ascertain (nd isolate.the source of the reactivity~~
~~. insertion. Also on the source range channel is the high flux at shutdown alarm. The setpoint for this ala is p6rmally placed at 5 times the~~
~~! source level. Even assuming that t aperator doesn't recognize the j increasing count rate..the alarm wili~~
~~cur at approximately 70 minutes into the transient. Thus there wo61d st 11 be 21 minutes for the~~
~~. operator to stop the dilution. /~~
/
~~For dilution during full powe'r operation:~~
/
~~1. With the Mctor jr(automatic control, the po r and temperature~~
~~. . increase from bpron dilution results in inserti of the rod cluster control assemblies and a decrease in the shutdown argin.~~
The opera r is alerted to an uncontrolled reactivity n$ertion by the rod nsertion limit alarms. Two insertion limit al ms are avail le: The first occurs when the rods are 10 steps a ve the ins tion limit (Lo Insertion Limit Alarm) and the second o urs at t insertion limit (Lo-Lo Insertion Limit Alarm). The analy 5 ssumed that the operator is alerted to the need for action by e Lo-Lo Alarm although action would be taken when the first alarm occurs. Thus the analysis already assumes a 10 step allowance for / k l 15.2-15 COC4/0115F
asi- i Delet e red position indicator inaccuracies. Even eith this conservatism, j ere are stt11 47 minutes available from the time of alarm untti g nutdown margin is lost. In addition to the above, other al indic ons are available. The main indication would be a of the 4x 1 offset control band which would result in a eactor lation ) trip (reduc n in overtemperature At setpoint). - 2'. With the reactor manual control and if no o ator action is taken, the power and moerature rise will se the reactor to reach the overtemperatur T trip setpol . The boron dilution accident in_this case is es t1411y)4entical to a rod cluster control assembly withdrawal at eor. The maalmum reactivity ' insertion rate for boron' dilutlen ' less than 1.6 X 10" $ Ak/second and is seen to be ithin t range of_ insertion rates
~~analyzed. Prior to the temperature trip, an over- ,~~
temperature AT alarm turbine runback wo d be actuated. ! The analysis sh that for the maximum dilution te possible, trip occurs in abp 1 minute. Thus, 46 minutes are avai le for the g operator Je recogni:e the abnormal CVCS status indicati and/or the abDormal neutron flur indications and stop the dilutto low., , ! The, ge'nsitivity and alarm thresholds are already assumed to b graced'to the maximum extent allowable for the overtemperature j
~~,/at trio function (see section 15.2.2).~~
For Dilution Following Reactor Shutdown: l In providing a description of a boron dilution event initiated immediately after scram, it is appropriate to analyze two initial *
~~)' i i conditions. These are:~~
~~1. BOL, Equilibrium Xe '~~
i This will result in the longest time following scram until the Source Range Nuclear Instrumentation System (NIS) is available . to provide an indication of a dilution event.
~~2. BOL, Clean Core This will result in a very short time following scram for the source range NIS to become evallable, however, it yields the-most rapid boron dilution (return to criticality) case.~~
Figure 15.2.4-1 shows the relative change in boron concentration with time for the two cases. The dilution rates are consistent with the time for the two cases. Figure 15.2.4-2 shows the condition of the core consistent with the boron concentrations of Figure 15.2.4-1 and Xe build-up fc110 wing trip for the Eq Xe cases.
~~. )'~~
15.2-16 COC4/0115F i l
SQN 1 i rigure 15.2.4-3 shows the information available to the operator on the core relative power based on the Nuclear Instrumentation System for the ( EQ xe case. As shown there is essentially an instantaneous decrease in nuclear power from 100% to 7.5% (< 5 seconds). From 7.5% the standard 80 second period is used until the precursor isotopes have been depleted. / From the point shown, an 18 day half life is assumed. For the case without EQ Xe, the NIS. stable reading on source range is achieved very 4 rapidly, ( 5 minutes as opposed to 21 minutes for the IQ Xe case. Secuence of events Tables are attached for both cases (see Tables 15.2.4-1, 15.2.4-2). These show that for both cases > 15 minutes of optrator action time is available. Therefore the acceptance criteria.for this event is met.- In addition to the High Flur at Shutdown Alarm, there is also the High Pressurizer Level Trip and alarm available. In order to return critical a very large total dilution volume is recuired. The only means of accommodating this large volume is to allow the pressurizer to , start filling. As shown, however, this results in a High Pressuriter ! Level Alarm very early in the transient. , These two alarms would provide the operator an adeguate set of l incitations that a boron dilution event was in progress and also allom aceauate time for operator corrective action. . 15.2.5 Partial
~~Loss of Forced Reactor Coolant F1p , ,~~
15.2.5.1 Identification of causes and Accident Description i A partial loss of coolant flow accident can result from a mechanical or electrical f ailure in a reattor coolant pump, or from a fault in the i power supply to the pump. If the reactor is at power at the time of the
~~- accident, the immediate effect of loss of coolant flow is a rapid -~~
increase in the coolant temperature. This increase could result in DNB with sucseQuent fuel damage if the reactor. is not tripped promptly. The necessary protection egainst a partial loss of coolant flow accident is provided by the low primary coolant flow reactor. trip which is ; actuated by two out of three low flow signals in any reactor Coolant loop. Above approximately 351 power (Permissive 8), low flow in any loop
~~. . will actuate a reactor trip. Between approximately 10% power (Permissive :~~
7) and the power level corresponding to Permissive 8 low flow in any two loops will actuete a reactor trip. A reactor trip signal from the pump breaker position is provided as an anticipatory signal which serves as a '
backup to the low flow signal. It functions essentially identically to the low flow trip so that above Permissive 8 a breaker open signal from any pump will actuate a reactor trip and between Permissive 7 and - Permissive 8 a breaker open signal from any two pumps will actuate a reactor trip.
+
~~e i e i 15.2-17 CO 4/0115r . .~~
~~.v. - -.. -. - - ...w.. + .,~~
~~l' IN$!RT A i Dilution Durino Startun I~~
\
In this mode, the plant is bein i Het Standby, to another, Power.g taken from Typically, the one plant long term mode in is maintained of the operation, ! Startup mode only for the purpose of startup testing at the beginning of each ! cycle. During this mode of operation rod control is in manual. All normal j actions required to change power level, either up or down, require operator i initiation. Cond +. ions assumed for the analysis are: - i
~~1. Dilution flow is the maximum capacity of the makeup water pumps, !~~
~~300 gpm. l~~
~~2. A minimum RCS water volume of 9019 ft?. This corresponds to the ;~~
active RCS volume excluding the pressurizer and the reactor vessel i upper head. I J
3. - The initial boron concentration is assumed to be 1800 ppm, which is a I conservative maximum value for the critical concentration at the condition of het zero power, rods to insertion limits, and no Xenon.

4. The critical boron concentraticn following reactor trip is assumed to l be 1600 apm, corresponding to the hot zero power, all rods inserted 1 (minus tie most reactive RCCA), no Xenon condition. The 200 ppm change )
~~from the initial condition noted above is a conservative minimum value. -' v. f l t i i l I i i e e~~
~~-n . - - - - , , - . . - - - - - , , ,,w. , . , . .~~
~~- - . - . _ . - . - . . ~ . - - - - - -.- - . . - - . - .- -. - .-~~
~~l INSERT B Dilution at Pewer In this mode, the plant may be operated in either automatic or manual rod i control. ; Conditions assumed for the analysis are:~~
1. Dilution flow at power is the maximum capacity of the makeup water pumps, 300 gpm. ,

2. A minimum RCS water volume of 9019 ft3. This corresponds to the active RCS volume excluding the pressurizer and the reactor vessel upper head.

3. The initial boron concentration is assumed to be 1575 ppm for Unit I and 1500 ppm for Unit 2, which are conservative maximum values for the critical concentrations at the condition of hot full power, rods to insertion limits, and no Xenon.
The critical boron concentration following reactor trip is assumed to 4. be 1195 ppm for Unit 1 and 1150 ppm for Unit 2, corresponding to the hot zero power, all rods inserted (minus the most reactive RCCA), no Xenon cond.ition. The 380 ppm (Unit 1) and 350 ppm (Unit 2) changes from the initial conditions noted above are conservative minimum values. p #
~~.g-e~~
a ; l l INSERT C l l For dilution during startup: l . This mode of operation is a transitory operational mode in which the operator intentionally dilutes and withdraws control rods to take the ; plant critical. During this mode, the plant is in manual control with the j operator required to maintain a high awareness of the plant status. For a l the operator must manually initiate a J normal approach limited dilution and to criticality,ly manually withdraw the control rods, a subsequent . ; process that takes several hours. The Technical Specifications require ! that the operator determine the estimated critical position of the control ' rods prior to approaching criticality, thus assuring that the reactor does o not go critical with the control rods below the insertion limits. Once , critical, the power escalation must be sufficiently slow to allow the 1 operator to manually block the source range reactor trip after receiving j P 6 from the intermediate range. Too fast a power escalation (due to an J unknown dilution) would result in reaching P 6 unexpectedly, leaving i insufficient time to manually block the source range reactor trip. ! ) , Failure to perform this manual action results in a reactor trip and imediate shutdown of the reactor. For dilution during startup, there are more than 19 minutes avaliable for operator action from the time of alarm (reactor trip on P 6) to loss of shutdown margin. For dilution during full power operation: ; With the reactor in automatic rod control, be power and temperature . increase from boron dilution results in insertion of the control rods and i a decrease in the available shutdown margin. The rod insertion limit alarms (LOW and LOW LOW settings) alert the operator that a dilution event is in progress. There are more than 42 minutes available for operator action from the time of alarm (LOW LOW rod insertion limit) to loss of - shutdown margin. .I With the reactor in manual control and no operator action taken to terminate the transient, the power and temperature-rise will cause the reactor to reach the Overtemperature AT trip setpoint re:ulting in a reactor trip. The boron dilution transient in this case is estentially the equivalent to an uncontrolled RCCA bank withdrawal at power. The .l maximum reactivity insertion rate for a boron dilution is conservatively estimated to be about 2.5 pe#see which is within the range of insertion rates analyzed. There are more than 40 minutes available for operator action from the time of alam (overtemperature delta T) to loss of ; shutdown margin. ..- j l l 1 i u
i
~~$QN-6 TABLE 15.2-1 (Sheet 2)~~
(
~~, (Continued) .~~
TIME SE0VENCE OF :VKNTS FOR I CONDITION ll :VI:NTS Accident [yanj, Tlee (Sec.) Uncontrolled Boron - Dilution
1. 011ution during . 3 refueling and startup Dilution begins 0 Operator isolates source of P dilution; minimum margin to reSuch'42:00 .., 0 % recluded criticality occurs
~~$+=M N0p;t:-h;40 > t1 f6~~
2. Ollution During Full Power Operation 3

4. Automatic
~~Reactor Control Shutdown margins lost M 2.520 l6~~
~~b. Manual~~
{ ; Reactor , V Control Dilution beg'fns 0 Reactor trip setpoint reached for overtemperature AT ,56 Al20 M ; t;;t O ' 11 '-;^ :: : 5? Shutdown margin is lost (if j dilution continued af ter t'i9). .aM*J >2400 i Partial loss of Forced Reactor Coolant Flow
1. All loops operating, two pumps coasting down Coastdown begins 0 3 Low flow reactor trip 1.72 Rods begin to drop 2.72 Minimum DNBR occurs 3.4, Added by Amendment 6 g
i COC4/0723r i
W 1 1 j l Changes to TSAR-Section 15.2.7 loss of External Electrical Load and/or Turbine Trip u i e
~~'W k~~
~~t h a S 9~~
~~-? .e.~~
f
SQN-6 l
)
7. No mixing .is assumed in the inlet plenum for the reactivity calculations. For conservatism the loop with the largeet i C temperature change. I.e., the inactive loop, was used for the calculation of nuclear power. j In the analysis reactor trip is conservatively assumed to be actuated by the high neutron flux reactor trip. The trip.setpoint was assumed to be { 116% of nominal full power. In practice, however, reactor trip would be { expected to occur on power range neutron flux exceeding the P-8 setpoint i (set at 35% of nominal full power). The P-8 setpoint will remain active until flow in the inactive loop reaches 90% of its nominal value, l6l j Results i l The results following the startup of the idle pump with the above listed { assumptions are shown in Figures 15.2.6-1 and 15.2.6-2. The minimum DNBR during the transient is never less than 1.30. The calculated sequence of events for the accident is shown on Table 15.2-1. 4 15.2.6.3 Conclusions ! The transient results show that the core is not adversely affected, i.e., : there is Considerable margin to a limiting DNBR of 1.30.
)
15.2.7 Loss Of External Electrical Load And/Or Turbine Trio I-15.2.7.1 Identification of Causes and Accident DescriDtion M6jor load loss on the plant can result from loss of external electrical load or from a turbine trip. For either casejoff site power is available for the continued operation of plant components such as the reactor ccolant pumps. The case of-loss of all AC power (station blackcut) is , analyzed in Subsection 15.2.9. Following the loss of generator load, an immediate fast closure of the turbine control valves will occur. This will cause a sudden reduction in steam flow, resulting in an increase in pressure and temperature in the steam generator shell. As a result,'the neat transfer rate in the steam generator is reduced, causing the reactor cc0lant temperature to rise, which in turn causes coolant exoansion, , l pressurizer insurge, and RCS pressure rise. For a turbine trip, the reactor would be tripped directly (unless below approximately 50% power) from a signal derived from the turbine autostop oil pressure (Westinghouse Turbine) and turbine stop valves. The turbine ; stop valves close on loss of autostop oil pressure actuated by one of a number of possible turbine trip signals. Turbine-trip initiation signals 6 include: ,
1. Generator Trip

2. Low Condenser Vacuum C
~~15.2-21 CCCa/Oll5F~~
~~.w.n~~
~~3. Loss of Lubricating 011 4~~
~~Turbine Thrust Bearing failure !~~
5. Turbine Overspeed

6. Manual Trip (}
L Uten initiation of step valve closure, steam flow to the turoine stocs ! abruptly. Sensors associated with the stoo valves cetect the turoine trip and initiate the turbine trip and initiate steam cump anc if a00ve 6 , 50 percent power, a reactor trip. The loss of steam flow results in an almost immeditte rise in secondary system temperature and pressure witi. . resultant primary system transient. : The automatic steam dump system would acccmmodate the excess steam generation. Reactor coolant temoeratures and pressure do not signifi-cantly increase if the steam cump system anc pressurizer pressure control ! system are functioning properly. If the turbine condenser were not availaole, the excess steam generation would be dumped to atmoschere ! through the steam generator relief ano safety valves. Acditionally, main feedwater flow would be lost if the treine concenser was not available. . For this situation feedaater System. feedwater flow would be maintained by the Auxiliary The SeQuoyah plant is designed to accept a load rejection of 50 percent of its rated electrical loac, and signals from the reactor protection ! system of percent will trip load. rated the plant for load rejections in sufficient excess of 50 In the event tht* steam dump valves fail to open following a large loss of load, the tripped bysteam the highgenerator safety valves may lift and the reactor may be water level signal prM he overtemperature \T signal F The steamressurizer wg gm generator shell si e pressure and reactor coolant temperatures will l horca t increase rapidly. safety valves are, The pressurizer safety /alves and steam generator- g~ um .owg _.e and steam generat wever, sized to protect the Reactor Coolant $ystem gainst overpressure for all _ load losses without assuming the operat on of the steam dumo system, pressurizer spray, + , pressuri:er power operated relief valves, automatic rod cluster control ' assembly control C r, W direct reactor trip on turbine trip. The steam generator safety valve capacity is sized to_ remove the steam flow 8t the percent Engineered of steam flow atSafeguards Design rating (approximately 105 rated power) from the steam genera esteeding 110 percent of the steam system design pressure. tor The without pressurl:er safety valve capacity is sized based on a-complete loss of - heat sink with the plant initially operating at the maximum calculated turbine load along with operation of the steam generator safety valves. The pressurizer safety valves are then able to maintain the Reactor Coolant System pressure within 110 percent of the Reactor Coolant System cesign pressure without direct or immediate reactor trip action. 15.2-22 C004/Oll5F
J SON-6 ! L *Cre CCmDiete di cussion of overpressure protection can te founc i.- Efferente T5. [( Normal poner for tne reacter coolant pumps is supplied through busses from a transformer connected to the generator. When a generator tri: ; otturs, the busses are automatically transferred .to a transformer sucolis: from external power lines, and the pumps will continue to su:: y coolant flow to the cere. Following any turbine trip where there are no l electrical f aults whicn re:;uire tripping the generator from the net.crk, the generator remains connected to the network for approalmately 30 / setends. The reactor coolant pumps remain connected to the generator, 4 thus ensuring flow for 30 seconds before any transfer is made. Dould the network bus transfer fail at 30 seconds A complete loss Of f;rced reactor coolant flow would result. This assumption is mace for the analysis of a comolete loss of lead at approximately .507. power witnout direct reactor trip. The immediate effect of loss of coolant , fl0w is a rapid increase in the coolant temperature in addition to tte increased coolant temoerature as a result of the turbine tric. This increase could result in DN3 with subsecuent fuel damage if the reactor
~~were no: tripped promptly.~~
ine f011cwing signals provide the necessary protection against a Ccm01ste ' icis of flow accident: 1. Reactor coolant pump power supply uncervoltage or underfrecuency ( 2. Low reactor coolant 1000 flow The reec:or trip on reactor coolant pump undervoltage is provided to protect aghinst conditions which can cause a loss of voltage to all rett or :colant pumps, i .e., station blackout. below 3;croximately 10 percent power (PermissiveThis 7). function is bicc(e: TM ructor trip on reactor coolant pump underfrequency is provided to one if 0 m;ter coolant pumo breakers ano trip the reactor for an uMerfmuent) ne p r x : grid. condition, resultinQ frcm frequency disturDances on t9e ~~ b The trip '_isengages the reactor coolant oumes frem tre coser&,m coes: crid sc that the pumt's Ainetic energy is available for full-The reactor trip on low primary coolant loop flew it provided to protect against Icso b foss cf flow conditions which affect only one reactor coolant per reactori is function is gtherated by two out of three low flow signals coolant loop. (Dermitiive 7) Ettween approxinately 10 percent power lau the power flow in any two loops will actuate level corresponding to t ermissive 8, low a reactor trip. 15.2-23 C00a!0115~
~~~ . m.- :~~
15.2.7.2 enalysis o' Rffects see Consemences Met %dt ef analvsis The total loss of load transients are analy:ed by ertploying the ::etailec digital computer program LOFTRAN. C-The program simulatts the neutron linetics, Reactor Coolant System, pressurizer, pressuri:er relief and safety valves, pressuri:er spray, steam generator, and steam generator safety valves. The program ccmputes pertinent plant variables including te.nteratures, pressures, and co er level. The core limits as illustrated in Figure 15.1.3-1 are useo as inout to LOfiRAN to determine the minimum DNBR of the typical or thimble cell during the transient. In this analysis, the behavior o' "le unit is evaluated for a complete loss of steam load frem 102 percen', of full power without a direct reactor trio primarily to show the adequacy cf the or 'sure relieving devicet and altos ,# to Demonstrate core protection rnargins. 6 lysts is also pertormeo to ,g,g evaluate a complete loss of steam loao rcm 52 percent of full power without reactor trip. all the sensors for reactor trip on the turbine stop valves.The The turpine is assume assumotion oeiays reactor trip until conoitions in the RCS result a trip due to other signals. Thus, the analysis assumes a worst transient. addition, no credit is taken for steam cumo. Main feedwater flow is In , terminated at the time of turbine trip, witn no credit taken for auxiliary feedwater to mitigate the consequences of the transient. A fast bus transfer is attemptec 30 seconds following the loss of steam load from 52 cercent power. Tne transfer to an external power source is assumed to fail which results in a cceplete loss of flow transient initiated from the loss of load conditions. The loss of flow transient, due to the assumed failure of the fast buss transfer, is analyzed by employing the detailed digital computer codes LOFTRAN (4), FACTRAN (3), and THINC Subparagraph 4.4.3.4 The FACTRAN Code calculates the heat flux transient based on the nuclear power and 'g I flow from LOFTRAN. Finally, the IHINC Code calculates the DNER curing the transient based en the heat flux from FACTRAN and flow from LOFTRAN. Typical Assumptions are: 1. Initial Operatino Concitions - the initial reactor power and Reactor Coolant System temperatures are assumed at their maximum values consistent with the steady state full power operation and 52 percent errors. operation including allowances for calibration and instrument power The initial Reactor Coolant system pressure is assumed at a minimum value consistent with the steady state full power operation and 52 percent power operation including allowances for calibration and instrument errors. This results in the maximum power difference for the load loss, and the minimum margin to core protection limits at the initiation of the accident. Table 15.2.7-1 summar12es the initial conoitions assumed. 15.2-24 COC /Oll5F 1 s
SQNo5
~~2. %oderator and Deepler Coefficients of Reactivity - the total loss of loacisanalyZeeforboththebeginninfoflifeandendoflife~~
( . conditions. Moderator temperature coefficients of tero at beginning of life and a large (absolute value) negative value at end of life are used. A conservatively large (absole+e value) Doppler power coefficient is used for all cases.
~~3. Reactor Control - from the standpoint of the maximum pressures attained it is conservative to assume that the reactor is in manual~~
~~control.~~
~~4. Steam Release - no credit is taken for the operation of the steam -~~
cump system or steam generator power operated relief valves. The steam generator pressure rises to the safety valve setpoint where steam release through safety valves limits secondary steam pressure at the setpoint value,
5. Pressurizer Sorav and Power Operated Relief valves - two cases for both tne beginning anO end of life are analyzed;

a. Full credit is taken for the effect of pressuriter spray and power operated relief valves in reducing or limiting the coolant pressure,.

b. No credit is taken for the effect of pressurizer sprey and power operated relief valves in reducing or limiting the coolant pressure. Pressurizer heater operation is assumed since heater t operation on high pressurt2er wat:r level wil) tend to increase the maximum surge rate through the pressurt2er safety valves.

6. Feedwater Flow - main feedwater flow to the steam generators is ,
~~assumed to be lost at the time of turbine trip. No credit is taken for auxiMary feedwater flow since a stabilized plant cordition will~~
be reached before auxiliary feedwater initiation is normally assumed to occur; however, the auxiliary feedwater pumps would be expected to start on a trip of the main feedwater pumps. The auxiliary feedwater flow would remove core decay heat following plant stabilization.
7. Reactor trip is actuated by the first Reactor Protection System trip setpoint reached with no credit taken for the direct reactor trip on the turbine trip. Trip signals are expected due to high pressurizer pressure3 overtemperature AT, high pressurizer water level, low 5 reactor coolant loop flow, see reactor coolant pump power supply undervoltage, and low-bw steam acena.reder we.4er le ve.l .
Except as discussed above'," normal reactor control system and Engineered. Safety Systems are not required to function. The Reactor Protection System may be required to function following a' turbine trip. Pressurizer safety valves and/or steam generator safety valves may be required to open to maintain system pressures below allowable limits. No single active failure will prevent operation of any-( system required to function. i 15.2-25 COC4/Oll5F
SQN Results The transient responses for a total loss of load from 102% full power operation are shown for four cases; two cases for the beginning of core i
~~- life and two cases for the end of core life, in Figures 15.2.7-1 through ,~~
l 15.2.7-8. Figures 15.2.71 and 15.2.7-2 show the transient responses for the total i loss of steam load at beginning of Ilfe with zero moderator temperature l coefficient assuming full credit for the pressurizer spray and pressurizer power operated relief valves. No credit is taken for the . steam dump. The reactor is tripped by the 0" -t:r;: Ot;r: l' tr!r high res P s Lie t te r th e"et. The minimum DNBR is well above the 1.30 value. PrerscAre signe / 4 Figures 15.2.7-3 and 15.2.7-4 show the responses for the total loss of load at end of itfe assuming a large (absolute value) negative moderator i temperature coefficient. All other plant parameters are the same as case above. "N; 7;.s t !! t-i ppe d i, C.: :" -t ;: :t2 ? ^T t'i? 'h'- C' !
~~'t: D3? 17 ::;;:: t' : ;h;. th: 'r: :':-t On: n; . :- 1 ;: t;le. i t;-~~
M u i ..le:. See, lnsert en Mf $3C ; IE 7 i i'?-a in 5$.[.[i;fl"'!",*I['.vsvu;}r,;.[.i'i_i;l"**"'ntnetrannen., r.y .. _ ... .-,. The total loss of load accident was also studied. assuming the plant to be initially operating at 102% of full power with no credit taken for, the , pressurizer spray, pressurizer power operated relief valves, or steam dump. The reactor is tripped on the high pressurizer pressure signal. Figures 15.2.7-5 and 15.2.7-6 the beginning of life transients'with zero , moderator coefficient. The neutron flux remains constant at 102% of full power untti the reactor is tripped. The DNBR generally increases throughout the transient. In this case the pressurizer safety valves are actuated. Figures 15.2.7-7 and 15.2.7-8 are the transients at the end of life with , the other assumptions being the same as in Figure 15.2.7-5 and 15.2.7-6. Again the DNBR increases throughout the transient and the pressurizer safety valves are actuated. l l t 1 15.2-26j CDC4/0115r
~~. . . _ . , ., , - - m_..,~~
t
- g g 7 ,3 c 0 m b n E p g3 l5,9 7 6 A w pw.qyph The magnitude of the reactivity coefficients used in the analysis in conjunction with the pressurizer pressure control assumed cause this case to differ from the previous case. The pressure control limits peak RCS pressure, initially vi6 pressurizer relief valves, and the large reactivity feedback causes the nuclear power to be eeduced due to the rapid rise in average temperature. High pressurizer pressure reactor trip is prevented by the relief valve action followed by the r duction in nuclear power due to reactivity feedback effects. If less conservative reactivity coefficients were used, the reactor trip would occur earlier, on overpressure, as in the previous case.
The system stabilizes at about 65 percent of nominal nuclear power due to the combined effects of the reactivity coefficients-and the steam flow through the steam generator safety valves. This stablized condi-tion continues ur.til sufficient secon::ry side mass is lost through the steam generator safety valves to cause a reactor trip on low-low steam generator water level. This is a highly conservative analysis because, in actuality, the level in the steam generator downcomer would rapidly decrease to the low-low level trip setpoint within the first 10 seconds due to shrinkage in the secondary fluid caused by the increase in secondary side pressure and the rapid primary side power reduction. The DNBR increases throughout the transient and never drops below its initial value. Pressurizer relief valves and spray prevent primary system overpressurization as described above; steam generator safety , valves prevent overpressurization in the secondary side. The pressurizer safety val.ves are not actuated for this _ case.
E TAELE 15.2-1_(Continued) , f
~
TIME SE01'ENCE OT EVENTS TOR CONDITION 11 EVENTS l A :ident Event Tine (Sec.) , Partial Loss of Forced . Reactor Coolant T2cw (cont.) l
~~2. All but one loop .~~
O rer <==. coasting down r >- coastdown begins O Lev flow reactor trip 3. 2,1 Reds begin to drop 4.21 [ Minimum DNBR occurs 4.8 I Startup of an Inactive Reactor Coolant Loop Initiation. of pun:p startup 0 Power reaches high nuclear flux trip 9.3 Rods begin to drop 9.8 Minimum DNBR occurs 11.0 m - Loss of External -* Electrical Lead '
~~1. With pressurizer~~
( , control (BOL) Loss of electrical load 0 l Initiation of steam release from ! steam generator safety valves ->W T . O j O 0 +...;. i:::r: _
~~-~r !?w~Reactor Trip Setpoint reached .A3 T . C)~~
~~Rods begin to drop .4G+lC).9 O -~~
~~#1=t=== oxia ===r- 22.5 Peak pressurizer pressure occurs 4,4- l ~J . O i i~~
l (Sheet 2) .I H i
r SON-3 TABLE 1$,21 (Sheet 4) [ (Continued) TIME SE0',lENCE Or EVENTS FOR C0hD1110N 1] (V[hi$ Accident g Time (Sec. ) Loss of External Electrical Load (Cont.) ,
2. With pressuri2er i' control (EOL) Loss of electrical load 0 Initiation of steam release M S,0 .
~~from steam generator safety { valves ! nhM.I A d.bM eactor~~
~~Trip Point Reached ,$df'7.6,*2- j~~
~~. Rods begin to drop M 17 1 j Minimum DNBR occurs (1)~~
~~Peak pressurl2er pressure JtC' 9, 5 ) occurs . [~~
~~3. Without pressurizer i control (BOL) Loss of electrical load ,~~
~~O Initiation of steam release ' .;MJ p, , from steam generator safety , valves High pressurl2er pressure~~
~~, reactor trip point reached ,4dr (o,"5 ~ '~~
Rods begin to drop .L4P F, 3 l Minimum DNBR occurs , (1) Peak pressurizer pressure lk7 9,0 , occurs- l (1) DNSR does not decreast below its initial value. I v Revised by Arrendment 3 s l COC4/0723F -
I t s ; 50N-3 ! i TABLE 15.2-1 (Sheet 5) ( (Continued) j
~~. Tlwt st0VENCE OF EVENTS roR COhDITION II EVENTS Accident Event Tlee (Sec.) {~~
Loss of External - f Electrical Load (Cont.) l 4 Without pressurlZer f control (EOL) Loss of elettrical lead 0 Initiation of steam release from steam generator safety valves M S.O F High pressurizer pressure
~~.P3 6 .1 reactor trio point reached Rods begin to drop M T.1-Minimum DNBR occurs (1) , ,~~
Peak pressurizer pressure occurs .24 3.5- ; loss of Normal feedwater low-low ste,m generator water and Lots of Off-site level reactor trip; reactor . Fo.er to the Station coolant pumps begin to coast ! Aust11arles (Station) down 0 Blackout) - # Rods begin to drop 2 Two steam generators begin to receive auxiliary feed from one > motor-driven auxillary feed. ; water pump 60 , Peak water level in pres 5Url2er occurs ' 2900 Excessive feedwater at ! full load One main feedwater control valve falls fully open 0
'~
Minimum DNBR occurs 15.2 feedwater flow isolated due to high-high steam generator level 14.0 t (1) DNBR does not decrease below its initial value. t Revised by /*ercent 3
~~CDC4/0723r T~~
~~~ . , , . - --e , n , , - - . - -~~
~~Replace, wi fk Figwe,s on 4he Ollodm3 pge., I l.2~~
~~s l l l l g 0 -~~
g 0.5 - g % 0.6 - 0.4 - u , a 0.2 - O ' NM 2300 - g g 2200 - 2l00 - yE5 2000 - E 1900 - 1800 11! $70 e / 3 560 - N h [ 550 - g 540 530 2.00 \
~
~~1.90 7- . l.80 - E i, o _~~
~~.60 - -~~
1.50 ! ! ! I 0 10 20 g- 30 40 - 50 TIME (SECONDS) Figure 15.2.7-1 Loss of land Accident with Pressurizer Spray and Fower ( Operated Relief Volves,Beginning of Lif e 8 lr
~~! Revised by Amendment 5 e~~
~~1.2 < C ' g~~
~~.6< ,g .6<~~
~~et bg 4< V .2< Q t t.~~
~~8. 10. 20. E0. 40. 50. E0. 70. 60. 80. 100. '~~
~~11Mt 15tt! _ 2ECO. < 2 2400. < 5% Wb gw 2200. E@2000. E 1E00. D. 10 .. 20. 50. 40. $0. (C. 70. 60. *C. 100. Tlet t$tti 600. '~~
" ~*
~~540. en w d$ gg- . 3 EE 570. w ?- gg Sec. um I 550. j 54e.~~
~~e. 10. 20. 50. 40. 50. 60. 70. 80. 90. 100.~~
~~. . TIMt ISCCI 4.<~~
5.5 - E. ce g 2.5 - ... 2. 1.5 < . 1. B. 10, 20. 50. 40. 50. ED. 70. 60. 40. 100. 11MC (SEC) FIGURE 15.2.7-1 Loss of load Accident with Pressurizer Spray and Power Operated Relief Valves, Beginning of Life
4 Rep ace 50 ELSu.res on Nie bl/odng ( po.p n i .w
~~.S }4 JA /~~
~~71 5 8 d8 b i 1 - 2 't~~
~~/ ~~~
~~l 3-= q. 3 t 4~~
~~/ - - ~~~
~~R" b~~
) \ es m %8 II sa -; ; I I N I / 8 g 4!
~~3 E E~~
~~- - _I a. a. g 3 ,,~~
~~(do) E : , (1734 Disno) .- unioA niva nzivnens, nniendwn seynny o3 g, ' c - 9 1 f~~
)
~~i i 1000. < f 1400. 5E . ,.~~
~~$w 1200. ; *v j 5~~
~~Mg iece. - - _ a h"'600. c ;~~
\
g see. C. 10. 20. !C. 40. 50. 60. 70. 62. 40. 100. ! TIME l$ttl ~ I l i l t 640. w ! B g (23. . a hm EC3. r wE EEO. or
~~@Y w E60. <~~
~~E w E40.~~
$ 0. 10. 20. 50. 40, 50. 60. 70. 60. Sc. 100. " TIME 15CCI FIGURE 15.2.7-2 Loss of Load Accident, with Pressurizer Spray and Power Operated Relief Valves, Beginning of Life .
~~ggate. Wllh S $ wes on he, S ollo w in 3 p a g o~~
~~\ ci2 0 I I I I i~~
g 5 0. - g & 0.6 * , li e! 0.4 - 20.2 w L
~
~~0 2400 x 2300 - g 22LC -~~
~~= 5 2l00 -~~
d2
~~i ~ 2000 r~~
/
~~y 1900 - 1800~~
~~# 570 h 560 - ! \~~
~~k C 550 - w e. g Suo w E $30 2.00 l.90 - E I.80 - o I.70 - I '*'~~
! I I I 60 O 10 20 30 40 Sk j TIE (SECONDS) - \
~~Tigure 15.2.7-3 Less of Lood Accident with Preuvrizer Sprey end Power Operated~~
~~, Relief Volves, End of Life 1~~
4
~~- 1.20 Y z 3,, ^~~
~~ac ~ WE gE .6< g8 .6<~~
~~~ } $E 4<~~
~~w-Eu 6 w~~
~~.2' L ~~~
0.
~~2. 10. 22, 50. 40. 50. 60. 70. 60. $0. 100. ,,~~
llet ISEC) 2600. < *
~
~~er f. 2400. '~~
~~-N.E.~~
h,p 2222. ORm NW 2000. < E 1900. C. it. 20. 50, 40. ge. 60. 70. 62. 90, 100. 11MC IECCI 600. < C 5:0. < . E ts e. is0. 5'0.
~~$g Etc. "ta: 550.~~
N E40. O. 30. 20. 50. 40. 50. 60. 70. 60. 90. 100. . . TIME istcl 4 E.5 < . 5. E g 2.5 - 2. 1.5 - . 1. P. 10. 22. 50, 40. 50. EO. 70. E2, 93, 103, j f!"E ISEC FIGUP.E 15.2.7 3 Loss of Load Accident with Pressurizer Spray :d Power Operated Relief Valves. End of Life
d Wilowmj P3e l
~~\ ) ~~~
l'
/
~~B }~~
/ 1 / / i, / R & /
~~le 1 h t~~
~~/ 2.~~
~~g A ( -~~
~~!\ -~~
~~R ;- p L~~
~~~ / \ q ~~~
~~E' i/ i i i i i~~
~~! ! ! $ $ $ $ $ $i $ '~~
~~(and Disno) (do) I' wnoA unyx univnssiva annivuidwal sovv3Av 3:03 g e ( s kp~~
~
~~1603.< h 1400. sE e- w 120w.< /.-~~
~~ac 1000. .~~
ME . gU Si . W 600. E 0. 10. 20. b?. 48. EB. 60. 72. 60. 93, 100. 11Mt t$tti W 640. - l g (20. I t yC 600.
~~"8 "~~
~~560. h-k 560. , E 540* 8 e. 10. 20. 50. 40, 50. 60. 10. 90. 90. 100.~~
~~. . TIME l$tCI FIGURE 15.2.7-4 Loss of Load Accident, with Pressurizer Spray and Power Operated Relief Valves, End of Life~~
~~6 k act. Ul- ' LC3geS on. koW/(n3 fa.f (-~~
\
~~1.2 1 I I I l.0 - E IS .. 0.6 -~~
~~?J$[0. -~~
~~b 'A E g = 0.4 - C.2 g f 2600 2500 -~~
~~$w 2400 -~~
~~g $ @ 2300 -~~
,/
~~2 0 - 2000 -~~
~~/ _~~
1900 s , w 570 s g 560 x / g - t = ' g7 550 -- j N - 540 5 M 530 . f
/
~~520 , s 2.00 ,. i~~
/
~~l.90 -~~
~~,/ /~~
~~i.80 -~~
~~/ s E I.70 -~~
1 1.60 - 1.50 0 li 20 30 40 50 TIME (SECONDS) sure 15.2.7-5 Lou of Lood Accident, Without Pressurizer Sprey end Perwer Operated Relief Volves, Beginning of Life [ d 4
~~'h . i 3 1. 2 <~~
~~at ar f. 1.~~
~~] .6< . 6 E" .6< $5 .~~
4 1 r , U !
$ .2- l w >
~~e. 10.. 20. 50. 40, 50, 60, 70. 80 .' 90. 100. -~~
~~TIME ISEC) 26CO. . 2~~
~~'6 G 2400.~~
~~Ng-Ew~~
~~<>m 22c0. -~~
05 l E$ " 4 g 2000. . 1000. O. 10. 20. 50. 40. 50, 60. 70. 80. 90. 100,- I!ME ISCCI 600, n w :. m 583. v WO w -- SEO. Yw
~~-g w e-s70. ;~~
~~85 sco.~~
~~"t 5~~
~~W s50. 540. D. 10. 20. 50. 40. 50. 60. 70. 80. 90. 100. '.~~
. . TIME (SEC1 ,, I 'l 4, 1 5.5 r
5. ce .. cm :- 25 E '*' I
)
~~2. 1.5 - I l 1.~~
~~2. 10. 20. -50. 40. 50. 60. 70. 80..'Ac. 100.~~
~~n' {~~
.11 ME . ISECl FIGURE 15.2.7-5 Loss of Load Accide.nt, without Pressurizer Spray and; 1 Power Operated Reli'ef Valves, Beginning of Life ]
~~t e~~
~~m. .. -., . . . . . . _ . , . _- _. _ . . . . - _ . . . . - . . . _ . _ _ _ . ~ . ~ . . . . . _ . . . _ . . .~~
~~g iM6..~ LA$M 'Ylg(,At'f S- 0/l 4~~
~~.( 0 M ((oWI ?go S- .~~
2 L 1 y j a [ 5 2 ,'
~~m. )~~
~~g_. * ; s. g~~
~~/ s y .g .~~
~~x y e~~
~~./ "~~
~~i-I N =. y- .y~~
's 2C g-o i l .gg .
~~.. .}% t~~
' N s.P. ,/ - . 4 .!. ) / $. l s , / Ts = .54 a>
~~I k l l l. l~~
/ !. g .g ..a ^
7 o ! _I R: e. .. _ l u.) d ; um mn>>smm . annivvami nymuy nos j unym vaziansim e i' = g l. l. 1
~~...- l~~
~~-l' 1400. <~~
~~W-gp 1200. Wu .~~
~~.gm 1000.~~
~~y - g g_ 8e0. m~~
h. 600.

8. le. 20. 58. 48. St. 68.' 78. 60; 90, 188.
~~TIME -ISCCI 640. h .,~~
*N 620. .
~~t b[ wE 600. C 3, 560. E w E 560. . w E W E40.~~
~~0. 10. 20. 58. 48. 50. '80. 78. 80. 90. Ige.~~
~~TIME- ISCCI FIGURE 15.2.7-6 Loss'*6fLoadAccident,withoutPressurizerSprayand-- Power Operated Relief Valves, Beginning of Life r~~
-4
~~gg,-W(h ' M8S Od " O ktlev>,n3pa.y (~~
~~, l l l- I g .0 S~~
l- 0.8 - g 8 0.6 - g 0.4 - 0.2 - 2600 N / w E
~~.g 2400 - .~~
~~W 6 - g 5 2200 - 2L - 2000 - 0 1800 (~~
~~.w 570 / - \~~
~~E 560 -~~
~~/ N -~~
~~f g p 550 -~~
$ ~ Sil0 5 " $30 2.00 / . -\- - .
~~1.90 - E E l 80 - t 0 -~~
~~~~~~~
I.60 0 10 20 30 -40 5 TIME (SECONOS) . Figure 15.2.7 7 Loss of Load Accident,' Without Pressuri er Spray and Power - Operated Relief Volves, End of Ufe .( 4 i
.e
1
~~~O 10' E~~
~~CC - 1.~~
~~.8-a $ .6< ,~~
~~4- {{~~
~~.2 t~~
~~{ 8.~~
~~8. 10. 20. 50. 40. 50. 60. 70. . 80. 90. 100 - - ,~~
TIME ISECl 1 2600. , I E i
~~$G 2400.~~
~~DD . N 2200. 05;~~
. 20 ge - 2000. j .i 1900. '
O. IV. 20. 50. 40. 50. 60. 70. 60. 90. 100. -l TIME isEcl j g 600. 2. p - 590. gg 500. q cc w$ j 570. ; EE vg 560. 4 5 r 550. 540. 1
~~8. 10. 20. 50. 40, 50. be. 70. 80. 90. 100. (~~
~~TIME ISECl~~
~~4. l 5.5 '~~
~~E. E 2.5 ' . 2. I 1.5 -~~
~~1. .. . . .~~
0; 10. 20. 50. 40. 50. 60. 70. G0. 96 .T TIME ISEC) FIGURE-15.2.7-7 Loss of Load Accidept, without Pressurizer Spray.and . Power Operated Relief Valves, End of Life e
~~) Ce lh .~~
~~f-Mllodtn3 77 I I I I 1300 ,~~
~
~~I W~~
~~1200 -~~
~~5 O. Gi~~
*u 1100 5~
~~tt E - R~1000 m w E 900 - l~~
-i
(
~~800~~
/
~~g 610 3~~
=
~~j 600 - w _.~~
~~.k 590 -~~
~~w .E 580~ - - -j~~
. g- .1 670 . m i W .1 - 1 5
W ! 8 550 O l'0 20- 30 40 50 TIME (SECONDS)
~~l Figure 15.2.7-8 Less of Lood Accident, Ylithout Pressurizer Spray and Power- 1~~
.( Operated Relief Volves, End of Life: 1 -1 x < ..
~~1 [~~
~~.( 1400, i~~
~~cr p - 1200. <~~
~~.UU -~~
5" v 1000. UE - [ 600. w .1 g 600.
~~2. 10. 20. 50. 48. 50. 60. 70. 80.- ' 90. j e0, TIME ISCC1' i~~
.1 . =I .: i
~~{ 640. Ce N' 620. W 600. ! wT - C3 E 563.~~
~~.1 w~~
~~E 560. I w , 5 u 540.~~
90. 100.

2. .10. 20. 50. de. 50.--60. 79; 80.
-
~~TINC (SEC)~~
~~... '-i .~~
~~FIGURE 15.2.7-8 Loss of Load Accident, without Pressurizer Spray and: -!~~
~~.- l Power Operated Relief Valves,- End of Life W~~
y
~
~
~~Changos to FSAR Section 15.2.8 Loss'of Normal Feedwater and Section 15.2.9' Loss of Off-Site Power-to the. . Station Auxiliaries (Station Blackout)~~
~~-i i~~
~~I i 1 i i~~
~~^b i -i i .1 9-~~
d
~~^ .M '~~
15.2.7.3 conclusions I Results of the~ analyses, including those in Reference 8 show that the plant design.15 such that a total loss of external electrical load-without a direct er.Immediate reactor trip presents no hazard to the integrity of the RCS or the main steam system. Pressure relieving devices-incorporated in the two systems are adequate to limit the maximum-pressures-to within the design-lialts. - The integrity of the core 1s malatajned by opeatled of'the React'or Protection System,1.e., the DNBR v111 be maintained above the 1.30: value. Thus there w111 be.no cladding damage and no release *of ft.ssion . products to the Reactor Coolant System. . ..
~~' 15.2.8 ~ Loss of Normal Feedwater -~~
i
, 15.2.8.1 Identif1 cation of causes and Aceident Descrintion - I A loss of normal feedwater (from pump failures, valve. malfunctions. or. !
~~loss of offstte AC power) results in a reduction in capability of the . R~~
. secondary system to remove the heat generated in the reactor core. If i the reactor were not tripped during this accident; core damage would-possibly occur from a sudden loss of heat sink. If an altdraative supply. --
of feedwater were not supplied to the plant, residual heat following' t reactor trip would heat DJi mary, system water to the point where water a relief from the pressori SteurJr Significant loss of water from the Reactor Coolant System (RCS) could conceivably lead to core camage. Since the plant is tripped well before 'the steam generator heat transfer capability is reduced, the primary system variables never approach a DNB condition. ;
The following provides the necessary protection'against a-loss of normal feedwater: .
a
1. Roactor trip on low-low water level in any- steam generator.. 1 I*
~~5 :. H N I d i i' :' i h 5 i N - ^" ' " I5 "' I ' " *"3 "' * " W ""' '~~
6 * *
~~,."* " : : _ n r . : r ; - - . r . ". W. .~~
~~ov.......... . . . . . . . . . . . . . ...~~
~~~-' - - --= **' ' ' u" " ' *"* 6 6 " "'~~
~~2, f. Two motor driven auxiliary feedwater pumps which are started on: i~~
~~a. Law-low level in any steam generator- o .~~
~~. i~~
b. Trip of all main feedwater pumps

c. Asy safety injection signal-i
~~[ l 1~~
. 15.2 Coc4/011hr . , t . I3
~~1. O 5081-2 i~~
~~d. Loss of offsite power I k e. Manual actuation~~
. f. Loss of one main feedwater pump and turbine load 807.. I I - ~
3 [. One turbine-driven'auntilary feedwater pump which-utilizes steam y from the steam generators is started on: ' a Low-low level in any two steam generators, or - - - hw.s of,e#sde apeasar'. _ _ _ . _ _ , _ _ _ _ , _
~~b. .... - .... - ..., ... ....... .~~
~~..... v. , _.....~~
~~c. Manual a,ctuation ared flema **.fga eaa em. a adsmueg rep 4e~~
~~e g,4e.*g th{-~~
SMg, # Lee's't'fven l 'The actor.dr avu111ary feedwater pumps M dieselsupp11 are ! generators if afrom loss the ofgfs utilizes To";tr^ J:t' r steam __ .gr
~~.y occurs system. and Both thetypeturbine-driven numps are destoned pump 7- ;~~
1
~~'rt; even 1,f a loss of 61 PAC power occurs. vaea-emergang .~~
~~l simultaneously with loss of normal feedwater. The turbine erhausts the used "q , ,~~
~~--' ; steam to the atmosphere. Tee auxiltary pumps take suction from team '~~
~~Anthe condtnsate4&a. s,$t storage-h % '.tank for delivery 4t 13. x - - d go L the }n4. du generators. . "~~
-T&e analysisa shens that following a Ioss of normal feedwater, the aux 11i J "y feJe water system is papable of removing the stored and. residual heat T4husrpreventing either overpressurization of the Reactor Coolant System or loss of water from the reactor core.
~~15.2.3.2 - Analysis of Fffects and.cp.ngou,e,scel ,~~
~ Method of Analysis .. ....- ODPf'GMJ Y, - . . A detailed analysis usin't the4M0tR (Referen,ce #F code is performed in <:
~~order to obtain the plant transient following a loss:of normal feedrater. The~~
~
Coolant' Sysjpnneluding n,}Amulation the natural describes the plant ctreulation, pressurizerthermal kinetics,- Reactor steam generatorsr. and feedwater systen. The digital' program temputes pertinent variables including the steam generator. level, pre'ssurizer . water level, and reactor coolant average temperature.. c d. Major assumptions are: i .
't; t.". ^. ". ;; ; ^ ;;. . ;;;; :t : .;^ ., le ..; ; ;., .; ; . ;;; ,..,,, . iv. . , .c tY "= ;f ;;;;; t-'r ?: St ; ;;;;;r::ttW ?= !;;;!, ;..., i...
~~h:r n:;; . ..... !; .;'. t;; . - l' l/. The plant is Inttlally operating at 1021, of.. pL% . . .... .MIN. i~~
~~.:-*:; = : et' 7 -? :.~~
~~4 1~~
. 15.2-29 COC4/011SF. l & g __ ,
a
1
$QN-3 ! % ore rest va e atton W g 'I operation at the er lev og the trip. g.
~~4. A heat tr cient in the steam laat System natural cireviation.~~
stoclated with ')- 3 /. Only one motop-driven austilary feedwaterI pump is available one sipute --!d sametadafter 44 ewy the3%Aws..a.be.ba geneeste. ph tenemarbrag 8ede% 8+$a&L e,S g;p g 4. ,,g g .; of/.,AuxiliaryfeedwaterO !!;r; Ig wo a eat ge erators.
,5 f. ' Secondary system steam relief is achieved through the self-actukted , - ' safety valves. Note that steam relief will. In fact, be through the
~ 3 power operated relief valves or condenser.duse valves for most cases of loss of normal feedvater. However, for the sake of analysis '* these have been assumed snavailable. H 5.S*F b f. The laittal reactor coolant average temperature tsadaP lower than .F the nominal value since this results in a greater expansion of ( Reactor Coolant' System water during the transient and, thus In a-higher water level in the p Nssurizer. 88 8 132 _ A t, s.2.e,- 4 _ Figures 15.2.8-1Ashows plar.t parameters following a loss ~ of normal .. feedwater. Following the reactor and tuttine trip from full load, the - j - water level in the steam generators will fall due to the reduction of E steam generator void fraction and because steam flow through the safety j valves continues to dissipate the stored and generated heat.- One minute following the initiation of the low-low level trip. O. n;titrNT 3ea#*ae.as k feedwater pump is automatically started. reducing the rate of water level , , , , , , , , , i E decrease.
~~. ene.~~
r- The capacity of the auntliary'feedwater pump is such that the water level r in the steam generatoist^t;; M does not recede below the lowest level e - . at which sufficient heat transfer area is available to dissipate core k MN',%tJ;tg g rg,e,f, gg4tgCgigr.gety valves.N % , 3 15AS-E % 1gure 15.2.8-1 tt'can be seen that at no time is the tube-shee ~ uncover he steam generators receiving aux 111ary feedwa and that at no time are water reitef from the pressu . . If the a'urillary feed deliver reater than that - actor driven-pump, u the initial reactor power is an the engineered saYeguards " design rating, or the steam tor level in one or more steam- - l generators is above th ow level trip po the time of trip'then 1 the result w.111 steam generator alataum water e her than nereased margin to the point at which reactor shown a - water occurs. q= . .
~~]e4%1ERTC gq M ,~~
~~a 15 ,2-30 coc4/oiisr . 1 y . l~~
i INSERT A The worst postulated loss of normal feedwater event is one initiated by a loss of offsite AC >ower which is described in Section 15.2.9. This is durt to the decreased capa>ility of the reactor coolant to remove residual core heat as a result of the RCP coastdown..
The following events occur upon loss of. normal feedw~ater (assuming main. i feedwater pump f611ures or valve malfunctions): 3
~~'A. As the steam system pressure rises following the trip, the 4 team~~
generator power-operated relief valves are automatically opened to-the atmospiere. Steam dump to the condenser is assumed not to be available. If the: steam flow rate through the power-operated relief valves is not available, the steam generator safety valves'may lift to dissipate the sensible heat of the fuel and coolant plus the residual decay heat produced in the reactor.-
~~8. As the no-load temperature is approached, the steam generator~~
~
power-operated relief valves (or safety valves if the power-operated - . relief valves are not available) are used to dissipate the residual- - -- " decay heat and to maintain the plant at the hot shutdown condition. INSERT B -
~~2. Core residual heat generation is based on.the 1979 version of ANS-5.1 (Reference 9) plus two standard deviations for uncertainty.~~
~~ANSI /ANS-S.1-1979 is a conservative representation of.the decay heat-release rates. 2 1~~
~~. INSERT C- -~~
The calculated sequence of events for this accident is listed in Table. 15.2.-l. As shown;in Figures 15.2.8-1 through 15.2.8-4, the plant approaches a stabilized condition following reactor trip and auxiliary feedwater initiation. x . b ee
~~+ \~~
~~. j~~
. .. - . . _ - - - - - - - - - - ~ - - - - _ _ - - . _ _ - O L . 15.2.8.3 Conclusions Results of the analysis show that a loss of normal feedwater does not adversely affect the core, the RCS, or the steam system since the i auxillary feedwater capacity.ls.such that the reactor coolant water is l not
....trelieved a u. from the ...m pressurtter .... m . ....r.elief.or ... ..a_....safety ... valve.s , a.nd 4he-wetoe m l
i
~~.. . 2..u..~~
~~t 15.2.9 Loss of Off-Site power to the Station Austliaries (Station,, ,,,~~
~~Blackout) l.~~
I 1 15.2.9.1 Identification of Causes and Accide'nt Descrintion ! In the event.of a complete loss of offstte power and a turbine trip there , 1 will be a loss of power to =the plant aus111eries, t.e., the reactor ' coolant pumps, condensate pumps, etc. The events fo11ow199 a loss of AC power with turbine and reactor trip are described in the sequence listed below: -
~~1. plant vital. instruments are supplied by emergency power sources. :~~
, y 2.. As the steam system pressure rises l.Qwing the trip, the steam system power operated relief valve . i".futomatically opened to the .
~~atmosphere. Steam dump to the condenser ts-assumed not to be~~
~~available. If the steam flow rate throu~~
~~' 1s not available. the steam generator. ;.gh~~
the power n'....' relief a safety valves valve's may lift to dissipate the sensible helt of the feel anti ecclent plus-the residual ghgat produced in the reactor. .
~~3. As the no-load erature'Is-a power.eperMJ~~
~~[ ret tef valves (or O. x" _...pproached, the steam '~~
...; safety valves. if the power apsestd rettef valves are not available) are used to dissipate the residual Jeug heat and to maintain the plant at the hot shutdown condition.-. 8 -
~~4. TheNSJbWelel gener'ators st;rted on loss of voltage on the~~
~~plant emergency busses y begin to, supply plant vital-loads. i' l~~
The avrillary feedwater systin is started a 11y as discussed tu > the loss of norni feedwater analysis. tven aus111a,'y r . feedwater pump utilizes steam from the the atmosphere. The actor driven aux 111a systemandexhauststo 4 j by power from the diesel gener'ators. The, r s are sup tion f.;..". plied '7: fron'the condensate storage tank for delivery to the steam generators, , I i
.i e \ / i I
~~i a 15.2-31 .$ COC4/0115F ! 1~~
~~, ... 1~~
4
~~h. SQH !~~
g NsE5rt "D3 l Upon e necessary for core coolin reactor coolant b 1 heat is maintained n the reactor toolant loops. 15.2.9.2 Analysts of Effects and contenuences Method of Analysts. - ' M,*E th ocus 4 - 44 1 AtM ;tdetailed analysis following r;'. ;;. d.... . .. using'e,.e 9tROtM a station b (Reference /)lackoutcode
~~.. The simu-T lation describes the plant thermal kinetics. RCS.lacluding the. natural --~~
ctrr. elation, pressurtaer, steam generators and feedvater system. The - digital program computes pertinent variables including the steam t generator temperature. level, pressurlaer water level, and reactor coolant average
, @gggy g} , .The first few seconds et the transion a 1c@slyresembleaslaulation of the complete ^1oss of flow incident (see Subseetton 15.3.4).1.e., core damage _due to rapidly increasing core temperatures is prevented by -
~~promptly ri ping the reactor. After the reactor trip, stored and residua 4 t must be removed to prevent damage to either the RCS or the~~
~~$5 3.$" ' ' ~ ~ ~ ~~~
15.2.9.3 Conclus1ons j [' , yk h ]g Results of the ' Complete Loss 'ofi;;S:tr" Forced;;..h.i. Reactor Coolant;3.;,h Gee;;... Flow" analysis (Section show that 'r . '.;;; c' M' r
'.C pewee no 64 verse. 1 ions occur in the reactor core. The DNBR 1s maintained above e Reactor C wlant System is not overpressurlaed and no water relief will occur through the pressurlaer relief or safety valves. I damage and no release of fission products to the RCS.-Thus there will be no cladding-15.2.10 Excessive Heat Removal Due to Feedwater Svstem Malfunctions 15.2.10.1 Identifteation of Causes and Accident Descrintion Reductions in feedwater temperature or additions of excessive feedwater are means of increasing core power above full-power. Such. transients are - attenuated by the thermal capacity of the secondary plant and of the RCS. The overpowse - overtem overtemperature and overpower.perature protectioc (neutron overpower. .
which could lead to a DNBR less than 1.30.AT trips) prevents any. power.lacrease Escessive feedwater flow could be caused by a l'ull opening of a feedwater control valve due to a feedwater control system malfunction or an operator error. At power this excess flow causes a greater load demand on the plant at RCS due to increased subenoling in the steam generator. With the I
~.
~~15.2-33 COC4/011SF~~
~~% 9 f~~
~~lt 1~~
. INSERT D-l Following the RCP coastdown caused by the loss of AC power, the natural i circulation capsbility of the RCS will remove residual and decay heat from the ,
core, aided'by auxiliary feedwater in the secondary system. An analysis is - presented here to show that the ettural circulation flow in the RCS following.a loss of AC power event is suffh 'ont to remove residual heat from the core.- l INSERT E- .. -.
~~. . ~~~
~~l Major assumptions-are:~~
~~1. The plant is initially operating at:102% of 3423 MWt. q~~
2. Core residual heat generation is based on:the 1979 version of ANS-5.It j (Reference 9) plus two standard deviations for uncertainty. ('
~~ANSI /ANS-5.1 1979 is a conservative representation of the decay heat; .- release rates.~~
~~3. Only one motor-driven auxiliary feedwater pump is available one minute -~~
~~after the-low-low steam generator level signal is. initiated in any steam generator. . _ l~~
4. Auxiliary feedwater flow rate of 410 gal / min is split unifomly between two steam generators. ;

5. e temperature is 5.5'F lower than' The-initial the nominal reactor value since coolant avera$1s in a greater expansion of Reacto]
this resu : Coolar,i. Systen Date during the transient:and a higher water level:in the j pressurizer. i l 15.2.9.3 R32011 , The transient response of the RCS following a loss of AC power is shown in )
. Figures 15.2.8-5.through_15.2.8-8. The calculated seque7ce of events for this event is listed in-Table 15.2-1. , } j i
i . l -l 1 1 I
SQN
6. Turbine Load - Turbine load was' assumed constant untti the electro-hydraulle governor drives the throttle valve wide open. Then turbine load' drops as steam pressure drops.

7. Reactor Trip - Reactor Trip was initiated by low pressurizer pressure assumed at a conservatively low value of 1775 psia.
Result _s - The transient response is shown in Figures 15.2 I'4-1 and 15.2.14- 2. Natclear power starts decreasing lamediately due to_ boron injection but steam flow does not decrease entti 15 seconds into the transient when the turbine throttle valve goes wide open. The alsmatch between load an'd ' acclear power causes T..., pressuriger water level, and pressurlaer pressure to drop. The low pressure trip set point is reached at 64 : seconds and rods ~ start soving into the core at 66 seconds. After' trip, pressures and temperatures slowly rise since the turbine is l tripped and the reactor is producing some power due to delayed neutron fissions and decay heat.
~~15.2.14.3 Conclusions . <~~
Results of the analysis show that spurious safety injection with or without tamediate reactor trip presents no hazard to the integrity of the Reactor Coolant System. - DN8 ratto is never less than the initial. value. Thus there will be no cladding damage and no release of fission products to the reactor coolant ' system. :: . If the reactor does not trip immediately, the low pressure reactor triiP' " ' l will be actuated. This trips the turbine and prevents excess cooldown thereby expediting recovery from the incident. 15.2.15 References I
~~1. -W. 'C. Gangloff "An Evaluation of' Anticipated Operational Transients ;~~
~~in Westinghouse Pressurtzed Water Reactors," WCAP-7486, May 1971.,~~
2. D. B.-Fairbrother, d. G. Nargrove, " NIT-6 Reactor. Transient Analysis Computer Program Description " WCAP-7980, November 1972. ;

3. C. Munin, "FACTRAN, A Fortran Code-for Thermal Transients in U0, Fuel Rod," WCAP.-7908, June' 1972. - '

c. T. ". 4 -h tt, O. O. ",2 ?? - . *. *. ".:::, "'.07T7f.T on, tp :. ' C,,.
~~e 9.. ".". . e , r s_ m R~~
~~4. hd4/r,W:T.g a\,, 'LoPreM cab /,pb," ucersq-M l (Fr*P0MhWCA?-7907-4 (A.' Rop /dgp) Ap/,(. IoM, , .~~
~~15.2-44 C004/0115F~~
*'"a * , ,y
e
. . y sgn.5 .
~~b j'~~
, 5. . S. Altomare, R. F. Barry, "The TURTLE = 24.0 Diffusion Depletion ;
Code " WCAP-7758, September 1971. - ( 6. F. M. Bordelon, " Calculation of Flow Coastdown Af ter Lost, of Reactor Coolant Pump (PH0ENIX Code),' WCAP.7969. September 1972.-
7. J. M. deets, ' MARVEL - A Digital Computer Code For-Transient Analysis of a Multiloop PNR System ' NCAP-7909,'Jone 1972.

8. M. A. Mangan, " Overpressure Protection For Westinghouse Pressu.r13ed 9F Water Reactors
~~NCAP-776), October 1971. .-- 1~~
/
~~9. P::~~
~~2.u.". ,0.:. .;,~.t~~
.. -- . . . -.. . . "_,e. .. .m. . " . - -,, , ;. .;; . ' . 'S_ . . '. _ - .'. . .r : ? ::t l. ' :t;;; ? : " :;7::
o
~~.y_ a o .~~
10. ' Power Distributton Control and Lon:f Following Procedures,' '
~~WCAP-8403, September 1974. b-~~
~~< < _v l~~
~~l ..~~
~~%D . k gW WAe AJors, "ssr/Ms h r.t -m%J MT5~~
~~t. * .~~
I l - J
~~. 1~~
./
k I L . ] .
=
) - 1 -
~~.{ .I g~~
i 15.2-45 CCC4/0115F l V 1
& 4d . g g w y =- - , , , .e.. ,-w-w,-- e.--..,.,,,. ..-.-w,w,,,.m-ei........-..-..,._,... .,,...,4
l 50N-3 , TABLt-15.21 (Sheet 5) (Continued) , 1 TIME SE M NCE OF EV NTS FOR I CON)l" ION II EV [NTS. i Accident - [g,n}1 . . Time (Sec.)
.]
~~Loss of External "~~
~l Electrical Load (Cont.) * '
~~4. Without pressurizer . .~~
control-(EOL) Loss of electrical load 0 Initiation of steam release from steam generator safety, valves 7.5 ' litgh pressurizer pressure . reactor trip point reached 5.5 -. Rods begin to drop 7.5 ,
~
Minlaus DN8R occurs (1) , , Peak pressurizer pressure occurs 7.0 , Lo Normal Feedwater Low-low steam. generator water " and Loss Off-site level reactor-trip; reactor Power to the ton- coolant pumps begir. to c( , Austliaries (Stat . down 0~ Blackout)
. Rods begin rop ~ 2 MSEW. at f ,
s generators begin to receive.au ry feed from one motor-driven au ry feed-water pump - 60
~~Peak water level in pressurlae occurs , ,~~
e ; Excessive feedwater at full load One main feedwater control-valve; - l
-i . falls fully open 0- 1 Minimum DNiiR occurs 15.2 :
Feedwater flow isolated due to-high-high steam' generator level. 14.0 (1) DNBR does not decrease below its initial value. _. 3 Revised by Amendmerit 3 4 COC4/0723F i a
~~>= - '~~
~~e e ww. -,ww w -c,w. ,,=-,--e---- --4m - _mm__ w- u-. m_u- __ _ _--_._s - - - -~~
~~- - +_ . _ _ _ _~~
~~3 INSERT F-Loss of Normal Feedwater Main feedwater flow 10 stopped-Low Low steam generator 58-~~
~~' water level trip-setpoint ^~~
reached ! i Rods begin to drop- ' ,, ,60 -- Peak water level in the 644 pressurizer occurs - '
- :t Two steam generators 11g i begin to receive auxil- i tary feedwater from one "{
~~motor-driven auxiliary -- feedwater pump , l Cold auxiliary feedwater -556-~~
~
~~is delivered to the' l~~
~~. steam generators - "~~
f
~~.q - Core decay heat plus pump -6200.~~
heat decreases to auxil-iary feedwater heat ~ remova,1 capacity i 1 Loss of Off-Site Power Main feedwater flow- 10: to.the Station Auxiliaries stopped :
~~, 1 Low-Low steam generator 59- ;~~
~~water level trip setpoint 1 reached-Rods beginlto drop~~
~~61~~~
~~Power lost to the reactor 63 coolant pumps' * -~~
~~- Two steam generators ' 120 1~~
begin to receive auxil-iary feedwater from one motor-driven-auxiliary 'l feedwater pump i Cold auxiliary feedwater 557 : l is delivered to the steam generators - Core decay. heat decreases -1500 to auxiliary feedwater~ heat removal capacity Pea ~k water level in 28 4 pressurizer occurs
~~- - i~~
SQN . TABLE 15.2.9-1 NATURAL CIRCULATION FLOW - Power Natura treulktion'- " -- lPereat
~~v. p.re.nt -~~
~~5 4.0 5.22~~
~~. . '3.5 4,g4 1~~
~~3.0 4.60 2.5 4,37-~~
~~. , l 2.0 ,o4 i.5 1/-~~
~~3.. " l 3.n It - a* .~~
}.
~~t ' a 11 e i~~
~~'l COC4/0571F N r~~
~~.I - ([ ' .~~
~~sto~~
~~; ; ; g i /~~
~~w .-- 0 - - 600 - g 590-~~
~~- - )~~
~~l ,, - - l~~
'570 i -@gr 1600 [ - .i
[ g gas 00 - h 1200 r_ e Im0 . -
\
~~2400 y~~
~~/ -~~
~~17 E 2300 \ ( -~~
~~.N .~~
~~i g 22= i,2l00 - i~~
. . . . - - 2000 40- / \ ; 30 - '!b sisAit stataalces atetivlas c Wg 20 Auxii.taar Fttomatta - ,
~~g ,, sitam stateatoes not sectivias~~
~~' AU11 W ).laaf FEE 9Batta . 0' ,~~
l l. - l 1 0 1000 .2000 8000 88000 5000 6 0 T.16E .(SECCIIDS) 71 re 15.2.51 Core Averaae Temperature. Steam Generator Water Level and Pressurizer Water . Volume as a Function o'f, Time. ..' (. Loss of Nomal Feed Accident-Vauc.e LeenA wic; A60265 l 5.~2. E- ( rHe.casR 15.2..B-8
--5 .1 .o . .j i
I.. l 0 8.8 g - i
~~. b~~
1 .
=
~~t .. . . 1 E 4' i~~
~~.4 .~~
348 tel. . SGI '308 48*
~~. . ting ISECl~~
l i.. i ( -I l g i.
~~. p" .E-~~
~~i. ~ _~~
~~s .~~
^
1
- $8 y .s j - 1 L
~~p .. .~~
~~.d -~~
e
i
~~, , . i- -~~
V .- , -- t 1 '- e.
~~.: i.e ... .~~
1:a . eci , Figure 15.2.8-1.aLoss Function of Time of Normal Feedwater Event, Nuclear Power and,RCS Flow a 8 . 9 .y .,- B s
e . 4: 8800.
. . .i -i .a . e . . . .' .c_
~~8400. < .~~
~~.l . ./ . . .~~
~~g ... g,< .~~
~~g. * '~~
g ... . . . . u " 1 E . 8 00 0. < *
~~. g $499. =~~
~~148 tel . 808 . 888 898 I~~
, itRE (SEC8 . .' I - 1 I
w St e t. < = ' 4 g test. < , t , S t.o o. < ' 8 ..... Y stes.
~~3000.~~
g ' e"- g .... ac *
~~, -i k .i ses. .~~
~~k g .... E ,,.. .~~
e. I 848 ISI 408- 308
' 864 fint segCI ~,
~~Figure 15.2.8-2 Loss of Normal Feedwater Event,. Pressurizer Pressure and .~~
)
~~Pressurizer Water Volume as a Function of- Time i~~
*= .
~~@& t '!//Mo g k)/ IMAGE EVALUATION ///jf ,,5 4~~
~~4) .~~
~~1Es1 1 = eEr <m1.a>~~
~~,, jp \ ' Ru R a l.0 5 5 IL2a S u Ef M a~~
~~1.25 1.4 1.6 4 150mm >~~
< 6" *
~~k.,,,,~~
~~___ ., S WEBST R C Y 14580~~
~~+ +9 O~~
L
/
~~ll/I/4 0~~
%* IMAGE EVALUATION S# //jjg//
f 4 TEST TARGET (MT-3) #4 4)%ys,+,/ '*%j/ ef ,
\
I.0 5E2 BE I y lj IL?4 l-l ? u m Me r 1.25 1.4 1.6 4 150mm >
~~< 6" >~~
~~b 4%# ' 4<$//p o%g,,,,~~
~ ' *""? =n' : " ""' " ')g+kf+)[*
~~WEBSTER NE YORK 14580 (716) 265 1600~~
*4*&4 I////fo IMAGE EVALUAtlON g) d,@9 s 1Es1 meEr <=12) 77,p ///// ,f p4 9 + .
l.0 !!EE E yll llE l,i imILM l.8 J I.25 1.4 g 4 150mm O 4 6" *
~~?5'L / i~~
~~,,W , _=_~~
mesjggg,eeo .+
4+ OL
$*sf IMAGE EVALUA,10N , ,, /// p ,4 g/gN@ + re 1.0 'd m Ha iEET ylj llu y if EB L8 ==
~~l.4 1.6 4 150mm >~~
~~< 6" >~~
~~+sk % i/ / 'b~~
~~+ Q ,,,,S//j~~
~~_ = ,,s , _ $A..;;.4 WEBSTER NE YORK 14580 (716) 265-1600~~
. . __-.- .-.- _ ..- _ _ _ _.___ _ _ . . _ _ - _ . _ _ _ _ , _ _ _ ~ . - - - _ - - _ _ _ . _ _ . - _ _ _ _ _ _ _ _ _ . - - _ - - _ _ _ . - _ _ _ _ . _ . _
~~l 4 i l~~
.o ... ...
I l . . see. .. '.l. , a c .s . . ST. _-. " $s'... - [
~~.g,,...~~
~~war~~
~~s. ..~~
(: . .00 . '
~~! W see.<~~
I g see.; cobb l a - See. . Ste. ' . l' 600. .. 508 403 408- 308 304 s TIME ISECI *
~~. e ,St.~~
r l c .... sur - f e.... . g .... or - - - E see. - co
....... _ _ . . 3,,,,,. . . --t. . . . ,
~~Ste.~~
500. i 18' . 468 308 338 god
~~, ;[~~
~~p' ,_.~~
. . . TIME . . t t't t i ' ~
~~Figure 15.2.8-3 Loss of Normal Feedwater Event, Loop 1 and 3 Cold Leg,- Hot Leg, and-Saturation Temperatures as a Function 'of Time l l l' l~~
- - . . ~ - - - . . .. . . . - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - . - -
~~s-1~~
. -l s-l ? .l ,- t 1600. , g lete. E'
~~(-. ,,,,, . I e g .... i g i W .....~~
~~. i.~~
~~t I I 4es.~~
~~W 864. .~~
I
, . L , s.
~~I 808 89I-843~~
~~, le* 48 s.~~
i[
~~. flat ettC8 ** '~~
l I
.14008'6 l .. .s ! .Ittet+6 b . ieser.6 i
m . g .
~
E.......< . STEM damas4 mas REce.tVmad i Auxeugty y ...eef.s Faspionysg W y . y .****t**< J> rem 6 EmeRS M i K raus % A w s u A n y m e d
.eeeer.s **' - so* ee ,,a ,, . ,
_. vaat - este. . 1 Figure 15.2.8-4 1 Loss of Normal Feedwater Event, Steam Generator Pressure and i Steam Generator Mass as a Function of Time
~~.1 l~~
~~l l i~~
.G b b ,
w- v --r -- -, ..e.. . . , , . -.m.-- . . . . - ,. 2- re <- .. , - - w.+e., e. .--..vm , .m. --.I
~. . . _ . . - - - . . - . . , . .- . .. j ' l ..1 )
~~l i i d~~
. j 4.4 l . .+
t
,48' **
~~g .~~
s g 3,. e ... . .I
~
~~g .. . g . 4~~
.t - . 8. ,. ,, g,3 + get , 848 <- 88 #
~~Ting IStti . l e s.4 I g i.8 .~~
~~t. p '~~
~~l s I~~
\
~~8. .~~
~~-( .. h o~~
~~g .0' ,~~
~~.6-en .N 4, d~~
E
~~".8' .~~
~~8. 448 itI 198 438 10 8~~
~~, ting setts - ~~~
~~Figure 15.2.8 5 Loss of At Power to the Station Auxiliaries Event, Nuclear Power and RCS Flow as a Function of Time e l ..~~
~~. e.~~
\_ t 1 i ( ( . , , , ..-. . i [ , e.... L
* ~
l g tete. , [ g' , l =. ' i l ..... -
, f '
l
. E e.. . . . .g 1 .
m g...... . . E . . 3000. < ' 1688. 308 808 ,308 - 308 ted' fing 'estte . 2000 .- 7. j y seee. ' a i l M 3600. . - 7 r h \ l $st
~~;h 909.~~
400. *
~~'7 800. <~~
~~e. -~~
ej i.' ..' ... . Tint gegC6 Figure 15.2.8-6 Loss of AC Power to the Station Auxiliaries Event, Prc.;surizer Pressure and Pressurizer Water Volume as a Function of Time
. f a
v - + w - .-,-vs.,-v -,v.,.. . . m.~,n., ,,----.-#,-- --- --. ,.e -- -- , ,- ,
~
o' , 9... , sst p s 4.. " g.... ,7 i
~~. w..... .~~
~~6...~~
~~Cohb - '~~
~~l...' 6... gg. 363 - 388 - II' ' ftME' ' SECS~~
~~?...~~
~~l .... '~~
~~.... SAT -~~
~~w ....~~
$(b7 = .... . ;
w..... ' i e t
~~. ... I s .. ..~~
~~CO L.D~~
~~64.. .~~
s s... . gg. a. get att .08 .4 ttnt tttCl Figure 15.2.8-7 , l Loss of AC Power to the Station Auxiliaries Event, Loop 1 and_. 3 Cold Leg, Hot Leg, and Saturation Temperatures as a Function of Time I 4 (
e l
'9 . 1 0 -
8640. q J 1800 4840. ' m 8000. -
. g .... . l 8
w ....< . W . g .... . m - ees.
e. (
888' 408 309 se8 go e ting estts
.L -i .4406E.6
i
.34098 6 !.,,,,,.. . me u-w '
~~! w rec.Etui% Apot.mRy~~
~~I FEEDUATER I~~
2 w ... . .. : 8 1 . g ......., m m4.s== = wr - L :;; - R6CEtM AumungnebuesA , 7
..e..... J '~ "' . **8 i.e ... . ,,. _
vine essei Figure 15.2.8-8 1 ! Loss of AC Power to the Station Auxiliaries Event, Steam l Generator Time Pressure-and Steam Generator Mass as a Function o i 1 I .
- om-__. _ _. , ,. # , , , , ._._,,,,o, - - ,,_ ,y,. ,-w., g.....,-,,, , ,,7,_ , ~w.. - _ . - . .,y . . . . . , . ,..7,. , . , , -
~~1 i Changes to FSAR Section 15.2.12~~
~
~~Accidental Depressurization of the Reactor Coolant System i i 9~~
~~- 2.~~
~~l- r ! a t~~
~
l . . , a s 4 '9* 4 h
k SQN-1. Figures 15.2.11-5 through 15.2.ll-8' illustrate the transient assuming the ;
.( reactor is in the automatic c, trol mode. Both the beginning of life and ' , the end of life cases show that core power increases, thereby reducing the rate of decrease in ccdiMt average temperature and pressurizer pressure. For both the beginning of life and end of life cases, the ;
~~minimum DNBR remains above 1.30. 15.2.11.3 Conclusions It has been demonstrated that for an excessive load increase the minimum ..~~
~~-i DNBR during the transient will not be below the limit of 1.30. -1 1~~
~~15.2.12 Accidental Depressurization of the Reactor Coolant System~~
j 15.2.12.1 Identification of Causes and Accident Description The most severe core conditions resultitig from an accidental-depressurization of the Reactor Coolant System (RCS) are associated with -
~~an inadvertent opening of a pressurizer safety valve. Initially the~~
~~, event results in a rapidly decreasing RCS pressure until this pressure <~~
reaches a value corresponding to the hot leg saturation pressure. At that time, the pressure decrease is slowed considerably. The pressure j continues to decrease, however, throughout the transient. The effect of the pressure decrease would be to decrtase the neutron flux via the - moderator density feedback, but the' reactor control system (if.In the l automatic mode) functions to maintain the power essentially constant i , i throughout the initial stage of the transient. The average-coolant temperature decreases-slowly, but the pre.ssurizer
~~
~~level increases until l/ l .the reactor trip. CW. Cf The reactor will be tripped byAthe following Reactor Protection System signals: ,~~
1. Pressurizer low pressure '

2. Overtemperature AT 15.2.12.2 Analvsis of Effe:ts and Consecuences Methods of Analvsis The accidental depressurization transient is analyzed by employing the detailed digital computer code LOFTRAN (Reference 4). =The code simulates the neutron kinetics, Reactor Coolant System, pressurizer, pressurizer relief and safety valves, pressuriter spray, steam generator, and steam
generator safety valves. The code computes pertinent plant variables including temperatures, pressures, and power level. The core limits as-illustrated in Figure 15.1.3-1 are used as input to LOFTRAN to determine ,
the minimum DNBR of the typical or thimble cell during the transient. { . 1 1 15.2 N.,7 COC4/0115F
SQN-6 i In calculating the DNBR the following conservative assumptions are made: ,
~~1. Initial conditions of maximum core power and reactor coolant~~
~~, temperatures and minimum reactor coolant pressure resulting in the minimum initial margin to DNS (See Section 15.1).~~
2. A reto moderator coefficient of reactivity conservative for

beginning of life operation in order to provice a conservatively low i amount of negative reactivity feedback. due to changes in moderator density. The spatial effect of void due to local or subcooled $
l boiling is not considered in the analysis with respect to reactivity feedback or core power shape. l 3. A high (absolute value) Doppler coefficient of rsectivity such that the resultant amount of positive feedback is constrvatively high in order to retard any power decrease due to moderator reactivity feedback.
~~4. The pressurizer safety valve capacity is asstmed to be 10% greater 5 ,~~
to increase the depressurization rate. ' It should also be noted that in the analysis power peaking factors are kept constant at the design values while, in fact, the core feedback effects would resul.t in considerable flattening of the power distribution. This would significantly increase the calculated DNBR; however, no credit is taken for this effect. Results ru&os=.pcw Figure 15.2.12-1 illustrates the h ^ transient following the accident. Reactor trip on overtemperature AT occurs as shown in Figure 15.2.12-1. The pressureNtransientsduring the accident, including a 101 increase-in capacityuff given in Figure 15.2.12-2. The resulting DNBR never goes 5 below 1. F as shown in Figure 15.2.12-3. I 15.2.12.3 Conclusions Ord Waar\ 036 NO\d NWO\'E@MC
~~The pressurizer low pressure and the overtemperature AT Reactor- .~~
~~Protection System signals provide adequate protection against this~~
~~. . accident, and the minimum NNBR remains in excess of 1.30, 15.2.13 Accidental Depressurization of the Main Steam System /~~
15.2.13.1 Identification of causes and Accident Description-The most severe core conditions resulting from an accidental depressurization of the mairt steam system are associated with an inadvertent opening of a single steam dump, relief or safety valve. The l analyses performed assuming a rupture of a main steam pipe are given in Section 15.4. 15.2-3Bd COC4/0115F
~~-w . _ , -. - - - - , , , - ..-.,.,c r .~~
, 1 l l L ; , SON-3 1
4 TABLE 15.2-1 (Sheet 6) I h (Continued)- ! TIME SE0VENCE OF EVENTS FOR ! CONDITION ]] EVENTS Accident Event Time (Sec.) I Excessive Load Increase .-
~~l. Manual Reactor 10% step load increase 0 ,~~
~~Control (BOL) Equilibrium conditions reached (apprortmate times only) 200T~~
~~2. Manual Reactor Control (EOL) 10% step load increase 0 Equilibrium condt'tlons reached~~
~~, (approximate times only). .50 <~~
~~3. Automatic Reactor 1~~
~~Control (BOL) . 10% step load increase O' ! Equilibrium conditions reached (3) 1 4. Automatic Reactor~~
,. Control (EOL) 10% step Icad increase -0 Equilibrium conditions reached (approximate times only) 50 l Accidental p pressurtration of the Reactor Coolant System Inadvertent Opening of one RCS Safety Valve O- ,
~~I eagtgr rip CA @CAgrMurGeaf t/b.T~~
~~.. Minimum DNBR occurs. 95Mr % .S Accidental depressurization .~~
6 \n 40 dfCg") %.'b i of the Main Steam Safety System Inadvertent-Op ning of one main steam safety or relief valve O Pressur.12er~ Empties 160 20.000 ppm boron reaches RCS loops -197 UHI initiation time 221 (3) Did not reach equilibrium within the time scale of Figure 15.2.11-2 Revised by Amendment 3 ', l COC4/0723F
.[
~~1 kep\acc WM 6 ure on L l~~
%c fo\\ouowg y c. . )j l 1 . -1 1
l , NUCLEAR FLUX (FRACTION OF NOMINAL) l m ? ? .9 9 .~ .~ . i I ec o a ~
~~- .. . . ~~~
2 ' l I I I I f - l N e r 1 g 1 k5 t-m
=
~~l _ , me ' em e w g3 - 1 i i~~
1
=
~~1 . t . R, R8~~
=*
m F 1 ,g _ : I e. 8 4 N 1 1 ! l
~~l. \~~
~~Z E-SC 94 4 l~~
~~-l l~~
I
~~. - . , , . : 7 :.- .--. .... :.,..-.~~
~~h 1 l~~
. l l
~~I J l~~
-0 L i - 1. 4 < . .;
\ . g 1.2 < -
~~-g 8'~~
g ) ,. t i W .8 - e w k .6 <
. W -E g .4 -t .2-c.
~~2. E. 10. 15. 20. 25, 50. ' 55. de dS c 50. 55. 60. s~~
! 11ME ISEC I' e . vipe. 15. 2 .12 - 1 Nul**r Og Ac.didehtal 1)e.pcesswiw+. ion q m
~ . 3-I L .t=
~~i 1~~
, . , , . - - - - - - - - - - , - ,-S.-r .-, , - . < , a y ~ - ,.t'
l . 7636 33 ( . 8 t. a I- . I _ yg l . l
g. -

c. ]
t w e;.
,3
~~d .~~
I I I I I - .. r E : f a k k _ k_- ._ - k. . E (visd) Juusud a zius saud 1 i ( 7byt\ bucosso M vo sid3 Wcn wo\
~~. a 1 7-= =e -y w ,,--..-.,,e-.,-a,-- __m-__________= e .,,--w-c=^ws.. --._ ww,+= .s-m-s~~
1
~~. 4 j~~
~~2400. <~~
~~$2200.~~
~~e 2000. , 1 1 m 1900. 1 {~~
J
= 1600. + .M. -
~~gi.08. o , i k1200. 1800. < 600.~~
~~0. - 5. 10. 15. 20. 25. 50.-55. 40. 45. 50, 55. 60. ,~~
TINC (5CCI t 620. < 600. . l x. C 500. , e 1 1 8 SEC. 5 l ' \ s A 0. < 1 520. 1 500, t
~~9. 5. 10. 15. 20. 25. 50. 55. 40. 45. 50. 55. 60., , ;~~
~~TINC (SCCI~~
~~.B op n. w.a.\a $ 9(esw\2cr 9(ecentc Gnd a\wc 1amsMs :~~
fbo\och for .Gleesdato Temp \ Vegenud20Aton ! l 5
t 6 D\ M f5-O n S X e q yoge ) l l DNBR , Y " "
~~~. ' '~~
~~u u ."a e o m e .-~~
o^
~~u , o l l l l l l' l l k.~~
w
,~ o - . 0 4 6 ;;; - =
a b I l N w ' w Eo - U
~~$_. g i~~
~~S '~~
)
~~a g f h. t 0, -~~
~~= ~~~
~~f N I~~
~~i. a 1 L~~
~~g _8 - o L 4~~
),
tit St94 , E l l -
E
~
~~4.- 5.5 < ..~~
~~5. <~~
~~2.5 -~~
2. - ,

1. 5 -

1. -j D. 5. 10. 15. 20. 25. 50. 55. 40. d5. 50. 55. 60.
~~' {' TIME - tCEC)- - 1 I i~~
~!
~~( l~~
~~, G ,)* )&Ok fCLO$l60 O .~~
~~Accidenlo / Def r* ** U 'I'" ; l~~
~~.. 4 s4" . i 0~~
1 . m
h 9 Changes to FSAR'Section 15.4.2.2
~
~~Major Rupture of a Main St'eam Pipe~~
~~.T L~~
~~k f 4 b L a I i~~
~~.e. .~~
e 1
SON-6 , . For Case D (Inside break with loss of offsite power), the point most i likely to have the lowest DNBR is the point with the highest power / flow ratio. Usually, this point is the one with the highest. power. As with Case B, either the preceding or succeeding point is also analyzed. 3hould any of the points analyzed result in DNBR's near 1.30 additional-points may be analyzed to insure that the point with the minimum DNBR condition has been analys:ed. The points analyzed for this application had DNBR's greater than-1.30. " Thus, it is concluded that the minimum DNBR for a steam break is greater than 1.30. - The maximum linear heat rate for the most Ilmiting- s'teambreak case presented in the FSAR was less than 10 kW/ft, which is less than the linear heat rate which results in fuel melting. There islno known j failure mechanism associated with this peak linear heat rate. -; 15.4.2.2 Maior Ruoture of a Main reedwater Pipe 15.4.2.2.1 Identification of Causes and Accident Description A major feedwater line. rupture is defined as a break ir i feedwater pipe large enough to prevent the addition of sufficient feedwater to the steam generators to siaintain she11 side fluid inventory in the steam generators. If the break is postulated in a feedline between the check valve and the steam generator, fluid from the steam generator any also be discharged through the break. Further, a break in-this location could preclude the subsequent addition of auxillity feedwater to;the affected s' team generator. (A break upstream of the feedline check valve would affect the Nuclear Steam Supply System only as a loss of feedwater. This > case is covered by the evaluation in Subsection 15.2.8.) Decending upon the size of the break and the plant operating conditions at the time of the break, the break could cause either & RCS cooldown (by excessive energy discharge through the break), or a RCS heatup.. Potential RCS cecidown resulting from a secondary pipe rupture is evaluated in Paragraph 15.4.2.1, " Rupture of a Main Steam Line." Therefore, only the RCS heatup effects are evaluated for-a feedline 6 i
~~. . rupture.~~
A feedline rupture reduces the ability to remove. heat generated by the core from the RCS because of the following reasons:- ,-
1. Feedvater to the steam generators is reduced. Since feedwater is subcooled, its loss may cause reactor coolant temperatures-to increase prior to reactor trip;

2. Liquid in the steam generator may be discharged through the break, and would then not be available for decay heat removal after trip; :

3. The break may be large enough to prevent the addition of any main feedwater after trip.
l \ 1
\
~~l 15.4-28 .0117F/COC4 , l g yw-ge'r- +e -n ,-4.w+.- ---m-v-e-.w~~
~~%-c- * --- _ ,,y., s-- -+ e -w.i, -- , we-e, *,v--- - .-4 ,n-m mm,q-,,g .M t-w~~
~~.q d~~
1 a s SQN-6 I (, An Auxiliary Feedwater System is provided to assure that adequate feedwater will be available such that:-
1. No substantial overpressurization of the RCS shall occur; and

2. Liquid in the RCS shall be sufficient to cover the reactor core at p/S G"RT all times.
WRITE UP T - FRoM The following provides the necessary protection against a main feedwater pot,u) Wig, upture; , PAfrE '
1. A reactor trip on any of the following conditions: "

4. High pressurizer pressure.

b. Overtemperature delta-T,

c. Low-low steam generator water level in any steam generator,
%. Le- :t::r ;;r.;r-j w sevel pive m m ...!Te;G Tk Ci;; ;ct,17, rl/ -- - , a y e ::er ck Safety injection signals from either of the following:
1. Low steam line pressure b ; cine;;ent .;Ui . wh iteea. Tiw

11. High containment pressure
~~'<i dt ;t, ;;;;; ' he p <r r.eHe a1 n e :;.7; (Refer to Chapter 7 for a description of the actuation system.)~~
( 2. An Auxillary Feedwater System to provYe an assured source of . t feedwater to the steam generators for decay heat removal. (Refer to l Section 10.4.7.2 for' description of the Auxillary Feedwater System.) 6 15.4.2.2.2 Analysis of Effects and Conteauences Method of Analysis LM=TNJ 2A A detailed analysis using the Mub.-(Reference 2f) code is performed in order to determine the plant transient-following a feeditne rupture. The code describes the plant thermal kinetics. RCS including natural circulation, pressurizer, steam generators and feedwater system, and computes pertinent variables including the-pressurizer pressure, pressurtrer water. level, and reactor coolant average temperature. Major assumptions are:
~~//02#4 b '~~
~~1. The plant is initially operatin(4t[the engineered safeguards design rating.~~
2. Initial reactor coolant average temperature is.kfF above the M nominal value, and the initial pressurtzer pressure is 30 psi above its nominal value. ,
~~'3. No credit is taken for the F nuik., ;,;eer- ;;r:Md r:H a* 99/b:~~
e pressurizer spray. ( . 1 , 15.4-29 0117F/COC4 4 1
~~,,.-a , , , , ,-.--,,,m..,.... ,, ,,,e.-w, , - - - . .~~
~~The severity of the feedwater-line break transient depends on a number of .~~
, system parameters including break size, initial reactor power, and credit taken for the functioning of various control and safety systems. The.most limiting feedwater line ruptures are the double ended rupture of the largest feedwater line, occurring at full power with and without loss of offsite power, with no credit taken for pressurizer control.
t e T 4 4 e, t I 1
l SON-1 l
~~4. No credit 1$ taken for the high pressurizer pressure reactor trip. -'~~
Note: This assumption is made for_ calculational convenience. Pressurizer power-operated relief valves and spray could act to delay the high pressure trip. Assumptions 3 and 4 permit evaluation-of one hypothetical, limiting case rather than two possible cases: one with a high pressure trip and no pressure controls; and one with pressure control but no high pressure trip.
~~5. Main feed to all steam generators is assumed to stop at the time the - >~~
~~break occurs.~~
~~. 4~~
~~6. Saturated liguld discharge only (no steam)-is assumed from the. ;~~
affected steam generator through the'feedline rupture. This ;l assumption minimizes energy removal from the NSSS during blowdown. ' W "o cred't it t9 ea 'e- the 1e - ta- wete- teuel tris & the t'#ected ;
~~--st":e ; ner:ter er.tt! ::ent! !!y e!' " utd !: d!!: hor;-d #r^= the 1 Stei ;enerater th:11. .~~
~~g 4 ,. d 'd R J dyr p .'u C j 7f. The worst possible break area is assumed; 1.e., one thatV: ;ti:; t' "'N'H 4~~
-- qo -S" - ....... ted 5t:'e m.... ;i.m.i n:rit:r h..ed ......,.&_..A
~~:::t;r~~
~~. u. trt; ...u_ ;- 1; !;^ ._ < ;te.e;.~~
~~r;r.. m. ,' z'n '..'. . 1. . -'. __ . in.., n s;~~
. . . _ . . _ _ 7"..'.:__.U . . . . :_:1 . . .. 'H . . . ::.~ _**_ .... . ., G',.3 #er 'a" ':;;l w.m. a vviu .. J. ;te;;/'::d '!r c'::: :5. This ption etntmizes the;stete ;----=ter '!r!d i="--tery at th" t!--
~~Ju f m49 , uy a r. ..Ip, : d tn:r:iy :::!:!::: th: :::"!t::t 50:tep O' th' - e:ter~~
~~.,/,,tg R g . $^^!a"* 'Fer Ot!:"lettent! :! ;!!:!ty, th- 't-"if d!::har; 're -~~
4'r. '"A <* M'*he t'fect^d St*=e ;enerater three;5 the re=t"re !! itte--d re-stant k f p,,,cd et Aed- the3200 water !he!: 16 velc.) in The reactor trip the unaffected the narrow range instrument spane! .:, , 0.0" f;;; i:S-mormal -wter - steamwas assumed generator to be actuated decreases to,,Pt' whenf'off,, I 9,. 1 . FJ. No credit is taken for heat energy deposited in reactor coolant system metal during the Reactor Coolant System heatup. 6(* 9N. No credit is taken for. charging,4 letdown, Or ;;f;t; ' ',- ti 7 l
~~/o,W. Loss of offsite electrical power is assumed after the reactor trip, !~~
~~and reactor coolant flow decreases to natural circulation. 'j~~
~~//R. Steam generator heat transfer area is assumed to decrease as.the ,~~
shellside liguld inventory decreases. AM. Conservative core residual heat generation is assumed based upon long term operation at the* initial power level 0 M. The auxiliary feedwater Q A. is"-'+^d e;;. .dM by:% M a"tr i e* preceding the tri
@ gJ ergl.t" t- " W +h th: f;;d 7 t: ;' ". T JJ.. ^^m dditt:::!'!0 -'-"te: t: ....o.a A.<y. tw, s .g u ... ... og --a ..a +6. ,. i . . . .; m q ,,
T" ^0^ i s s u d. l i;.73 7;ed. iwi ridri a Liis ..ef7;;t;C ^ tee; ^;r.;r:terb l T%e cuyhuy feeda:<k syck,v is a.ssamed. -tv supply c. Ae4a.I "V I 4@ 9P<n +o % uncMec.+ed stecun SMus @#d" M*i M A\\ces : - 15.4-30'
~~, AO "#~~
OllN/COC4 W j (
l
~~-1 i .~~
~~3) The tur,bine driven pur.p is assumed to fail.~~
i
2) The motor-driven pump- supplying the faulted steam generator is assumed - I to conservatively spill all its flow out the break. The intact steam generator aligned to that pump is therefore assumed to' receive no flow.

3) The remaining motor driven pump r,upplies flow to two intact steam '
generators. . i A 60 second delay was assumed following the low-low level signal to allow
~~time for startup of the emergency diesel generators and the auxiliary _~~
~~L. feedwater pumps. O i l 0~ e av e l l 2 f w i~~
~~, , , n .~~
~~t , J q J~~
~~-1 '~~
~~SQN~~
~~-l 1~~
~~13. T y.7.}. y iiquic re g i g.,,..j jipi. qw...y egy, .- )~~
.....,.'......"'..,_,;'<'-'"""p' . 7. .... , , , n ;",I' " W " , ; r rg ~ -
MARVEL code is used to caleviate'the course of the system trans t ont the time that bulk boiling begins in the RCS. After this . It-is as d that all excess heat addition (core residual heat g eratten 1 l less aux lary feedwater heat removal capacityF boils satura d liquid.in .l the RCS an that'further, the steam thus generated forces turated
~~I Ilquid out o the pressurizar. This assumption maxistre he calculated ;~~
loss of reacto coolant. gggg (A J l T / / f $ U C Q 'pt Figure 15.4.2-8 shows he calculated plant parame ers following a l o feedline rupture. The sumed auxiliary feedwatt'r flow rate is capable of removing decay heat I seconds after trip. After this time, core decay heat decreases below he auxillary ftpdwater heat removal-capacity,
~~- Md[ 3 , and reactor coolant temperath(es and pres res decrease. Calculated results for these are listed bilow.~~
t Calculated reactor coolant discharge j 24,290 lbs - through pressurizer safety valves , prior to bulk boiling. /( Calculated reactor coolant discharge- \-
/\ 92.000 lbs through pressurizer safety valves between <
time of bulk boiling and time pf auxillary 's feedwater heat removal capactty equals N decay heat j/ N
/
Fluid remaining in reactor' coolant- 409,000 lbs ' system , T-Maximum required relief' rate from
~~pressurtzer safety valves:~~
a. Steam relief, prior to 40 ft*/sec filling pressurtzer

b. Waterrelief/priortobulk 1. ft*/sec l bolling f

c. Water relief, after bulk-bolling 13 ft see-These results show, that even with the conservative method of !
calculation, the remaining.. liquid volume is more than adequate to 11 4 the RCS to the top of the reactor. core. Therefore, the reactor core would remain covered. Decay heat would be removed by steam,-- generate'd-. in the core by decay heat, condensing in the steam generator and flowing \ . back into the core. Y 15.4 31 1 0117F/COC4
~~.--i. -. -4,. -.-.--,.-.,-4_,,,,r, --,....ry,_o.,.n -+,%.,m.,-. , - , , ..w, --.- ,~~
~~I 1~~
~~'l 3~~
RESULTS Figure 15.4.2-8.a (sheets 1 and 2) shows the calculated plant parameters following a feedline rupture for.the case with offsite power. Figure e 15.4.2 8.b (sheets 1 and 2) shows the calculated plant parameters ~ following a feedline rupture with loss of offsite power. The calculated - ; sequences of events for both cases analyzed are presented in Table-15.4.1-12(sheet 4).- . The systes response'following the feedwater line rupture is similar for both cases analyzed.. Results figures show that pressures in the RCS and ' main steam system remain below 110 percent of the respective design j pressures. Pressurizer. pressure increases until reactor trip occurs on ; low low steam generator narrow range level. Pressure then decreases, due to the loss of heat. input, until the safety injection system is actuated. on low steamline pressure in the ruptured loop. Coolant expansion occurs. I due to reduced heat transfer capability in.the-steam generators. The pressurizer safety valves open to maintain primary pressure at an ( acceptable value. The calculated relief rates are weil within the relief . capacity of the pressurizer safety valves.. Addition of the safety . t. -! injection flow aids in cooling down the primary and helps to ensure that sufficient fluid exists to keep the. core covered with water. 3 f The reactor core remains covered with water throughout the-transient as water relief due to thermal expansion is limited by the heat removal .' capability of the Auxiliary Feedwater System and makeup is provided by the safety injection system. Bulk boiling does not occur in the RCS at any time in the transient. .. t i a h i 4 l l
~~.I i~~
*.. \
y
~~,, \~~
~~.SQN 14L9n. the cticulated required relief rates ara ' ell witnin the relief capacity of the pressor :2 . ;;f.., ==ive .~~
Therefore, no overpressur-1 Iration oGettrT within the RCS. ! m The time sequence of events'ist showns on Table 15.4,1-12. l 15.4.2.2.3 Conclusion Results of the analysis show that for the postulated feedline rupture, the assumed Auxiliary Feedwater System capacity is adequate to remove decay heat, to prevent overpressurizing the RCS and'to prevent uncovering the reactor core. ; i p 15.4.3 Steam Generator Tube Ruoture , i 15.4.3.1 Identification of Causes and Accident Description . l-L ! The accident examined'is the complete severance of a single steam p generator tube. The accident is assumed to take place at power with the reactor coolant contaminated with; fission products corresponding to-continuous operation with a limited amount of ~ defective fuel rods. The I . accident leads to an increase in contamination of the secondary system due to leakage of radioactive coolant from the RCS, In the event of a coincident loss of offsite power, or failure of the condenser dump system, discharge of activity to the atmosphere takes place via the steam generator safety and/or power operated relief valves.
~
In view of the fact that the steam generator tube material is Inconel 600 and is highly ductile material, it is considered that the assumption of a C, complete severance is somewhat conservative. The more probable mode of tube failure would be one or more minor leaks of undetermined origin. Activity in the steam and power conversion system'is subject- to continual , surveillance and an accumulation of minor leaks which exceed the limits established Unit operation in the Technical Specifications is not permitted during-the: The operator is expected to determine that a steam generator tube rupture has occurred, and to identify and isolate' the faulty-steam generator on a restricted time scale in order to minimize contamination of the secondary
l. system from theand ensure faulty termination of radioactive release to the~ atmosphere unit. The recovery procedure can be carried.out on a time scale which ensures that break flow to the secondary system ls terminated before water' level in the affected steam-generator rises into the main steam pipe.
Sufficient indications and controls are provided to enable the operator to carry out these functions satisfactorily, i ! Consideration of the indications provided at the control board, together. i with the magnitude of the break flow, leads to the conclusion that the , isolation procedure can be completed within 30 minutes of accident L initiation. , 4 15.4-32 Oll7F/COC4 i
~~- - _ _ _ _ _ - ~ - - - . - - . - _ _ - _ _ - _ . . _ - - _ _ _. _~~
~~i a SQN-6 r~~
14. Commonwenith-Edison Company, Zion Station Final Safety Analysis Report. Amendment 20 Appendix 14E. U. S. Atomic Energy Commission Docket Numbers 50-295 and 50-304, May. 1972. i

15. Olttus, F. W. 'and L. M. K. Soelter, University of California !
~~(Berkely), Pubis, Eng., 2, 433 (1930).~~
~~15. Jens W. H., and P. A. Lottes, " Analysis of Heat Transfer, Burnout.~~
Pressure Drop, and Density Data for High Pressure Mater," USAEC i Report ANL-4627 (1951). -
~~17. Macbeth, R. V., " Burnout Analysis, Pt. 2. The Basis Burn-out Curve,"~~
U. K. Report AEEW-R 167, Winfrith (1963). Also Pt. 3. "The { Low-Velocity Burnout Regimes," AEEW-R 222 (1963) Pt. 4 " App 1tcation .
~~. of Local Conditions Nypothesis to World Data for Uniformly Heated '~~
~~Round Tubes and. Rectangular Channels," AEEW-R 267 (1963).-~~
. 18. Douga11. R. S., and W M. Rohsensow,- Film Bolling on the Inside of 4 Vertical Tubes with Upward Flow of Fluid at Low Quantitles, 3 MIT Report 9079-26. '
19. D. M. McE11got, L. W. Ormand and H. C. Perkins, Jr., " Internal -Low Reynolds - Number Turbulent and Transitional Gas Flow with Heat Transfer " Journal of Heat Transfer, 88, 239-245 (May 1966). }
~~, 20. W. H. McAdam, Heat Transmission, McGraw-Hill 3rd Edition,1954, i~~
~~p. 172. .~~
-?
~~I j l 21. J. M. Geets, " MARVEL - A D191t141 Computer Code for Transient l Analysts of a Multiloop PHR System," WCAP-7909, June 1972.~~
22. F. S. Moody, Transactions of the ASME, Journal of Heat Transfer, i February 1965,-Figure 3, page 134.

23. F. M. Bordelon, " Calculation of Flow Coastdown After Loss of Reactor
~~! Coolant Pump (PHOENIX Code) " WCAP 7969 September 1972. l6 ft.ck~~
~~24. T. W. T. Burnett, C. J. Ah.ty ., J. t "dr,--A< -P r-Aose, "LQFTRAN Coce Descr1ption,"~~
~~P-7907 ,:.... ; ^ (#cn-8 yi e+vy , Apn i MF4 ., . A (l'. cPr w+Ary), ycAP-7(m]-g~~
~~25. C. Hunin, "FACTRAN, A Fortran IV Code for Thermal Transients in a +~~
~~UD, Fuel Rod," WCAP-7908, June 1972.~~
~~26. T. N.-T. Burnett, " Reactor Protection System Diversity in _~~
~~q Westinghouse Pressurized Water Reactor," NCAP-7306, April, 1969..~~
27. T. G. Taxe11us, ed. " Annual Report - Spert Project October 1968 September 1969", Idaho Nuclear Corporation IN-1370, June, 1970.

28. R. C. Lilmatainen, and F. J. Testa, " Studies in TREAT of .
~~21rcaloy-2-Clad, UO -Core Simulated Fuel Elements," ANL-7225 January - June 1966, p.177, November 1966. 4~~
~~, 15.4-50 0117F/C0(.4 1~~
~~i ( t 7s 5-u~~
- "_ a 1 I I I I I .-
~~, p 660 - lH/ /~~
~~$W =w 0~~
y - w t / 6 l $80 . i ! ?
~~. /~~
~~1900 '~~
~~' ,b e~~
~~I / g 1700 -~~
~~/. $m L' '~~
I $ 1500 - - 3-t:: - g 13o0 ,-,' - e E '
!!00 - ' / /
2000 w , g -i
~~.7 y _ 2600 -~~
l m2 \ -* er - g ' 2@ - - a w E I I I I I I 2200 N . 200 *500 600 800 1000 2000 l TIME'(SECONDS)
~~a Figur l~~
~~- 13.4,2 8 Main Feedline Rupture Accident - Cee Avwage Tempw ve Pressurlaw Pressure and Wotw Volume as a Function of . Time R e g lc,c. e d E F g d e 15.4. 2 -?. R , Shee t I ad 2 i~~
~~IS .4. 2 - ?. b , su e t I ~ d 2~~
~~-_. - - -- E - - - -~~
4 i l t i 4 4600. 1446. ' 5 l l C ssee. ; 1 l l400. d i l ' i* 1400. . b* 1998.
~~$ l 8 .... ;~~
2 . E ..... 8 w g 00. . I too. ' O. b [ lt le3 80I 103 It' ..j flPE IltCI e t i f 4640. t t.49. l- g ( G tiet. l !- A t W 2000. ut
~
(4
~~$ eseo.~~
t 8 1690. 3 ..... l 0 l f i 4 1200. , 9'
~~-g.~~
~~I960. #~~
~~#e eto.~~
i,. i.i ..# i.: i.. l tini .src. < FIGURE 15.4.2-8.a (sheet 1) MAJOR RUPTURE OF A MAIN FEEDWATER PIPE.-WITH . OFFSITE POWER - Pressurizer Pressure and Water Volume as a Function of Time (
~~-e- +~~
~~} )~~
0 j i . l i 960. j TSAT . 654. ' [ [-
~~\ '~~
~~TH0T~~
~~- t g .00. .~~
{ lle. TCOLD - k.a - 600 < *
~~.50. .~~
I
~~.00.~~
~~It' 198 lli 198 30 8 tiet ettts~~
~~. ? ,h 966.~~
~~I ** Tggy E TH0T~~
~~\ '~~
~~600. Ilit. TCOLD e\~~
~~\ / !~~
~~1 k lit . ' w j E .44 7 . ; l~~
~~41. ,~~
~~n'-~~
~~j. il i ... ite 104 ,,.~~
Tiet entts FIGURE 15.4.28.4(sheet 2) MAJOR RUPTURE OF A KAIN FEEDWATER P! pet $1TH OFFSITE POWER Loop Temperature as a Function of Time l 1 ;
~~= i s ,~~
~~4....-. - allt. < E . [ 1640. S e t t. ' I~~
~~> 3400.~~
~~b 3000. ' + 1 g....<~~
~~.m 4 g .. . . . .~~
~~m g..... .~~
~~e. _~~
~~= 348 30 8 300 308 30 8 !~~
~~tint altte 4066. ,. g t a t t.~~
~~Seet. - 1 g~~
G~ tttt. Xc : tw N g g #t88. o W u . lat e. < Ist1. ,,. IB8 16 8 80I 168 10 8 flat sitta 1 FIGURE 15.4.2 8.b (sheet 1) MOOR RUPTURE OF A MAIN FEEDWATER P!PE,* WITHOUT OFFSITE POWER - Pressurizer Pressure and Water Voleme as a Function of Time : 1
~~$ l~~
~~i i i~~
~~,ee. !~~
l ..... l 4 I C ...., SAT x j
~~..e..~~
~~g NOT~~
~~.,e.~~
~~g _ W ..... t ... . . E s.. . . TCOLD~~
~
~~Y..... m see.' i~~
~~. l~~
~~see.~~
~~,ee ies i.e see it' fint 48t88 l~~
~~eee. . -~~
~~. . . . , SAT x ; ..e..~~
~~THOT ' see. i.e. TCOLD s.... i.e.. see.~~
,ee i: ie se8 is' tint Ilttt FIGURE 15.4.28.b(sheett) MJOR RUPTURE OF A MlH FEEDWATER PIPE, WITHOUT OFF5ITE POWER - Loop Temperature as a Function of Time e
Table 15.4. I-12. (shwt4) Major Rupture of a Main feedline rupture occurs 10.0 Mein Feedwater l Pipe Low low steam generator level reactor 15.0 With Offsite Power available Rods begin to drop 17.0 1 Auxiliary feedwater started 75.0 - l
~~$!$ Low Pressurizer Pressure 140. 0 Low steamline pressure setpoint reached 192.5 Peak relief rates from pressurizer 767.0 safeties,(1.267lb/ft3)~~
~~Core decay heat plus pump heat decreases < 8000 l~~
~~. to auxiliary feedwater heat removal '~~
capacity Major Rupture of a Main feedline rupture occurs 10.0 , Main Feedwater Pipe Low low steam generator level reactor 15.0 trip Without Offsite Power Available Rods begin to drop" 17.0 Auxiliary feedwater started 75.0
~~' Low steamline pressure setpoint reached 100.3 Peak relief rates from pressurizer 363.0 safeties (3.677 lb/ft3) ;~~
l Core decay heat plus pump heat decreases < 3000 to auxiliary feedwater heat removal capacity i 9 1 at' e d i m
i ENCLOSURE 5 SEQUOYAH NUCLEAR PLANT DRAFT FSAR CHAPTER 7 MARKUPS . FOR EAGLE 21 RPS UPGRADE i s i a t t i i k i l r e F
*4
~~I SQN b$* l ef /u j (~~
~~7.1 INTRODUCTION~~
CHAPTER 7.0 INSTRUMENTATION AND CDNTROLS 1 l This chapter presents the various plant Instrumentation and Control l ! Systems by relating the functional performance requirements, design ' bases, system descriptions, design evaluations, and tests and inspections for each. The information provided in this chapter emphasized those
~~instruments and associated equipment which constitute the protection i~~
system as defined in IEEE Std. 279-1971 "!EEE Standard: Criteria for . Protection Systems for Nuclear Power Generating Stations". ; The primary purpose of the Instrumentation and Control Systems is to provide automatic protection against unsafe and improper reactor i operation during steady state and transient power operations (Conditions l ! !!, !!!) and to provide initiating signals to mitigate the consequences of faulted conditions (Condition IV). For a discussion of the four conditions see Chapter 15. The information presented in this chapter emphasizes those Instrumentation and Control Systems which are central to ; assuring that the reactor can be operated to produce power in a manner that insures no undue risk to the health and safety of the public. It is shown that the applicable criteria and codes, such as the Nuclear Regulatory Commission's General Design Criteria and IEEE Standards, concerned with the safe generation of nuclear power are met by these systems. Definitions
~~. The definitions below establish the meaning of words in the context of ;~~
their use in Chapter 7. . s Channel - An arrangement of components and modules as required to generate a single protective action signal when required b a plant condition. A channel loses its identity where single action signals are combined. DNBR - (Departure From Nucleate Boiling Ratto) - The ratio of the critical heat flux (defined as the transttlon from nucleate bolling to , film boiling) to the actual local heat flux. l Module - Any assembly of interconnected components which constitutes an toentifiable device, instrument, or piece of equipment. A module can be
disconnected, removed as a unit, and replaced with a spare. It has definable performance characteristics which permit it to be tested as a unit. A module could be a card or other subassembly of a larger device, provided it meets the requirements of this definition.
Components - Items from which the system is assembled (e.g., resistors, capacitors, wires, connectors, transistors, tubes, switches, springs, , etc.). o 7.1-1 0067F/COC4 s - 6 -y -- ,
i I 2. SON : i 1 ; 5tnote Falture - Any single event which results in a loss of function of a component or components of a system. Multiple failures resulting from
~~a single event will be treated as a single failure.~~
Protective t.ction - A protective action can be at the channel or the ' system level. A protective action at the channel level is the initiation of a signal by a single channel when the variable sensed onceeds a listt. . 1 A protective action at the system level is the intttation of the operation of a sufflctent number of actuators to effect a protective function. t Pronscutve Function - A protective function is the sensing of one or more i var' ab es associated with a particular generating station condition signal processing and the initiation and completion of the protective action at values of the variable established in the design basis. Twee Tests - Tests made on one or more units to verify adequacy of design. Decree of Redundancy - The difference between the number of channels monitoring a var' able and the number of channels which when tripped, will . cause an automatic system trip. - 4 Minimus Decree of Rodundancy - The degree of redundancy below which . operation is prohlb ted, or otherwise restricted by the Technical Spectftcations. Reeroducibility - This definition is taken from SANA Standard . PMC-202-1970. Process Neasurement and Control Terminology: " the closeness of agreement among repeated measurements of the output for the > same value of input, under normal operatin time, approaching from both directions.*.!g conditions t includes over a period of drift due to environmental effects, hysterests, long- term drift, and repeatability. l 1.ongterm drift (aging of components, etc.) is not an important factor in I accuracy requirements since, in general, the drift.15 not significant with respect to the time elapsed between testing. Therefore, long-term drif t may be eliminated from this definition. Reproducibility, in most
~~. cases, is a part of the definition of accuracy (see below).~~
Accuracy - This defintt1on is derived from SANA Standard PMC-202-Ig70, Process Neasurement and Control Terminology. An accuracy statement for a i device falls under Note 2 of the deftaltion of accuracy, which means l reference accuracy or the accuracy of that device at reference operating-conditions: " Reference accuracy includes conformity, hysterests and repeatability." To adequately define the accuracy of a system. the term reproducibility is useful as it covers normal operating conditions. The . I following terms.
~~trip accuracy" and " Indicated accuracy" etc., w111 then ,~~
include conformity and reproducibility under normal operating conditions. . Where the final result does not have to conform to an actual process variable but is related to another value estab11thed by testing, conformity may be eitnineted, and the term reproducibility may be
~~- substituted for accuracy. ,~~
~~7.1-2 0067F/C0C4 T~~
~~, m. . - - - , - - _ -,,--,,---,,,_,,,__,,,-,,.--m.- -,mm--~~
~~$0#-3 l~~
l
~~Readout tD y.hti - for consistency the final device of a complete channe)~~
, k. is consideres a readout device. This includes Indicators, recorders, j isolators (nonadjustable), and controllers. Channe) Atturacy - This definttlon includes accuracy of primary element, transaltter and rack modules. It does not include rendout devices or ' rack environmental effects, but does include process and environmental ( effects on field sounted hardware. Rack enviromental effects are
~
Included in the next two definttions to avoid dupilcation due to dual inputs. Indicated and/or Recorded Accuraev - This definition includes channel i i accuracy, accuracy of readout dev'ces and rack environmental effects. l I Trto Accuracy - This definition includes comparator accuracy. channel l accuracy, for each input, and rack environmental effects. This is the t tolerance expressed in process terms (or percent of span) within which the complete channel must perform its intended trip function. This , I includes all Instrument errors but no process effects such as streaming. The term " actuation accuracy" any be used where the word ' trip" sight cause confusion (for example, when starting pumps and other equipment). 3 Actuation Accuraev - Synonymous with trip accuracy, but used where the word " trip" may cause ambiguity. I i Cold Shuudown - The reactor is in the cold shutdown condition when the reactor is subtritical by at least 1 percent Ak/k and T.., is 1200'F with T.. defined as the average temperature across a reactc,r vessel as ( esatured,by the hot and cold leg temperature detectors. O Not Shutdown - The reacfor is in the het shutdown condition when the reactor is subtritical by an amount greater than or equal to the margin as specified in the 50N Technical Specifications and T., is greater than 200'T but less than or equal to T.,., where T..., is defined as any temperature at which the reactor is critical, llatted by the SQN Technical Specifications. Phase A Containment Isolation - Closure of all non-essential process lines which penetrate containment initiated by the safety injection signal. Phnso B Containmenu !solatiot - Closure of reestning process Itnes, In't ated by conta neent higa hlgh pressure. ' j T.1.1 Identification of safety Related tvstems 7.1.1.1 Safety Related Systeet The instrumentation required to function to achieve the system responses ( assumed in the safety evaluations, and those needed to shut down the plant safely are given in this section. , 7.1-3 0067F/COC4
)
4
%s
_ ,. ,_,,.---,..--,--ww+-
Y I SQN
)
~~7.1.1.1.1 Reactor Trte tvstem~~
~~l The Reactor Trip System is a function 411y defined systes described in l~~
' Section 7.2. The equipment which provides the trip functions is ,, ; identifled and discussed in Section 7.2. Design bases for the Reactor > i Trip System are given in Paragraph 7.1.2.1. i 7.1.1.1.2 Enoineered Safety Features Actuation System I The Engineered Safety Features Actuation System is a functionally defined j systes described in Section 7.3. The soutpoent %1ch provides the ,
~~; actuation functions is identifled and discussed in Section 7.3. Design bases for the Engineered Safety Features Actuation System are given in _~~
Paragraph 7.1.2.1. ; 1 7.1.1.1.3 Vital Power Sunniv systes , Design bases for the Vital Power Supply System are given in Paragraph 7.1.2.1. Further description of the system is provided in Section 7.6. 7.1.1.1.4 Austilary Control Air Syste's The Auxillary Control Air System supplies essentist control att to safety related items such as the availlary feedwater control valvest the Vacuum Relief System containment isolation valves; and dampers in the Austilary , : Guilding Gas Treatment System, the Emergency Gas Treatment System, and ' the Control lutiding NVAC system. Further description of the system Is given in Subsection g.3.1. . 7.1.1.2 tafety Related Disclav Instrumentation l Display instrumentation provides the operator with information to enable him to monitor the results of Engineered Safety Features actions following a Condition !!! or IV event. Section 7.5 and Tables 7.5.1-1 i and 7.5.1-2 provide information required to maintain the plant in a hot I shutdown condttion, or to proceed to cold shutdown. : 7.1.1.3 Instrumentation and Control tvstem Detteners l All systems discussed in Chapter 7 have definitive functional requirements developed on the basis of the Westinghouse NS$$ destyn. The systees are , supplied by Westinghouse with the exception of the vutal power and aux 111ary control air systems that were designed and supplied by TVA. 7.1.1.4 Plant Comnarlton .; Systes functions for all systems discussed in Chapter 7 are stellar to those of D. C. Cook Nuclear Plant and Trojan Nuclear Plant. Detailed comparison is provided in Section 1.3. , I i
~~, -l 7.1-4 0067F/COC4 j I g~~
~~SON I~~
~~(* 7.I.2 Identification of Safety Criterla~~
~~t 4~~
Paragraph 7.1.2.1 gives design bases for the systems given in Paragraph 7.1.1.1. Design bases for non-safet Jettlons which describe the systems.y Conservative related systems are provided considerations for in the instrument errors are lacluded in the accident analyses presented in , , , Chapter 15. Functional requirements, developed on the basis of the , i ( results of the accident analyses, which have uttllred conservative ! assumptions and parameters are used in designing these s'ystees and a '
~~pre-operational testing program vertfles the adequacy of the design.~~
j Accuracles are given in Sections 7.2. 7.3 and 7.5. The documents Itsted below were considered in the design of the systems (t given in Subsection 7.1.1. In general, the scope of these documents is : given in the document itself. This determines the systems or parts of systems to which the document is appilcable. A discussion of. compliance t
with each document for systees in its scope is provided in the referenced i sections.
~~Secause some documents were issued after design and testing had been completed, the equipment documentation may not meet the format requirements of some standards. The documents. considered are:~~
1. " General Design Criteria for Nuclear Power Plants'. Appendix A to I Title 10 CFR Part 50. July 7, 1971. (See Sections 7.2. 7.3. 7.4. and i
~~7.7). l -~~
~~2. ' Regulatory Guide 1.11 (March,1971) -Instrument Lines Penetrating~~
( Primary Reactor Contatament' Reculatory Guides for Water Cooled : Nucloar Power Plants. 01 vision of Reactor Standards. Atomic Energy
Comm.ssion. (5ee Paragraph 6.2.4.1 and 6.2.4.3). *

3. " Regulatory Guide 1.22 (February 1972) -Periodic Testing of Protection System Actuation functions". Raoulatory Guides for Water j
Cooled Nucloar Power Plants. Olvtslon of Reactor Standards. Atomic j Energy Come ssion. (See Paragraph 7.1.2.8). ,' l 4. The Institute of Electrical and Electronte Engineers. Tr-i
~~'IEEE Standard: Criteria for Protection Systems for Nunear Power :~~
l , Generating Stations". IEEE Std. 279-1971. (See Sections 7.2. 7.3. ! ' j 7.6). , , I ( 5. The Institute of Electrical and Electronic Engineer ~s. Inc..
*!EEE Standard Criteria for Class IE Electric S Power Generating Stations". IEEE Std. 3081971.ystems for Nuclear (See Section 7.6). -
~~6. The Institute of Electrical and Electronic Engineers. Inc.,~~
~~a i 'IEEE Standard for Electrical Penetration Assemblies in containment 1 f Structures for Nuclear Fueled power Generating Stations'.~~
~~1 IEEE Std. 317-1971. (See Paragraph 7.1.2.4).~~
~
1 , i k 7.1 5 0067F/COC4 2
1 i ( 4 ! l 1 SQN
~~7. The Institute of Electrical and Electronic Engineers. Inc.,~~
*!EEE Trial-Use Standard: General Guide for Qualifying Class ! j Electric Equipment for Nuclear Power Generating Stations". . l IEEE Std. 323-1971. (See Paragraph 7.1.2.5).
~~8. The Institute of Electricat and ElettronIc Engineers. Inc.. ;~~
~~'IEEE Trial-Use Gu1de for Type Tosts of Continuous-Duty C14ss ! !~~
~~Notors Installed Inside the Containment of Nuclear Power Generating~~
~~i Stations". IEEE Std. 334-1971 (see Paragraph 7.1.2.10).~~
~~9. The Institute of Electrical and Electronic Engineers. Inc.. !~~
*!EEE 5tandard Installation. Inspection, and "esting Requirements for l Instrumentation and Elet:tric Equipment During the Construction of Nuclear Power Generating Stations". IEEE Std. 336-1971. (See i Paragraph 7.1.2.6) j l
~~10. The Institute of Electrical and Electronte Engineers. Inc.. !~~
*!EEE Trial-Use Criteria for the Per1od1c Testing of Nuclear Power l
~~Generating Station Protection Systems.", IEEE Std. 334-1971. (See Paragraph 7.1.2.7).~~
~~11. The Institute of Electrical and Electronic Engineers. Inc.. -~~
~~'IEEE Trial-Use Guide for Selsate Qualtf tcation of Class ! 3 Electric Equipment for Nuclear Power Genetating Stations".~~
~~IEEE Std. 344-1971. (See Paragraph 7.1.2.11).~~
~~12. The Institute of Elecitical and Electronic Engineers. Inc. I~~
~~'!EEE Trial-Use Guide for the Appilcation of the $1ngle-Failure Criterton to Nuclear Power Generating Station Protection Systems". IEEE Std. 379-1972. (See Paragraph 7.1.2.12).~~
~~13. "qegulatory Guide 1.53 (June,1973) -Appittation of the $1ngle-Failure Criterton to Nuclear Power Plant Protection Systems".~~
~~Reculatory Guities for Water Cooled Nuclent Power Plants." Olvision of gg j P.eactor Stancards. Atom c Energy Conent ss'on. (See Paragraph 7.1.2.12).~~
~~'7.1.2.1 De 1en lases The technical design bases for the protection systems'are provided by~~
~~Westinghouse equipment spectf tcations which consider the functional requirements for these systems and appitcable criteria such as IEEE~~
, 279-1971. IEEE 317-1971. IEEE 323-1971 and the NRC General Design Criteria, , i . t 7.1.2.1.1 Reactor Tris System '
~~< The Reactor Trip System acts to Ilmit the consequences of Condition !!~~
1 events (faults of moderate frequency such as loss of feedwater flow) by, at most, a shutdown of the reactor and turbine, with the plant capable of returning to operation after corrective action. The Reactor Trip System
~~, features impose a limiting boundary region to plant operation which ,~~
j 7.1-6 0067F/COC4
ll SQN l ! !' ensures that the reactor safety lletts analyzed in Chapter 15 are not ! ! esteeded during Condition !! events and that these events can be i 1 accommodated wLthout developing into more severe condttions. l ! The tiesign requirements for the Reactor Trip System are derived by ;
analyses of plant operating and fault conditions where automatic rapid

i control rod insertion is necessary in order to prevent or listt core or * '
~~t reactor coolant boundary damage. The design 11alts for this system are:~~
1. Ninlaum DNBR will not be less than 1.30 as a result of any anticipated transient or a61 function (condition !! faults). ,
~~t (l 2. Power density will not esteed the rated linear power density for~~
~~ondition II faults. See Chapter 4 for fuel design Ilmits.~~
) ,
~~3. The stress Itatt of the Reactor Coolant lyttee for the various conditions will be as spectfled in Chapter 5.~~
i 4. Release of radioactive material will not be sufflctent to interrupt ' ! or restrict public use of those areas beyond the esclusion distance or to exceed the guidelines of 10 CFR 20. ' Standards For Protection Age. inst Radiation", as a result of any Condition !!! fault. , t
5. For any Condition IV fault, release of radioactive material shall not result in an undue risk to pubtlc health and safety nor will it onceed the guideltnes of 10 CFR 100, " Reactor $tte Criterta". ,
a 7.1.2.1.2 Encineered Safety Features Actuation System i s The Engineered Safety Features Actuation System acts to Ilmit the consequences of Condition !!! events (infrequent faults such as primary
coolant spillage from a small rupture which exceeds normal charging I system makeup and requires actuation of the safety injection system).
The Engineered Safety Features Actuation System acts to mitigate Condition IV events (limiting faults, which include the potential for i l significant release of redloactive material). The design bases for the Engineered Safety Features Actuation System are derived from the design bases given in Chapter 6 for the Engineered Safety Features. Design bases requirements of IEEE 279-1971 are addressed in Paragraph 7.3.1.2. General design requirements are given below. ! t- 1. Automattc Actuation Renutrements The primary functional requirement of the Engineered Safety Features
Actuation System is to receive input signals (Information) from the various on-going processes within the reactor plant and containment and automatically provide, as output, timely and effective signals to
( actuate the various components and subsystems comprising the ' Engineered Safety Features System. These signals must assure that i the Engineered Safety Features System will meet its performance , objectives as outlined le Chapter 6. 1 Y i 7.1-7 0067F/COC4 i
-- - - - - - - - - - - - - - - - . _ _ - - . _ - - - - ---.-______-____---,s __..--,--a. .-<.w-e---u-, , . , , , --rv~--,vsc. ,r ,.w. . , , ,, - , w. -,m,-,w-p ,,,*m,m-,-w-,--=-..=v--
i ! . i l 1 SQN l [ The functional diagrams presented in Fiqure 7.2.1 1 Sheets 5. 6. 7 j and 8 provide a graphic outilne of the ogic associated with the ESF l actuation syst6e. -
)
i 2. Manual actuation tenutrements j l The Engineered safety Features Actuation System has provisions for manual'y inttlating from the control room all of the functions of the Engineered Safety Features System. Manual actuation serves as backup to the automatic initiation and provides selective control of : Engineered Safety Features service features. . l. ' 7.1.2.1.3 Vital power tueely System . , The Vital Power Supply systes provides continuous, reliable, regulated l single phase AC power to all instrumentation and control equipment ! required for plant safety. Details of this system are provided in . Section 7.6. The design bases are given below:
1. The inverter shall have the capacity and regulation required for the AC output for proper operation of the equipment supplied.

2. Redundant loads shall be assigned to different distributton panels which are supplied from different inverters. ;

3. Aux 111ary devices that are required to operate dependent equipment ..
will be suppiled from the same distributton panel to prevent the loss ; of electric power in one protection set from causing the loss of equipment in another protection set. No single failure shall cause a loss of power supply to more than one distribution panel.
~~4. Each of the distributton panels will have access to an inverter and a ,~~
standby power supply. 7.1.2.1.4 Emeroency Power Design bases and system description for the emergency power supply are i provided in Chapter 8, 7.1.2.1.5 Interlocks Interlocks are discussed in Sections 7.2. 7.3. 7.6 and 7.7. The protection (P) interlocks are given on Tables 7.2.1-2 and 7.3.1 3. These interlocks are designed to meet the requirements of paragraph 4.12 of IEEE 279-1971. Control interlocks are identified on Table 7.7.1-1. Because control interlocks are not safety related, they have not been specifically designed to meet the requirements of IEEE Protection System Standards. < 7.1-8 0067F/COC4
]
q SQb6 7.1.2.1.6 g ggiggi 279-1971, ( , typasses are destened to meet the requirements of !![[A discussion of by paragraphs 4.11, 4.12. 4.13 and 4.14. is given in Sections 7.2 and 7.3. 4 7.1.2.1.7 toutement protection Equipment ; (! The triteria for equipment protection are given in Ch t i installed to protect it from damage. , accepted standards and criteria aused at providing reliable instr 1971 tion which is available under varying conditions. equipment is seismically quellfled in accordance with Ittt 344-l separation , (Reference 10). During construction, 279-1971, independence and iseith ' achieved, as required by ItttThis serves to protect against complete destruc separation. i system by fires, alsslies or other natural hazards. 7.1.2.1.8 pfversity Functional l Functional diversity has been designed into the system.-Logte f, l ( diversity is discussed in NCAP 7706, 'An tvaluation The entent of Solid State ) Reactor Protection in Anticipated Transients " Reference 1. of diverse system variables has been evaluated for ti 2. a wide varie postulated accidents as discussed in NCAP 7306, " Reactor Prote Systea Olversity in Westinghouse PressurizedllyWater Reactors l Generally, two or more diverse protection functions would automat terminate an accident before unacceptable consequences could occur For example, there are automatic reactor trips based t upon nucle measurements, reactor coolant loop temperature measurements, pr pressure and level measurements, and reactor coolant f pump and under voltage measurements as well as manually, and by init a safety injection signal. Regarding the Engineered Safety Features Actuation System fo l ( safety injection signal can be obtained manually or by automatic initiation from tro diverse parameter measurements. ^
~~1. Low pressurizer pressure, -~~
! High containment pressure. f 2. I II 55E.id12.535Id[3D' y yy, , 3 y yy y y 5t W5 5 Ww W ITD W E W FFD I I3D l 2. "t;h :t:= "= f t":mtta' ;-em e i' 0067F/COC4 a 7.1-9
~~,,,,,,.w-.,y,,,cw,,,.,,,,,,,,w,,,r.,-,..,,,-e.,.-,. - - , - . ,-~~
l lh 4 1 SQN-4 l - b5
~~,..-.h2!~~
~~bb ........... bh b !.b!b bb5...'......... ! bh . .. .... b N1 . - . . . . . .lii^~~
. . . N!I'iMki . . ,b!TN3.MifiU2 . . . . - . . . . . . . . . . . . NII . . . . . .IIC 7.1.2.1.g Setnoints l The Technical SpecificAttons for the Sequoyah Nuclear plant incorporate both nominal and limiting setpoints. Instrument spans are selected such i
that limiting setpoints are at least 5 percent from the end of the instrument span. Nominal settings of the setpoints are more conservative 4 ' i than the Itatting settings. This allows for cattbration uncertainty and instrument channel drift without violating the lletting setpoint. Automatic initiation of protective functions occur at the nominal setpoints. n
7.1.2.2 Indonendence of Redundant Safety Related lestems The safety related systems in Paragraph 7.1.1.1 are designed to meet the independence and separation requirements of criterton 22 of the 1g71 -
i General Design Criteria and Paragraph 4.6 of IEEE 27g.1g71. The ! administrative responsibility and control provided during the design and installation is discussed in Chapter 17 and TVA Topical Report TR76-1 which address the Quality Assurance programs applied by Westinghouse and TVA. The electrical power supply, instrumentation, and control conductors for redundant circMts of a nuclear plant have physical separation to preserve the etwndancy and to ensure that no single credible event will prevent operation of the associated function due to electrical conductor damage. Detailed information pertaining to electrical cable for safety related systems is given in Paragraph 8.3.1.4. Critical circuits and functions include power, control and analog instrumentation associated with the operation of the Reactor Trip System or Engineered Safety Features Actuation System. Credible events shall include, but not be limited to, the effects of short circuti;s, pipe rupture, missiles, etc. and are considered in the basic plant design. Control, board detalls are given in Paragraph 7.7.1.10. b 7.1. 2. 2.1 General
~~1. Cables of redundant ct' r cuits will be run in separate cable trays, conduits, ducts, penetrations, etc. .~~
l-
2. Circuits for non-redundant functions should be run in cable trays or conduit separated from those used for redundant circuits. Where this can not be accomplished, non-redundant circuits may be run in a cable I tray, condu'2, etc. assigned to a redundant function. When,so routed, it must remain with that particular redundant circuit routing and will not cross-over to other redundant groups.
l 7.1-10 0067F/COC4
i i 4
\
~~! (.' i I~~
~~3. Mortrontal and vertical separation will be maintained between cable trays, associated utth redundant circuits.~~
j
~~4. Where it is tupractical for reasons of equipment arrangement to provide separate cable trays, cables of redundant circuits any be~~
~~' Isolated by physical barriers or be installed in separate metallic .~~
~~condult. l 5. Power and control conductors rated at 600 volts or below shovid not be placed in cable trays wtth conductors rated above 600 volts.~~
)
6. Analog or other low levs1 type signal conductors will not be routed I in cable trays containing power or control cables.
, 7. 1.2.2.2 Snecific tvstems !
Channel independence is carried throughout the systes, estending from the sensor through to the devices actuating the protective function. Physical , separation is used to achieve separation of redundant transmitters. ~ Separation of wiring is achieved using separate wireways, cable trays, conduit runs and inment penetrat'ons for each redundant channel 1 set. Redundtnt equipment is separated l>y locating modules in I j different prot tion rack sets. Each redundant channel set is energized 1 from a separate AC power feed. , pre 4tetM Separation of There ara g u g parate proces p P rack sets. on w is malatained I redundant ....._, thannels begins at the process " ettion racks > in the flitid wiring, containment penetrations an dant, g :; channels _. i c racks. Red ___.
. to arethe separated redundant by locating trains inJQ: n different rack sets. Since all equipment within any rack,is associated with a single protection channel i r
' set, there is no requirement for separation of wir' ng and components withta the rack. l l l Independence of the logic trains is discussed in Reference 11. Two reactor trip breakers are actuated by two separate logic matrices which interrupt power to the control rod drive mechanisms. The breaker main . contacts are connected in series with the power supply so that opening l either breaker interrupts power to all full length control rod drive mechanisms, perettting the rods to fr6e fall into the core. I -
1. Reactor Trip System .

a. Separate routing is g g ined for the four basic Reactor Trip System channel sets.,_ _ _. sensing signals, bistable output signals and power suppites for such systems. The. separation of g
~~these four channet sets is maintained from sensors to instrument racks to logic system cabinets. T 9 7.1-11 0067F/COC4~~
*e ,c,-,- * , . - - . , - , . , , .
~~i II lt- l Y e l~~
b. Separate routing of the reactor trip signals from the redundant ;
~~logic system cabinets i separated from the .- _four,g~~
~~., channel g ained, sets. and in addition, they are l~~
2. EngineeredSafetyFeaturesActuationSystem ;
e '
a. Separate routing ; g tained for the four beste sets of ESF '
l i_,. sensing signals, bistable output signals ' Actuation System"es for such systems. The separation of these and power supp11 ' r i four channel sets Is maintained from sensors to instrument racks ) to logic system cabinets. i i
b. Separate routing of the ESF actuation signals from the redundant '
logli i fou(_ _gem -; channel cabinets sets. is matntained and ts separated from the l l . ! c. Separate routing of control and power circuits assottated with . the operation of engineered safety features equipment is required to retain redundanc'es provided in the system design and power , suppites,
3. Vital power Supply System ,
j The separat1on eriteria presented also apply to the power suppites for the load centers and busses distributing power to redundant ; components and to the control of these power supplies. p g :ter Trip System and Engineered Safety Features Actuation Sys the same power supply and channel set identity (I, !!, III or IV). , 7.1.2.2.3 Fire Protection Dotatis of fire protection are provided in Subsection g.5.1. 7.1.2.3 _ Physical identification of Safety Related tauinment , Adequate identification is provided to distinguish Reactor Trip, Engineered Safety Features and Instrumentation and Control Power Supply ' Systems as safety related. As previously stated there are four protection channel set racks. A color coded nameplate on each rack of each set is used to identify the protection sets. The color coding of the protection set nameplates ts: - -- Protection Set Color Coding , 1 Red with white lettering
!! Alack with White lettering 111 Blue with white lettertny ,
IV Yellcw with black letter.ng : l 7.1-12 0067F/COC4 ri,- - y- - e.,-,- , - . , , , , - 3 ,* -,y,--. i,,,,, , , , . . . - . . + . , , , , . . ------%,., ,.ir-.--e,_ s-+-e,.,i-,,.-a. - - - .. ,,,,s~.-e.--ww-ee-tw-_-ew "- mee v-e-a-w **
I
i ! t-
' H 50"-' l l (
l All non-rack mounted protective equipment and components are provided j i vtth an identifIcat1on tan or nameplate. Small electrical components All ; such as relays have namep' ates on the enclosure whtch house them. l i
~~cables are numered with identification tags. In congested areas. such f~~
l as under or over the control boards, instrument racks etc., cable trays and conduits containing redundant circuits shall be identified using permanent markings. The purpose of such markings, discussed in detall in Chapter 8. Paragraph 8.3.1.4. 18 to fact 11 tate cable routing identification of future' modification or additions, e i Positive permanent identification of cables and/or conductors shall be made at all terminal points. There are also identtf tcation nameplates on j ! the input panels of the solid state logic protection system. i l 7.1.2.4 Conformance to Itt! 317-1971 (Reference 3) f Electrical penetrations and conformance with IEEE 317-1971, ' Electrical ,
~~Penetration Assemblies in Containment Structures for Nuclear Fueled Power '~~
Generating $tations" are discussed in Chapter 8. Subparagraph 8.3.1.2.3. : 7.1.2.5 Conformance to Itti 323-1971 (Reference 4) - I . Reactor Trip System equipment is type tested to substantiate the adequacy of design. This is the preferred method as indicated in IEEE 323. Type l tests may not conform to the format guideltnel set forth in Section 5.2 of IEEE 323. since type tests on some equipment were performed prior to . ( issuance of the standard. However, it has been determined by Westinghouse that the testing and documentation was comparable to that required by 3 ! IEEE 323. ! 7.1.2.6 Conformance to Ittt 336-1971 (Reference 5) A discussion of conformance to IEEE 33614 given in Paragraph 8.3.1.2.2. 6 j j 7.1.2.7 Conformance to Itti 138-1971 (Reference 6) L 1. The reliability goals specified in Paragraph 4.2 of Reference 6 are , I being develened, and adequacy of test frequencies will be j demonstratec. 1
~~2. The periodic test frequency discussed in Paragraph.4.3 of Reference 6 l~~
~~and specified in the plant Technical Specifications, is~~
~~conservatively selected to assure that equipment associated with i~~
~~protection functions has not drif ted beyond its minimum performance requirements. If any protection channel appears to be marginal or~~
- requires more frequent adjustments due to plant condition changes. ' ' I the test frequency is accelerated to accommodate the situation untt) l the marginal performance is resolved.
i ; i l ; 4 l 3 7.1-13 0067F/COC4 , i F
~~,,...m ..._w,_ ,,_#__.y. . , . - _ . - . .. ..r,,. , , , , . - . _ , .-,._._,,....% ..... _. _ m._ ,,_,.__.,_y.yo..._.._m.- . . , . . . . . . , - .~~
)
~~l IT I~~
~~) $0N l~~
~~3. The test interval discussed in paragraph 5.2. Reference 6. Is :~~
developed primarily on past operating espertence and modified if l necessary to assure that system and subsystem protection is reliably
i provided. Analytic methods for deteretning reliability are not used i to determine test interval. :
7.1.2.8 Conformance yo Raoulatory Guide 1.22 (February. 1972) ! (Reference 7? l periodic testing of the Reactor Trip and Engineered Safety Features Actuotton Systems, as described in Subsections 7.2.2 and 7,3.2. compiles l with NRC Regulatory Guide 1.22. ' Periodic Testing of protection System Actuation Functions.' Under the present design, there are functions which are not tested at power, because to do so would render the plant in a less-safe condition. These are as follows:
1. Generation of a reactor trip by tripping the turbine; [

2. Generation of a reattor trip by sse of the manual trip switch;

3. Generation of a reactor trip by use of t.5e manual safety injection switch; .

4. Clot h the main steam line stop valves;

5. Pettst the feedwater control valves; i

6. C M bg of FN pump discharge valves. t The actuation logic for the functions listed is tested as described in
Sections 7.2 and 7.3. As recutred by Regulatory Guide 1.22, where actuated equipment is not tested during reactor operation it has been ;
~~determined that:~~
~~1. There is no practicable system design that would permit operation of . >~~
- the equipment without adversely affecting the safety or operability i of the plant; ,
~~2. The probability that the protettlon system will fall to initiate the~~
! operation of the equipment is, and can be maintained, acceptably low ! (. without testing the equipment during reactor operation; and 1
~~3. The equipment can routinely be tested when the reactor is shut down.~~
; Where the ability of a system to respond to a bona fide acetdent ! signal is intentionally bypassed for the purpose of performing a test during reactor operation, each bypass condition is automatic 6'Iy Indicated to the reactor operator in the main control room by a separate annunciator for the train in test. Test circuitry does not i
i allow two trains to be tested at the same time 50 thet extension of the bypass condition to redundant systems is prevented. l . 7.1-14 0067F/COC4 4
~~.,--.;,,,.,.,,.n.,,, n ...,w, ,-e . - - , . . , , , , ,-em--- ,,mq yw, ~.,.---y.~~
~~1 IQR l{ 1 7.1.2.9 Conformance of Itti 108 1971 (Reference g) fb See lection 7.6 fo* a discussion of the power supply for the Reactor Trip~~
~~Systes and comp 110nce with !!Et 304.~~
l 7.1.2.10 Conformance to f ttt 114-1971 (Reference 8) ) i ( There are no class I motors in the Reactor Trip System, thus Itti 334 , 1 does not apply. 7.1.2.11 Conformance to 1ttt 144-1971 (Reference 10) , l The setsett testing as discussed in lection 3.10 and the references of i ([~ Chapter 3 confore to the guidelines' set forth in !!!! 344 with the l' enceptions noted in lection 3.10. 7.1.2.11 Confwince to 11t 179-1972 and Renulatory Guide 1.51 ; ' Oune . ' em (Re"erences 12 and 3) The principles described in itit 5td. 379-72 were used in the deston of i the Westinghouse protection system. The system complies with the 'ntent l J of this standard and the additional requirements of Regulatory Guide : 1.53. The formal analyses required by the standard have not been j
~~documented enactly as outilned although parts of such analyses are pub 11shed la various documents such as References 1.12 and 13.~~
i ' Westinghouse has cone beyond the required analyses and has performed a fault tree analysis Reference 1. k The referenced Topical Reports provide details of the analyses of the protection systems previously ende to show conformance with single failure criterton set forth in Paragraph 4.2 of Itt! 5td. 279-1971. The : interpretation of a single-fativre criterton provided by Ittt-379 does not indicate substantial differences with the Westinghouse interpretation of the criterton except in the methods used to confira design reliability. ' Estabilshed design criteria in conjunction with sound engineering practices form the bases for the Westinghouse protection systems. The Reactor Trip and Engineered Safeguards Actuation Systems are each redundant safety systees. The required periodic testing of these systems will disclose any failures or loss of redundancy which could have occurred in the interval between tests, thus ensuring the availability of these systems. I 7.1.3 Electrical penetrations . 7.1.3.1 Desian Sas.es The electrical penetration assemb11es are designed to maintain containment temperature rise integrity underdurin! all design faul current basis events condttions. To assure including that electr1e ( power is continuously available to operate required equipment. penetrations for redundant cables are located in two or more separate l areas in the containment structure. l 7.1-15 0067F/C0C4 l
, N l 50N-6 7.1.3.2 httes Desertetton tither modular or canister t pe penetrations are used for all electrical 6 ;
~~conductors passing through t e primary containment. A double pressure , 6 1' barrier is formed by a heaGr plate at each end of the penetration nortle i~~
~~through which the conductors pass. There are three basic types of ;~~
l elettrical penetration assemblies: high-voltage, instrumentation, and , l low-voltage type. These three types are tabulated into five categories: high-voltage power, nuclear instrumentation system control rod position
~~I indication.10w-voltage power control and indication, and thermocouple.~~
One of each category of penetration assembly is shown in Figures 7.1.3-1 , . through 7.1.3-5. respectively. l The penetration assembly is designed for insertion from the outboard end , of the primary containment nogale and ts welded to the nortle by a weld , ring. Leak test equipment (drain valve and pressure gauge) is provided b ; l i on the outboard end of each assembly. The assemblies are destgned to , remain functional during and after design bests entents. I To provide suitable termination of cables at the penetration, junction bones or covered cable trays are provided inside containment. These i enclosures serve as an electrical spitting bon for fleid connection of conductors or for mounting of connectors. Supplied with each penetration assembly are the necessary special tooling, splice kits, and insulating ' materials that are required for terminating the cables. f thepenetrationassembitesaredestbned, accordance with 1971 edition. ASME iler and fabricated.andinspectedin Pressure Vessel Code. i Section !!!, subsection NE. and are code stamped. ! 7.1.3.3 Tests and Insnections ; The electrical penetration assemblies have been prototype tested, production tested, and field tested after installation for leakage. Prototype leak rate test results were comparable to the spectfled test requirement and found acceptable. The leak rate did not exceed ! 1.0 a 10** cubic centimeters per second of dry helium total for the , prototype assembly when pressurtred to 12 lblin'g with dry heltum in an ambient temperature of 150'F. fach penetration assembly is provided with a pressure connection and gauge to a110w pressurization of the assembly from outside the primary containment. Refer to Floures 7.1.3-1 through 7.1.3-5 for the heation of the gauge-valve assembly on each of the f' ve categottes of electrical penetration assembites. Each penetration assembly has passed the factory production leak rate' test. This requires the assembiles to be { pressurtred with helium out in the open and tested for trdividual leaks O ' using a snubber type leak detecter. I' b l 7.1-16 0067F/COC4 i i l
I I1 l l ) 50N.6 l-. After all assemb11es were installed and became an integral part of the ( i primary containment system, they were leak rate tested. The installed , penetration assemblies were backfilled with dry nitrogen at approstaately i 12 lblin't knd tested to the following requirements: ' ( Pressure 12 lblin's for 24 hours , Ambient ($0'F to 120'F) * ! Temperature > Mastnum leakage rate 1 x 10*8 cubic centimeters for each assembly per second of dry nitrogen i I , I In addition, each conductor was given an insulation resistance test and an electrical continutty test after installation of the penetration assemblies. 1 7.1.4 control Room Dtintavs and Controls 7.1.4.1 Control Room Panels l : The control room panels and the displays and controls are shown in Figure 7.1.41. ,
, t' 7.1.4.2 tafety parameter Disclav system 7.1.4.2.1 System Descrintion k The principal purpose and function of the Safety Parameter Display System - (SPDS) is to aid control room personnel during abnormal and emergency conditions in determining the safety status of the plant and in assessing whether abnormal conditions warrant corrective action by operators to ,
avoid a degraded core. During emergencies the SPD$ serves as an old to evaluating the current safety status of the plant, esecuting function-oriented emergency procedures, and monitoring the impact of engineered
- safeguards or mitigation activities. The SPDS also operates during normal operations, continuously displaying information from which the g plant safety status can be readily and re,lably assessed, toch unit has its own $PDS running on a real time data acquisition and analysis computer system. This computer system also drives display equipment in the Technical Support Center (tSC) and provides plant data
' I to the off-site computer located at thegraphic Emergency Operations (EDF). Each unit SPDS has three color Cathrode-Ray Tubefacility) (CRT : i monitors in the control room, which continuously display *information on ' the Critical Safety Functions (CSF). The operators use keyboards and joysticks to request additional detailed - ( information about the parameters used to determine the CSF status as well as other plant conditions. This information is provided in three l l formats: mimic, tabular, and trend displays. l \ l 7.1-17 0067F/COC4 1
. i l )
1
~~- - - - . - - - - - . . . - - - . ... - . . - _ - - _ - . . _ , _ - - . . . - - ~ . ~ . - . .._.-.-..-_....--.-...'~~
I Ie 1 i l .
~~$0N-6 i~~
The data undergoes several validation steps before being presented to the l operators. When redundant sensors are used, the data retelved by the l l temputer can be processed by sof tware to determine if the quellty of one ; i or more points is questionable. . , j ! 7.1.4.2.2 Det ten Bases ! l Lotat1on of SPDS I i ! The SPDS is conveniently located to control room operators. Two CRis are > 1 ' located inside the horseshoe on panel 417. The third CRT is located . I
~~,iust outside the horseshoe on W16. This CRT is provided primarily for -~~
~~ese by the $hift Technical Advisor (ITA) and Shift Operetton Supervisor ;~~
~~(505).~~
Continuous and Reitable Display of Plant Safety Status Information ' I l 4 The SPDS displays information from which the plant safety status can be ; s readily and reliably assessed by control room personnel responsible for
~~the avoidance of degraded and damaged core events. This is accomplished ;~~
1 by presenting the status of each C$F on every $PDS display. Redundant , sensor algorithms are used to aid the operators in determining if displayed information is reliable, f i , The quality of the information is identified as being good, poor, bad, or ! manually entered. Data is tagged as poor if 1t is inconsistent vith redundant sensors. Data is tagged as bad if it is outside the process [ - 4
' sensor Ilmits, or data acquisition system span, or because hardware l checks indicate a malfunettoning input devlee. Data is tagged as .
manually entered, when the value is operator entered. If a point is not poor, bad, or manually entered it is considered good. Pseudo-points are i tagged as poor if any of their constituent points are not good. ,
l A general ' health indicator" is provided on every SpDS display which I . provides an overall $PDS condition (operating or fatted). Contite Diselav of Critttal plant Variables Y The SPDS provides a concise display of critical plant verlables which provide Information to plant operators about the following critical ' safety functions: 4 I
~~4. Reactivity control~~
~~l- ,~~
b. Reactor core cooling and heat removal from the primary system .

c. Reactor coolant system integrity ,

d. Radioactivity control

e. Containment conditions t
~~7.1-18 0067F/CDC4~~
, . , _ -7.-...,,,...,.-.,,_y -,---y-.. , m-----,,-.-.--e .--
I l l 0 l ,
$QN-6 Table 7.1.41 provides cross reference between these five critical safety )
functions and the parameters displayed by the SPDS. l l l When the $PDS loqtt determines the plant may not be in a safe condition. l l the operator is 'nformed of the problem. After the $PDS indication is 1 i
, vertfled to be correct, the operator is directed to follow appropriate j \ recovery procedures. .
~~i, j~~
j Numan Factors Human factors are taken into account in the design of the $PD$. Color !
coding is used to inform operators of the severity of $PDS alare : conditions. Flashing 15 used to draw operator attention to new altre ; l (t conditions. Page keys are used to page up, down left, and ri ' Alarms are acknowledged with a keystroke at any control PDS room $ght. . j
~~keyboard.~~
i $ i Alpha-numeric information is input from a standard QWERTY keyboard. Additional information is presented to control room personnel in numeric format, numeric displays, deviation barcharts and trend displays. ! I tiectrical and letsmic Quattftcation , j The SPD$ is not qualtfled It and is not powered from a It power source. As such the $PDS is suitably isolated from equipment and sensors used in , safety systems. . ( ' l The $PDS has three power sources: i l Normal: Rectified station unit board AC power inverted to 120V AC Alternate: Station battery 2$0V DC inverted to 120V AC Maintenance: Regulated 120V AC frca 480V AC station unit board The $PDS is not required to operate during or after a seismic event.
~~$PDS equipment is designed so that it will not adversely affect any equipment important to safety, either during or after a seismic event.~~
~~7.1.$ References~~
1. Gangloff. N. C. and M. D. Lof tus "An tvaluation of Solid-State Log c Reactor Protection in Anticipated Transients." NCAP-7706-L. '
~~(Nestinghouse NES Proprietary), and NCAP-7706. September 1971. 5~~
~
2. Burnett. T.N.T., ' Reactor Protection System Diverstty In Westinghouse .
~~Pressurtred Nater Reactors." NCAP-7306. Aprtl 1969. k s 7.1-19 ' Od67NCbC4 , 1 I k~~
. . - _ , _ . . . . _ . - . . _ . . _ . _ . . . . . . . . . . . . , . . . , _ . . _ . . , _ . . _ _ _ . . . . . _ . ~ . . . . - _ . - . . - , . . . _ - _ . .
~~p i 50N-6~~
~~3. The Institute of Electrical and Electronic Engineers. Inc., !~~
i
*!EEE Standard for tiectrical Penetration Assemblies in Containment Structures for Nuclear Fueled Power Generating Stations.* !
~~IEEE Standard 317-1971.~~
t
~~4. The Institute of tiectrical and tiectronic Engineers. Inc..~~
t
*!Ett frial-Use Standard. General Guide for Qualifying class I tjectrical feutoment for Nuclear Power Generating Stations. 5 :
~~IEEE Standard 323-1971. ;~~
~~5. The institute of Electrical and (1ectronic Engineers. Inc..~~
~~'!EEE 5tandard. Inspection and Testing Requirements for Instrumentation and Electrical toutpoent During the Construction of (~~
~~Nuclear Power Generating Stations.* Itti Standard 336-1971. l~~
~~6. The Institute of tiectrical and Electronic Engineers. Inc., i~~
~~'!EEE Trial-Use Criteria for the Periodic Testing of Nuclear Power ,~~
~~Generating Station Protection Systems." !EEE Standard 338-1971. l 7. Division of teactor Standards. Atomic Energy Commission.~~
~~Regulatory Guide 1.22 (February 1972). Periodic Testing of Protection System '~~
~~Actuation Functions." Regulatory Guides for Nater Cooled Nuclear i Power Plants,~~
~~4. The Institute of tiectrical and Electronic Engin6ers. Inc.,~~
*!EEE Trial-Use Guide for Type Tests of Continuous-Duty Class 1 1 Notors Installed Inside the Containment of Nuclear Power Generating Stations." !EEE Standard 308-1971.
~~9. The Institute of Electrical and Electronic Engineers. Inc.,~~
*!EEE Standard. Criteria for Class 1E Electrical Systems for Nuclear Power Generating Stations." IEF.E Standard 308-1971.
~~10. The Institute of Electrical and Electronic Engineers. Inc., '~~
*!EEE Trial-Use Guide for Seisalc Qualif1 cation of Class !
~~tiectrical Equipment for Nuclear Power Generating Stations." IEEE Standard 344-1971. l~~
~~11. Katz. D. N., *5clid-State Logic Protection System Description." ;~~
~~l WCAP-7672. June 1971. ,~~
~~12. Gangloff. W. C., 'An tvaluation of Anttelpated Operational Transients .~~
~~In Westinghouse Pressurtred Water Reactors.* NCAP-7686-L. March 1971 (Westinghouse NES Proprietary), and NCAP-7486. May 1971.~~
13. Salvatort. R., ' Anticipated Transtents Without Reactor Trip in Nestinghouse Pressurtred Water Reactors.* NCAP-8096. April 1973. v

14. The Institute of Electrical and Electronic Engineers. Inc.,
*!EEE Standard. Criteria for Protection Systems for Nuclear Power Generating Stations.* !Ett Standard 279-1971.
~~I5. Erin, L.f. "TopseJ r+ En le-1i Mka'roemer - Bea*4 k u h* h b'&. r~~
~~WW~ I1L1W, -Q%he 191~~
~~\ OdesQm Propelekry cim.rs 1) 7.1-20 0067F/COC4~~
~~%l \ $0N-6 i~~
( Table 7.1.4-1 Critical Safety function to $PDS Parameter Napping 4 .
~~Critleal Safety Function $PD$ ParaM ter(s) '~~
k Reactivity Control Subtriticality , Reactor Core Cooling and Core Cooling 6 Heat Removal from the Heat $1nk - Priatry System Post-LOCA Containment Neat Removal (1 Reactor Coolant System Pressurtred Thermal Shock i Integrity ~ i Radioactivity Control Effluent and Area Radioactivity I
~~, Containment Containment Conditions Containment Effluent and Area Redloactivity~~
~~( l i 1 ( 1 S O 7.1-21 0067F/COC4~~
~~- . - . - . - ._ . _._,..,,y-. ,.-_.,_2 , _. . , _ , _ . _ , _ , . . , _ _ . ,, ._, ,,,,y .,3,. ,_ ...-, -- ,_,,~~
~~$QN-6~~
! l 7.2 REACTOR TRIP sytTEM s 7.2.1 Descrietton - 7.2.1.1 Ivstem Descrietton The Reactor Trip System automatically keeps the reactor operating within
~~) a safe region by shutting down the reactor whenever the Ilmits of the '~~
i region are approached. The safe operettati reilton is defined by several - considerations such as mechanical /hydraul'c l'attations on equipment, heat transfer phenomena and nuclear phenomena. Therefore, the Reactor 6 Trip System keeps surveillance on process variables which are directly related to equipment mechanical limitations, such as pressure,
~~) pressurtzer water Inol (to prevent water discharge through safety valves, and uncoverin heaters) and also on variables which directly +~~
i affect the heat trans er capability of the reactor (e.g. flow, reactor coolant temperatures). $ttil other parameters utilized in the Reactor , Trip System are calculated from various process variables. In any event. whenever a direct process or calculated variable exceeds a setpoint the reactor will be shut down in order to protect analnst either gross damage
~~to fuel cladding or loss of system integrity wh ch could lead to release of radioactive fission products into the containment.~~
~~The following systems make up the Reactor Trip stem: Iteforesw I s.J l~~
1. process Instrumentation and Control System 't '; ; = U

2. Nuclehr Instrumentation System (Reference 2) *

3. Solid State Logic protection System (Reference 3)

4. Reactor Trip Switchgear (Reference 3) l 5. Manual Actuation Circuit The Reactor Trip System consists of up to four redundant sensors and associated $9e; process protection circuitry and two redundant digital ,
logic trains. The W e; process protection circuitry monitors various g plant parameters and provides inputs to the digital logic trains. The digital logic trains develop the logic necessary to automatically open the reactor trip breakers. Each of the two trains, A and 8, is capable of opening a separate and independent reactor trip breaker, RTA and RTB, respectively. The two i trip breakers in series connect three phase AC power from the rod drive motor generator sets to the rod drive power cabinets, as shown on Figure 7.2.1-1, Sheet 2. During plant power operation. 4 DC undervoltage coll
' on each' reactor trip breaker holds a trip plunger out against its spring, I . allowing the power to be available at the rod control power supply icabinets. -
l k 7.2-1 006BF/COC4 i _ - _ . _ _ .r, , _ . -
i
~~7) !~~
i
SQN.6 , j l l For reactor trip, a loss of DC voltage to the undervoltage cell releases ;
the trip plunger and trips open the breaker. When either of the trip i breakers opens, power is Interrupted to the rod drive power supply, and : the control rods fall, by gravity, into the core. The rods cannot be : withdrawn until an operator resets the trip breakers. The trip breakers . cannot be reset until the bistable which initiated the trip is - i
~~re-energited. Sypass breakers BYA and SYS are provided to permit testing l~~
of the trip breakers, as discussed in 7.2.2.2.3. , { ' j 7.2.1.1.1 Functional performante Renuirements + , l The Reactor Trip System automatica11y initiates reactor trip: :
~~1. Whenever necessary to prevent fuel damage for an anticipated 6~~
~~transient (Condition ID. ,! t~~
~~2. To limit core damage for infrequent faults (Condition !!D.~~
~~, 3. So that the energy generated in the core is compattble with the !~~
j deston provisions to protect the rN etor coolant pressure boundary < j for 'Imiting faults (Condition IV). The Reactor Trip Sistem initiates a turbine trip signal whenever reactor trip is initiated to prevent the reactivity insertion that would : otherwise result from escessive reactor system cooldown and to avoid i , unnecessary actuation of the Engineered Safety Features Actuation System. ; l The Reactor Trip System provides for manual initiation of reactor trip by operator action. l 7.2.1.1.2 teactor Tries t The various reactor trip circuits automatically open the reactor trip t l breakers whenever a condition monitored by the Reactor Trip System reaches a preset level. To ensure a reliable system, high quality . design, components, manufacturing, quality control and testing is used. In addition to redundant channels and trains, the design approach ; provides a reactor trip system which conttors numerous system variables by different means, t.e., protection system functional diversity. The ' extent of this diversity has been evaluated for a wide variety of l postulated accidents and is detailed in Reference 7.. s 9 i Table 7.2.1-1 provides a list of reactor trips which are described below. Mi M the at;;tt; ;;;t;;; et:5 =t=t: rn-ter t-' ' " " " jg;;h ;!=0 ::: '; :tter!!; ifrt!n? t t're '" t'! ^^^91 C M 3
~~1. Nuclear Overpower Trips Th'e specific trip functions generated are as follows:~~
~~( 7.2-2 0068F/COC4~~
~~--, vn + , --w, , ,. . - - - , - - . - ~ . - - , '~~
~~.i SON-6' ;~~
I'
~~, a. Power range high' neutron flux trip. -~~
~~L l The power range high neutron flus trip circuit trips the reactor when two of the four power range channels exceed the trip-setpoint.~~
) There are two independent bistables each with their own trip setting (a high and a low setting) per channel (four channels , 6 l total). The high trip setting provides pr6tection during normal power operation and is always active. The low trip set'ing,- c a
~~which provides protection during startup, can be manually bypassed when two out of the four power range channels read above~~
~~, approutmately 10 percent power (P-10). Three out of the four ,~~
~~I channels below 10 percent automatically reinstates the trip function. Refer to Table 7.2.1-2 for a listing of all protection system interlocks.~~
~~b. Intermediate range high neutron fluu trip a~~
~~The intermediato range high neutron flux trip circuit trips the~~
reactor when one out of the two intermediate range channels exceed the trip setpoint.. This trip..which provides protection j during reactor startup, can be manually blocked if two out of
~
four-power range channels are above approximatelyzl0 percent-power (P-10). Three out of the four power range channels below-this value automatically reinstates the intermediate range high-neutron flux trip. The intermediate range channels (including ' 5 detectors) are separate from the power range channels. . The intermediate range channels can be individually bypassed atjthe - nuclear instrumentation racks to permit channel testing at any time under prescribed administrative' procedures and only under the direction of authorized supervision ;This bypass action.ts i annunciated on the control-board.
~~.L l -~~
~~c. Source range high neutron flux trip The w :e range high neutron flux trip circuit. trips the reactor -~~
when ou of the two source range channels exceeds the trip - setpoint. This trip which provides protection during reactor-- - .
~
startup and plant shutdown, can be manually bypassed'when one of ! the two intermediate range channels reads above the P-6 setpoint
, ) value (source range cutoff power level) and is automatically reinstated when both intermediate range channels decrease below .
the P-6 value. This trip is also automatically bypassed by two out of four logic from the power range permissive (P-10). , i s This trip functlpn can also be reinstatod below P-10 by ar j administrative action requiring manual actuation of two control -1 i I - board mounted switches. Each switch will reinstate the' trip-function in one of the two protection logic trains ~. The source-
.~
range trip is set between the P-6 setpoint and the maximum source range level. The channels.can be individually blocked at the. nuclear instrumentation racks to permit channel testing at any s 7.2-3 0068F/COC4 i e w -= -+w- . . - .-e. ...w-e 3 e - .r c.--,ow ,-w.w y + ww y y e-+ey- --wwe- -y-m g- w,w p y "sj
i SQN-6' I'
~~a. Power range high neutron flux trip. ,~~
The power range high neutron flux trip circuit trips the reactor i when two of the four power range channels exceed the trip setpoint.
) There are two independent bistables each with their own trip l- setting (a high and a low setting) per channel (four channeis , 6 ]
total). The high trip setting provides protection during normal power operation and is always active. The low trip setting, ' which provides protection during startup, can be manually bypassed when two out of the four power range channels read above j approximatel 10 percent power (P-10). Three out of the four ; I channels bel w 10 percent automatically reinstates the trip function. Refer to Table 7.2.1-2 for a listing of all protection t
system interlocks.

b. Intermediate range high neutron flux trip The intermediate range high neutron flux trip circuit trips the i reactor when one out of the two intermediate range channels .
. exceed the trip setpoint. This trip, which provides protection during reactor startup, can be manually blocked if two out of
four power range channels are above approximately 10 percent power (P-10). Three out of the four power range channels below this value automatically reinstates the intermediate range high neutron flux trip. The Intermediate range. channels (including
~~, detectors) are separate from the power range channels- The .~~
Intermediate range channels can be individually bypassed at the nuclear instrumentation racks to permit channel testing at any 1 time under prescribed administrative procedures and only under-the direction of authorized supervision. < This bypass action is
~~i annunciated on the control board.~~
~~. 3 '-~~
c. Source range high neutron flux trip The source range high neutron flux trip circuit trips the reactor l when one of the two source range channels exceeds the trip -

setpoint. This trip, which provides protection during reactor .
startup and plant shutdown, can be manually bypassed when one of l the two intermediate range channels. reads above the P-6 setpoint -
; ) value (source range cutoff power level) and 15 automatically reinstated when both intermediate range channels decrease below the P-6 value. This trip is also automatically bypassed by two'
~~out of four logic.from the power range permissive.(P-10).~~
~~3 This tr1p functtpn can also be reinstated below P-10 by an g administrative action requiring manual actuation of two control~~
~~< i '~~
~~board mounted switches. Each switch will-reinstate the trip~~
function in one of the two protection-logic trains. 'The' source -

range trip is set between the P-6'setpoint and the maximum source range level. The channels can be Individually blocked at the nuclear instrumentation racks to permit channel testing at any 1
~~7.2-3 0068F/COC4 i~~
~~. . . _ . _ - . . . _ . . . . . . . _ . _ . . _ . , _ , . . _ _ _ _ . _ _ _.,.a.._..__, _ . . . . _ . . _ ,~~
^
ff j i 1 SQN-6 i time under prescribad administrative procedures and only under
~~-the direction of auths tred supervision. This blocking. action is annunciated on the centrol board.~~
d. Power range high positive neutron flux rate trip This circuit trips the reactor when an abnormal rate of increase J in nuclear power occurs in two cet of four power range channels. .
This trip provides protection inst rod ejection accidents of a low worth free old-power and is always active.
e. Power range high negative neutron flux rate trip-This circuit trips the reactor when an abnormal rate of decrease ,
in nuclear power occurs in two out of four power range channels. This trip provides protection against dropped rods and is always active. Figure 7.2.1-1, Sheets 3. 4 and 5 show the logic for all of the nuclear overpower and rate trips. A detailed functional , description of the equipment associated with this function is given in Reference 2.
2. Core Thermal Overpower Trips The specific trip functions generated are as follows:

a. Overtemperature AT trip This trip protects the r. ore against low DNBR-and trips the reactor on coincidence logic as listed in Table 7.2.1- I with one ;
set of temperature measurements per g The setpoint for this . trip is continuously calculated by ......, circuttry for. each loop by solving the following equation:^ i re AT. fg7 N ( 1 ) <AT - -P') - fi(AI)} i +tss l+tas , Nhere: 1 = Lag compensator on measured 49- Tkg 6
~~' " b1 I s lag ae massWen Mansured Teeld I + *'% $~~
~~tg* gI . Ties constants uttitred in the lag compensator for~~
??: r - - - 71,/ and Te,fd % d[ . Indicated AT at RATED THERMAL POWER Ki f, 1.15-K 0.011 1
~~7.2-4 0068F/COC4~~
. . . . . . , . . . . , . . 1 _ . .l ....g....
e
. = ...__..._..........._............~._.:.. . . . . .l .4 % .. .._ t ..._ A . [ ...] N, J. . .. k. s. N. ./ .4 Tn A j . .. . / + Tg1 . . . . . . . . _ . . . . . .. _ .. .. e-Tgs ... 4.._... . a t
__ . . . . . . . _ . . - . . _ . . _ . ._ __ ___..... ..._f .. . _ . . . . . . . . . . . . . . . . . _ .
.. . . . _ . - .... s . .. . ..p.. .... .. . . . . . . . . . -.. _.. ._.._ J'te St h .&. ItJt t .a n . .ZMEAT...
I i .... . . _ . . . . . _ . . \. . . - . $ & . a, . ... . . . . _ _ _ ._. . . . _ . _ . .._.. . . __ .. . . _ . _ . . . . -. _ .. - ..~ . .. _* . .
~~... _= . .. . ..~~
~~L-. y~~
.._.c ._.__._ . .. . . _ _ - . , . _ . . . . . . _ . . . . . . . . . . . .
~~z. .~~
~~J. 1.. n J y. &. .. $l3~~
~~._._ DOf . .J._-- - .._ -.. . -. ...- . . _ _ __ _.. .. . . .h.. _t~~
~~_ . . .. . .y.. .n. .s .T.. _ _H __.q~~
. . .. _ . .f . . . . . . ,. ._. _ .. . ., . ._ ._v_.. . _ _ _. ... . ).
~~_._..]~. .h. .s . . /A ./A. .1.a_*.. ((. .r.~~
- .. . ___ -_. ._ . ._ . . . .C V9 ,* .x_ . . . . . . . . . _ _ . . . . _ _ . . . . . . . -_ .&. . .li ..k'TO.c . . . . . . .er- . .&. .. . .= . . .. . f> .. . rt . .&. .. . p m t....I @ d.. ...I.A . . _ . . . . . . . . . . . . . . ~ . . . . . . . _ . . . . . . . . . . . . . . . . . . . . . . . n. , .._ ..... . . ... .. _ ... ..... ..._. i . . . . . . . . J f MC . . ran . .~1 . . . . . . . . . . . . . . . . . . _ . . . . . gy g . _. . . . , . . . . . . ....... . . .. .. . . .
_ . . . . . _ . .. . ...J . . . . _ . . . . . _ _ . . . .) . . .,. 7 9 .. . .. . . . . . . . . . . .
~~..&..-..,-f...........-......_-....._-.-....-.-.......- .. ... .. ., r.~~
~~3.. . . . . . . . . . . . . . . .. . . .~~
. . . . .. .. _ . JJ ._I .... j . .. . . . . . . . . . .n j. . . . . _ . . .fcy.. .. . .._.._. ._ . ... .. ... . . . _ .........f..._..............._._...__..._...__....__..... .... ... . ...... . .... ,. . . .L.. S. %. . _. /. ... .p... ..... .. .. .g, . . .
~~g .. n ..... _~~
~~... L ... ... . .. .. .~~
(
... .. ._ g . . . . . . . . . . . .. . --. .
~~s'._. 0 0 6LA . .. ..AY.. . . . .. .hR;h4k .. .. .~~
~~.*g. .~~
~~6WQ . . .~~
. ... . . . . . . . . . . . . 1 1 . .. . . . . . . . . = ._... . _........... _ - r y- = -
_s. . - _ _ .. . . _ . . _ . . . . . _ . . . ..1 _( -
. . . ~ . . - - .1 .\ . . . . . . . . . . . . . . . . . - . . . . . _ . _ _ . _ _ _ . . .
~~i _ . _ . ... . _ . . . . . . . . . . . . . _ _ . .~~
- - . - - . --_._- --- . ._ -- --. - .-- - - ~ .- - -- . e- If TuS 74 Gs% y 4 y / t 1gS r /*1tecuredhasemk]l Na lead ./9 c&oller ' ' Mar [ Tg- he $u44r w 9; wh//M /9 Neo MrCrI*$
I. - manw e . rect. s Odt'N Jec f. N~~" l \, 1 + tes . Thefunctiongeneraledbythelead-lagcontrollerfor l , 3 + tis T... dynamic compensation , L sg & tt . Time constants uttitred in the lead-lag controller for l 9 T.... sg = secs., tg . secs. l T w L . . . .. G ,,. . . ; .; . " 7 , l - ~ i _ i...-_...u. _ _ . . . . . . .
~~-y -7 .wwww- -ww --www ww ig.g f~~
~~l , e, -~~
~~'*r r Stet etrd '- t'0 rne ed ?.., ;~~
('
~~:n ner. :. = 2 :- :.~~
. 2 : n .2 r = = = n ... a = e : = pw => -
~~4 0.00037~~
~~'K' .~~
P = Pressurtzer pressure, Ib/in's Pe = 2235 lbrin'g (Nominal RCS operating pressure) S . Laplace transform operator (sec -') and fi (AI) is a function of the indicated difference between top and bottom detectors of the power-range nuclear ton chambers; with gains tb be selected based on measured instrument response during plant startup tests such that: ( i
~~. (1) For q .- q. between -29 percent and +5 percent fi(A I) .~~
0 (where q. and q. are percent MTED THERML POWER in the top and bottom halves of the core respectively, and ei + q. Is total THERML POWER in percent of MTED THERML POWER). (11) For each percent that the magnitude of (q. - q.) exceeds -29 percent, the AT trip setpoint shall be automatically reduced by. 1.50 percent of 1ts value at MTED THERML PONER. (111) For each percent that the magnitude of (q. -q.) exceeds +5
~~percent, the AT trip setpoint shall be automatically reduced by 0.86 percent of its value at MTED THERML POWER.~~
~~. The one pressurtzer p'ressure parameter required per loop is.~~
obtained from separate sensors which are connected to three pressure taps at the top of the pressurizer. The four pressurtzer pressure signals are obtained from the three taps by-connecting one of the taps to two pressure transmitters. Refer to Subparagraph 7.2.2.3.3 for an analysts of this arrangement. i i Figure 7.2.1-1 Sheet 5 shows the lott'c for the overtemperature AT. trip function. A detalled functional description of the process equipment associated with this function is contained in Reference 1. L
. 7.2-5 0068F/COC4 . . ~
~~_, .,..e_,..,_.... ,.wme. ,,..e.,,,w_-.,,,.-. .,.,.p.,.,, s.y...-- , y ,, ,,,,,y-,p .~~
-) g ., 1 l
~~d I SON-6 l~~
b. Overpower AT trip i l .

t. This trip protects against excessive power (fuel rod rating - i l
8 prot 6ction) and trips the reactor on coincidence as listed in Table-7.2.1-1. with one set of temperature measurements per loop. The setpoint for each channel is continuously calculated using the i following equation: P i L - g Overpow gg i 3' ( 1 )<AT / )-T.K. 1+t.s
~~. f (AI)}~~
~~_ .vii + tis 1+t.s Where:~~
~~- L;; ;; ;;7;;;e7 en ; ;;ered 4T ;~~
~~b ..; e- -~~
~~"r :r St:-t: et!::1 '- th: ?:; :; ;::::t:r " r ??, ; . 2 ;;;; .~~
1 y( = Indicated AT at RATED THERMAL POWER i K. I 1.087 Ki = 0.02/'F for increasing average temperature and 0 for decreasing average temperature tss . The function generated by the rate-lag controller for I + 53s T.., dynamic compensation j = Time constant ut1112ed in the rate-lag controller for 53 : l T. . . , ty . secs..
~~- L;; ;, .,,;..;;;r er, ;.;.;er;d 7..,~~
L .. : t- = "= ::: t::t ;t'it;;d 'r, th; ;;;er;d 7,,,
~~;r ;;r.;;.ter. .. ;;;;.~~
g K. . . 0.0011- for T > T" and K. 0 'for Ti T" L i - fc;.r;;. t.g;7; tere T- - Inun.. w J 7.., .6 ^e 7;; Ts;; A 'GW: (;.M L. . J.vi,
~~) ;.g:r;ter; f;r T :::tt: ::t:t!; , ; 5?".:'"~~
S = Laplace transform operator (sec -') , f (AI) . O for all & I i
~~AS Aefud hr 0 .07~ 4rq~~
#**al Ta'$
V5 i a+ W M*'d P ll ' TL M4 entrub:) Ly 44 . lea)-83 fr mwred _ O&AAar W68F/COC4-i
* % f Tc % wk v.u) In %. 7. 7. Jend- -4 h_ "Tw- CedroII4 . .sto. for ^^enuA <[c s W I - ,v w a wee
~~rrw-e sew-, w m. .-av-- - - - - - - - - - - - - - - - - - - u -- v --. e-- --m-n -~~
.. . . . . . _ . . . . . . . . . _ _ . . . . . .. . _ . . . . . 3. ,t -
~~3 Tj . .. n. . m ._l......L._. Zh7;o. As../~~
~~%.. [ av~~
~~w. . .r,. J 1~~
._ . _ . . . _ . . .m. .gr . . _ . . .L . . _ .y .. . _._ ... . .. . . . .. . . . . . . . . . . . . ...__ . _._ Q_ _ _. _ i. __%....._... ._v . ...(. ... .. ... -.... .... .. . .. y. ...j ;-
~~_.... .. ...f.(.a) 9. ... . } .. ._... .~~
~~.._ y tw ar .~~
_ . . _ _ . u...No .. . e. . . J . . . . . . _ _ . _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . l.._ _ _ . . . _ _ __.______ . _ _. _ _ _ _ .. _ _ _ _ . _ . ._ ..... .
. . . . . . _ . = ___..___ ___ __ __ . _ . _ _ . . . . . . . . _ . _ . . . . . .
I
.L......_._.._.._ _ _ _ _ . _ _ . . . _ _ _ _ . . . _ _ _ . . _ . . . . _ _ . . . . . . . . . ... - . . = . _ _ . . _..... _ .. .._. .. . . _ .
l . . . . . _ . . .. . . .. . . . . . . . . . . ..
~~.( . ._ .__. . .._ . - _ . .. o Ts._. ._.7 ci _A. ._ r . m. ws. k 1. _~~
~~_. A , . _ .. ._- . . . . 4 ..~~
~~- - - _ L . __ . _*.. .% ass _b aiwed.._Ma_i_3k. . ...__ ... .hn_~~
~~u B. __.A.a r_ . ..oJer%.. .perofo. ..re _ma. __Du..er. .._ __ ar__ _ _27 #_i e ..y. a ta }a s~~
._sre _e /Teh. , . 1.. a . _ -4.___ ..-
~~v../_.a. b_re .a_ o. . .. rew,o...... u _.:...___. _m 4~~
~~. .. d. e... .. .. _ h_ .__.._..J~~
~~_'( il.F._. . . ._nmesenuh.._Je_.Sh _ _ _ .. _ _ . ..A. .~~_3EC . . . _ _ .<~~
.M.asa __ _evw,, _ .___.... .9. iK. a. _.._M. ..- . . +.. . . . . .
k.__ ._ h.aam. rw3iF1.e.N _h _. . rs o cLs.......... . ..
... . ....c s . n a e e.... e x s e.+ . . . ..re/h. .. k. 6..._.'_..E ... .c & __._a.... . .h. . /,4[J2.. .+.
~~_ . . . . . . _ _ . . .:o . i.~~
% , . . .44 cm,.r.A. . . ... . # ocx4,cj2dwe . Lab . A. r. . .(
~~k . d. A a..J~~
. . . . d ia,.._. F. . mk. - ea2 . . . . .. .aies . .. . - _em _ .. . . ... , _ . . Te.d. .nic s.l . rpa c.. A.. enM. o. s., - . . . . _ ._ _ _-. . . . .. . . _d. __Ar+A .
_ . . _ _ . . . . . _ _ _ . _ _ _ _ __._.. = . _ .. ._. ._ ._ ._ .. .. _. ._. .. ._ ......_. .. . . _ . . . ._ 7
~~- ,- r-~~
~~SQN ; t The source of temperature and fluu information is identical to that~~
~~' ~~~
~~, of the overtemperature AT trip and the resultant AT setpoint is l compared to the same AT. Figure 7.2.1 1 Sheet 5 shows the logic~~
. for this, trip function. The detailed functional description of the t '_' process equipment associated with this function is contained in ,
Reference 17, I 3. Reactor Coolant System, Pressurizer Pressure and Level Trips , l' The spectftc trip functions generated are as-follows:
~~a. Pressurizer low pressure trip~~
) The purpose of this trip is to protect against low pressure which could lead to DNS and limit the necessary range of protection afforded by_the overtemperature AT trip. The parameter being '
sensed is reactor coolant pressure as measured in the pressurizer. Above P-7 the reactor is tripped when the - , compensated pressurtzer pressure measurements fall' below preset Ilmits. This trip is blocked below P-7 to permit startup. The trip logic and interlocks are given in Table 7.2.1-1, t T.he t.r..ip
~~. . . .. ..... logic 1s..s.h.own .. . m.o,n_Figure~~
~~_ _ _ . 7.2.1_1~~
~~. . _ . .Shoe.t~~
~~_ _ _ .... 6. .^. f.a'-".ed m .u~~
~~, '.C' .'l.:'.'; i . i:' Z: us_n i;'.on':' .'..:._ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. r - " '" ' ~ ~ ~ ' ' '"~~
~~b. Pressurtzer high pressure trip The purpose of this trip is to protect the Reactor Coolant System against system overpressure..~~
~~( ;~~
, The same sensors and transmitters used for the pressurtzer low l- pressure trip are used for the high pressure trip except that separate b1 stables are used for trip. These blstables trip when
. uncompensated pressurtzer pressure signals exceed preset ilmits L on coincidence as Itsted in Table 7.2.1-1. There are no interlocks or permissives associated with this trip function.
~
The logic for.this trip is shown on Figure 7.2.1-1. Sheet 6. -4he-3::: M:t 5:: !:::: f:::r';t": c' M: ; : : :r";-- ' _ - a s ;vi t aud .~'. ", "' '; ' " ;-- ^' d:I ' ."."'---::^ ' . I c. Pressurtzer litgh Water Level Trip This trip is provided as a backup to the high pressurtrer pressure trip and serves to prevent water relief through the pressurtzer safety valves. This trip is blocked below P-7 to peralt startup. The coincidence logic and Interlocks of L ) pressurizer high water level signals are given in Table 7.2.1-1. , 7.2-7 0068F/COC4
~~/ .______,..___m._ , , , _ , _ _ , _,~~
~~( {g !~~
~~$QN-6 The trip logic for this function is shown on Figure 7.2.1- 1 ^~~
~~Sheet 6. 1:t,--d 1::: '-t b - 0' th: F :- : : ;.;-^^.0 t l~~
~~4. Reactor Coolant System Low Flow Trips l ,~~
These trips protect the core'from DN8 in the event of a lovs of coolant flow s1tuat1on. The means of sensing the loss of coolant ' flow are as follows:
4. Law reactor coolant flow The parameter sensed is reactor coolant flow. Three elbow taps in each coolant loop are used as a flow device that indicates'tt.e status of reactor coolant flow. The baste function of this device is to provide information as to whether or not a reduction i in flow rate has occurred. An output signal from two out of the three bistables in a loop would indicate a-low flow in that loop. ,
This tr1p is blocked below P-7 to permtt startup. 6 The coincidence logic and interlocks are given in Table 7.2.1-1.: -
., u. -__.... - - . . . - - - . . n. a......a ....... ... 2.......... .~.:_:::::. .J . . ... . . .. .' ". .'. '. . .. 'Z: '. ';.. . ;il.'s . 'a: Z : . O 'i r .';'L ; . ' . . , . . . . . . . . . . . . . . . . . _ _ . ~ . _ _ ,
~~Figure 7.2.11. Sheet 5. shows the logic for the Reactor Coolant System low flow trips. At power levels above P-7 and below P-8. low flow in two or more~~
~~-loops causes a reactor trip. Above P-8. Iow flow in one loop causes a reactor trip, 6~~
~~b. Reactor coolant pump undervoltage trip -~~
This trip is required in order to protect against low flow which can result from loss of voltage to more than one reactor coolant pump (e.g. from plant blackout). There is one undervoltage sensing relay connected to the load side of each reactor coolant pump breaker. These relays provide an output signal when the pump voltage goes below approutmately 70 percent of rated voltage. ~ Signals from these relays are time delayed to prevent spurious trips caused by short term voltage perturbations. The coincidence logic and interlocks are given in Table 7.2.1-1. This trip is blocked below P-7 to permit startup. g
~~c. Reactor coolant pump under frequency trip This trip is required to protect against low flow resulting from bus under frequency; for example, a ma,jor power grid frequency -~~
~~disturbance. The function of this trip is to open the reactor coolant pump (RCP) breakers and trip the reactor for an under - L l 7.2-8 0068F/COC4- .~~
* -- e w -- -- b______________h
a SQN-6 j frequency condition.- The setpoint of the unoer frequency relays ' ' Is adjustable between 54 and $9 Hz. l There is one under frequency sensing relay connected to the load side of each reactor coolant pump breaker.. Power level above the. : P-7 setpoint and an underfrequency condition sensed by more than
~~) .one reactor coolant pump motor resu l ts in the tripping of all of 6 ;~~
the reactor coolant pump breakers as well as directly tripping . ! the reactor. Signals from these relays are time delayed to i prevent spurious trips caused by short-term frequency perturbe-tiens. Undervoltage sensing relays are provided across the power !
. feed to each under frequency sensor in order to ensure that each under frequency input to the Reactor Protection System will I indicate an under frequency condition entsts on loss of power to the sensing device. The. contacts of this undervoltage relay are in series with the output of the under frequency sensing relays in each channel. Figure 7.2.1-1 Sheet 5 shows the-logic. .l6:
As shown in F1gure 7.2.1-1, Sheet 5, the only inputs to the .% Reactor Protection System (RPS), associated with the RCP come from the undervoltage and under frequency sensors. . These sensors are located on the. load side of the RCP breakers, within a ; Setssic Category I structure, and are designed in accordance with t the requirements of IEEE 279-1g71. The trip signal for the reactor trip breakers, associated with '6 - the under frequency condition, is an output from the RPS, as shown in Figure 7.2.1-1, Sheet 5. l6 The Westinghouse analysis of the loss of flow accident has.shown that for frequency decay rates less than 6.8-Hz per sec, no RCP tr1p is necessary. TVA has-performed an analysts-to confirm that the worst case frequency decay rate at the RCP input terminals is i l below this 11mit. The results of the TVA analysts shows a i frequency decay rate of less than 5 Hz/sec l
~~5. !* e- 00 : it '9;; M ".DOMW y ins ay wii;6 ..;p 7.A;t h;; ;;;; t:f e : !! *^" ~! -~~
-:_ y e-- y + e n~ senn Ni a i h,hivI^!I^ I^ 7CCCI ~~[7[ { . C^_^ 7 ". Of 'III ,
~~a ina ="=,is _ = = _-i r- ~ ; - --- y,g y 7.pu, -- , n i*!..'.'",_?"...'"..I'.".".""'*"'"*"'"'~~'--'~--~~~~
~~. ~ . , . . .~~
ywi w .. , ...t , e www a New wy : $ bu Ns e Iiih [wA$I wC. ; II; ;;; :: Mt: M:i: ":: : ht-d 9 th th h t-4 i r I ! 7.2-9 0068F/COC4 e n _ .--_. . ...-... ..._ ... _- . . . . . - , . .-~ ~-- .-, - ...... ..--.r, - - . - . - ~ m.~.-4 .
1 O INSERT 4
~~5. Low-Low Steam Generator Water Level Trip .~~
+
(Including Environmental Allowance Modifier and Trip Time Delay) ! This trip protects the reactor from loss of heat sink in the event of ; a loss of feedwater to one or more. steam generators or a major g feedwater line rupture. This trip is actuated on two out of three low-low water level signals occurring in any steam generator. If a low-low water level condition is detected in one steam generator, signals shall be generated to trip the reactor and start the motor . driven auxiliary feedwater pumps. If a low-low water level condition is detected in two or more steam generators, a signal is generated to g start the turbine driven auxiliary feedwater pump as well. This trip includes an Environmental Allowance Modifier (EAM) which distinguishes between normal and adverse containment environmental , conditions. The EAM' selects a low setpoint for the steam generator low-low level trip which includes an environmental uncertainty associated with normal plant conditions. In the event that an adverse containment condition is sensed by the EAM, a higher steam generator low low level trip setpoint is automatically selected to account for - larger environmental uncertainties. associated with the harsh environmental conditions due to a feedwater rupture inside containment. By utilizing the two different setpoints, more operational flexibility is provided during normal conditions, while adequate protection is still provided during accident / adverse conditions. In addition, the signals to actuate reactor trip and start. auxiliary feedwater pumps are delayed through the use of a Trip Time Delay (TTD) system for reactor power levels below 50% of RTP. Low-low water level. in any protection set in any steam generator will generate a signal. which starts an elapsed time trip delay-timer. 'The allowable trip _ time delay is based upon the prevailing power level at the time the low-low level trip setpoint is reached and the number of steam generators that are affected. If power level rises after the trip time delay setpoints have been determined. the trip time' delay is- i re-determined (i.e., decreased) according to the increase in power l 1evel. However, the trip time delay setpoints are not: be changed if the power level decreases after the setpoints have been determined. I The use of this delay allows added time for. natural steam generator > level stabilization or operator intervention to avoid an undesirable 1 inadvertent protection system actuation. The logic is shown on Figure 7.2.1-1 Sheets 17, 18, & 19. ! I A i i 4
~~, _. .. ~ . . . - -._,m. ~ . - . . . , . - - . - . - - . . . . . - - . , - - . . . . __ _ . .-~~
~~3f I SQN-6 j~~
~~. ....,_2 , .___, >____,_._,u_ _______ __.. ____.~~
~~J'"'r's'" ' r" .. in -" nwiw1=n6= s.~~
".. "..n ... _' .' . .': .' 1". P. .' '". !! ' ".i.. p. o' n.v.w' "i: ..,u.. ". m."'
~~9. '.= ! = :t n: ;n n; O n ts !^.n ! t '- "~~
ts trip protects the reactor free loss of heat sink in eve sustained steam /feedwater flow misant insufflcten tude to cause a low fee low reactor-
,tr1p. This tr1p 1s ed on of three low-low water. i level signals occurring i am generator.
l The logic- on Figure 7.2.1-1, Shoe detalled fun description of the process equipment as d with-is trin is nrovided in Reference 1.
~~6. Turbine Trip-Reactor Trip ,.~~
The turbine trip-reactor trip is actuated by two out of three logic l from low autostop oil pressure signals or by all closed signals-from ) the turbine steam stop valves. A turbine trip causes a direct -I lb l reactor trip above P-9 setpoint. The reactor trip on turbine trip is an anticipatory trip input-signal g to the reactor protection system. This trip is anticipatory in that it is not assumed to occur in any of the Chapter 15 accident - analysis. This trip meets all of the requirements of IEEE 279-1971-including separation, redundancy, single fatture, and testability. l6
~~. Seismic location, qualtf tcation, or mounting of the sensors is not practical because of their= location in the nonsetssic Turbine Butiding.~~
High-high steam generator level signals in two out of three channels for any steam generator will actuate a turbine trip, trip the main I feedwater pumps and close the main and bypass feedvater control valves and main feedvater isolation valves.. The purpose is to-g protect the turbine and steam piping from excessive moisture - ,
. carryover caused by high-high steam generator level. Other turbine '
trips are discussed in Chapter 10. The logic for this trip is shown on Figure 7.2.1-1 Sheet 7. The. analog portion of the trip shown on Figure 7.2.1-1, Sheet 16, is represented by dashed ( - -) Itnes., 14 hen the turbine 1s tripped, turbine auto stop oil pressure drops, and the pressure is sensed by three pressure sensors. A digital output is provided from each sensor when the oil pressure drops below a preset value. These three- , outputs are transmitted to two redundant two out of three logic matriuts, e1ther of which trips the reactor if above P-9-setpotnt. l6-7.2-10 0068F/COC4 _ ._ _ _ _ . _ _ _ . . _ _ . _ . - - _ _ _ _ . _ _ . _ __ _ _ _ . _ . _.2
f 3y- ! 50W-6
~~l. .~~
{. l , The auto stop' oil pressure signal also dumps the stop emergency trip ! fluid, closing all of the turbine steam stop valves. When all stop l valves are closed, a reactor trip signal will be initiated if the
reactor is above P-g setpoint. This trip signal is generated by 6- i redundant (two each) limit switches on the stop valves.
i
7. Safety Injection Sign'l a Actuation Trip i A reactor trip occurs when the Safety Injection Syst'm e is actuated.
~~The means of actuating the Safety Injection. System are described in. Section 7.3. This trip protects the core against a loss of primary~~
~~,; or secondary coolant. b- -!~~
9"[?_M*h!'!II?!_8Lshows {pg_}ogjc fo[_ ppis {{{p. g j;jp ygf, _ , , _ rr urn,x rrrm c.nn:xcr - ----- --- -- -~ -
~~8. Manual Trip The manual trip consists of two switches with two outpu'ts on each [~~
switch. One output is used to actuate the train A trip breaker, the other output actuates the train 8 trip breaker. -Operating a manual trip switch removes the voltage from the undervoltage trip coil and energizes the shunt reactor trip breake'r trip cotl. . , 6 q There are no interlocks which can block this trip.- Figure 7.2.1-1 Sheet 3, shows the manual trip logic. 7.2.1.1. 3 Reactor Trio System Interlocks
1. Power Escalation Permissives l- The overpower protection provided by the out of core nuclear i l
Instrumentation consists of three discrete, but overlapping, levels, l Continuation of startup operation or power increase requires a permissive signal from the higher range instrumentation' channels before the lower range level trips can be manually blocked by the. operator. A one out of two intermediate range permissive signal (P-6) is- 6 required prior to source range level trip blocking and detector high vol tage . cutoff. Source range level tr4ps are automatically reactivated and high voltage restored when both intermediate range channels are below the permissive (P-6) level. There is a manual-
~~. reset switch for administratively reactivating the source range' level trip and detector high voltage when between the permissive P-6 and P-10 level, if required.~~
4 l' Source range level trip block and high voltage cutoff are always - maintained when above the permissive P-10 level. 7.2-11 0068F/COC4 l 0 b
~~- ..~ _ _ - _ _ __ _ . _ _ _ _ _ . - _ _.___ ,... _ . _ . . _ _ _ . . _ ,~~
a
~~-3V i i~~
~~SQN -~~
'The intermediate range level trip and power-range (Iow setpoint) trip can only be blocked after satisfactory operation and permissive t information are obtained from two of four power range channels.- ,
Individual blocking switches are provided so that.the low-ran e power-range trip and intermediate range trip can be independently b ocked - reactivated when any three of ithe four power range channels are be foreachtrain.-Thesetripsareautomatica11{owthepermissive l$
~~.(P-10) level, thus ensuring automatic activation to more restrictive trip protection.~~
The development of permissives P-6 and P-10 is shown on Figure i 7.2.1-1, Sheet 4. All of the permissives are digital;- they are - -f derived from analog bistable signals from'the four power range and . - the two intermediate range channels. b-
~
~~See Table 7.2.1-2 for the 11st of-protection system Interlocks.~~
l6

2. Blocks of Reactor Trips at Low Power Interlock P-7 blocks a reactor trip below approutmately 10 percent of full power on a low reactor coolant flow in more than one'1 cop.-
reactor coolant pump undervoltage and under frequency, pressurizer low pressure or high water level. See Figure 7.2.1-1, Sheets 5 and 6.. Interlock P-8 blocks a reactor-trip on 2/3 low reactor coolant. flow in any one loop. when-the-plant is below approntmately 35 percent of- -i full power. The block action (absence of the P-8 1nterlock signal) occurs when three out of four neutron flux power range signals are. below the setpoint. Thus, below the P-8 setpoint, the reactor will be allowed to operate with one inactive loop and trip will not occur .i untti two loops are indicating low flow. -See Figure 7.2.1-1. Sheet - c 4, for derivation of P-8. and Sheet 5 for appilcable logic. Interlock P-9 blocks a reactor trip following a turbine trip below 50 l percent power. The block action'(absence of the P-g interlock signal) occurs when three out of four neutron flux power range'
, signals are below the setpoint. Thus, below the P-9 setpoint, the l reactor will not be directly tripped by a-turbine trip, but-instead-the reactor control system and the steam dump system will auto-6 natically control the reactor to zero power conditions. See Figure- ..
7.2.1-1 Sheet 16 for the implementation of the P-9-interlock. See Figure 7.2.1-1 Sheet 4 for the derivation' of P-9. See Table 7.2.1-2 for the list of protection system blocks. I 7.2.1.1.4 Coolant Temperature Sensor Arranoement O,e T...si;ene; Je.i. ;,,t; en ef ti.e rees; u see h n; eja:e ;;7 ,;. .;. m
~~;en.-.;;w ; ail.;;aii ;;r.;er ;,77;;;; ;;t h ;h:2 " S::t h: 5.5. $~~
LEE' ZIMEAT [ , f ll k k j 7.2-12 0068F/COC4
~~-, . , - .. _ .._ ,- -__ . . - - . , - - _ _ . . _ _ . . _ . _ . _ _ _ . . . . . . . _ - . . - . - . . . . - . ~~~
i INSERT 5 PAGE 1 0F 3 - l 1 7.2.1.1.4 coolant Lahm sensor Awwi ard Nculational Methociolocrv: The individual narrW range cold ard hot leg tanparature signals required for igut to the reactor trip circuits and interlocks are obtained using RID's installed
~~-in each reactor coolant loop. j 1he cold leg tanparature measurement en eam loop is acocuplished ,~~
with two narzw range RIDS mountad in thermowells. The cold lag- . sensors _are irhely redundant in that either sensor can " adequately represent the cold leg tanparature sensurument. 4 Tenparature streaming in the cold leg is not a concern due to the [ mixing action of the reactor coolant punp.- '
'Iha hot leg tauperature measurement en each loop is = -- 711shed with three narrw range RIDS mounted in thernowells spaced 120 degrees apart arcund the circumference of the reactor coolant t pipe for spatial variations.. 'Ihese cold and hot lag narzw range RID signals are input' to the protection system digital electronics and gh as.follows: 'Ibe two cold leg temperature signals are subjected to range and; i consistency checks and then averaged to provide a group value for l T cold.
l l If either T cold input signal is cut of range high or lw, it l will be set to the high or low limit respectively. - \ I Next, a consistency check is performed en the T cold input l signals. If these signals agree within an %W.se interval _ : l (DEIRC), the group quality is set to GOOD. 'If the signals do-i not agree within the acceptance tolerance II2nc, .the grcup quality is set to BAD, and the individual signal qualities are set to POOR. The average of the two signals is used to 4 A the group in either case. II2nc is a fixed input parameter based en operating experience. One DE2nC value is required for and protection' set. i , Each of the three. hot leg tanparature signals is subjected to a ' range check, and utilized to almlata an estimated average hot leg tenperature which is then consistency checked against,the' other two estimates for average hat leg tamperature. ,' ! l I J 4 l L I 1 1- l j.
1 INSERT 5 PAGE2'0Fi N ( If any T hot input signal is out of range hi@ or lw, it will be set to the offscale hi@ or low limit respectively. . Next, an estimated average het leg tenparature is derived frun each T hot input signal as follows: I
~~e T,=g g -gSjj where I !~~
Thj is the filtered T hot signal for the jth RID (j = 1 to 3)~ in ' the ith loop (i = 1 to 4) Pai = power fraction being used to correct tho' bias.value being used for any power level P y =(T(g-T[.)/4T[ where l 4T[isthefullpoWardTintheithloop sh = manually input bias which corrects the individual T het Rm value to the loop average.
'Ihan, an average of the three estimated hot leg tenparatures.is cxmputed and the individual signals are checked to determine if they agree within i DEI 2A of the average value. If all' of the signals do agree within i DE12A of the average value, the group quality is set to GOOD. 'Jhe groqp value (T/w average of the three estimated average hot leg *M) tures. is set to the If the signal values do not all agree within i III21 of the ~
~~average, the algorithm will delete the signal value which is~~
' furthest frun the average. 'Iha quality of this signal will be-set to RXR and a consistency check will then be performed on the remaining GOOD signals. If these signals pass the consistency check, the group value will be taken.as the average of these GOOD t
signals and the group quality will be set to POOR. - However, if these signals again fail the consistency check (within M2HA), then the group value will be set to the average of these two signals; but the group quality will be set to BAD. All of the ! individual signals will have their quality set to RXR. DE12A is a fixed input parameter based upon t @ 61.dre J distribution tests within the hot leg. One IIIEA value is required for each protection set. l H l 1
~
U INSERT 5 31 ' l PAGE 3 0F 3 l I L Delta T ard T average are calculated as follows: ! 6Ti = TI.wer- Tie 3
. i ' Tov3;=(Tbwg+(.)/2.0 1 + 'Iha calcallated values for Delta T ard T Average are than utilized for both the remaindar of the overtanparature and overpower Dmita~
4 T protecticri channel and dannel outputs for cx:ritrol purposes. t I l [ i i l i i i I 4 e i i i .j e a 4 i n
M 7.2.1.1.5 pressurizer Water Level Reference Lee Arranoement The design of the pressurizer water level lastrumentation includes a slight modification of the usual tank level arrangement using ' differential pressure between an upper and a lower tap. The modification shown in Flgure 7.2.1-3, consists of the use of a sealed reference leg Instead of the conventional open column of water. Refer to 7.2.2.3.4 for l an analysis of this arrangement. N, 7.2.1.1.6A $ $ System The process :- W ; system is described in Reference 11. I 7.2.1.1.7 Solid State Locic Protection System The solld state logic protection system takes binary inputs (voltage /no voltage) from the process and nuclear instrument channels corresponding ; to conditions (normal / abnormal) of plant parameters. The system combines
~
these signals in the required logic combination and generates a trip signal-(no voltage) t6 the undervoltage coils of the reactor trip circuit 1 breakers when the necessary combination of signals occur. The system also provides annunctator, status Itcht and computer input signals which indicate the condition of blstable input signals, partial' trip and full trip functions and the status of the various blocking, permissive and ; actuation functions. In addition the system includes means for sent-automatic testing of the logic circults. A detailed description of this system is given in Reference 3. t
~~7. 2.1.1. 8 Isolation n~~
~~k n t~~
l. In certain appilcations Westinghouse considers-it advantageous to employ control stJg derived from individual protection channels through isolattog...,.. , er, contained in the protection channel, as permitted by IEEE-279.
In all of these cases, analog signals derived from protectieg for located non-protecgeJpctions are obtained through isolatf ofy..,gnels in the ......, protection ..;m. racks. By definition, non-protective functions indication,include and com those signals used for control, remote process
~~! solation; ;.. dew =puter monitoring.~~
qualification tests are described in Referencerd' rd LI 7.2.1.1.9 Enerov Suco1v and Environmental Variations The energy supply for the Reactor Trip System, including.the voltage and frequency variations. is described in Section 7.6. The environmental variations, throughout which the system will perform, are given in i Section 3.11. 7.2-13 0068F/COC4 er + - - -~ --- ,- - ..n., ,-. . . . , , . , , , - --- . . - > . . . , _ _ _ ,---,,,_a ,,,-,,,,--,,-.,,-,,---a_ ,,
4 4I
~~'i SQN-6 7.2.1.1.10 Trio Levels (Setpoints) y i~~
The levels that, when reached, will require trip action are given in the -{ SNP Technical Specifications. (Refer also to Subparagraph 7.1.2.1.9) , 7.2.1.1.11 Seiseie Des 1an l The seismic design considerations for the Reactor Trip System are given in Section 3.10. This design meets the requirements of Crtterion 2 of
-the 1971 General Design Criteria (GDC). j 7.2.1.2 Desion Bases Information The information given below presents'the design bases informatten per Section 3 of IEEE 279-1971 Reference 8. Functional logic diagrams are 6 j presented in Figure 7.2.1-1. 1 7.2.1.2.1 Generatino Station Conditions !
~~The following are the generating station conditions requiring reactor , trip. 1 1~~
1. DN8R approaching 1.30.

2. - Power density (kilowatts per foot) approachthe rated value for i Condition II faults (See Chapter 4 for fuel design 11alts).

3. Reactor Coolant System overpressure creating stresses aporoaching the 11alts specified in Chapter 5.
7.2.1.2.2 Generattna Station Variables i The following are the variables required to be sonttored in order to j provide reactor trips. (See Table 7.2.1-1) j
~~1. Neutron flux ;~~
~~. s ;~~
~~2. Reactor Coolant temperature~~
3
3. Reactor Coolant System pressure (pressurizer pressure)-

4. Pressurtzer water level

5. Reactor Coolant flow

6. Reactor Coolant pump operational status (Bus voltage and frequency,-
~~and breaker position)~~
~~7. 36.- ses. ore ec ';;Mt:r "h:~~
~~"? g Steam generator water level 7.2-14 006BF/COC4 ~~~
~~l 1 SQN-6 1 2 i and stop~~
~~.l. [ Turbine-generator operational status (autostop oil pressure valve position).~~
7.2.1.2.3 Seat 1 ally Denendent Variables The follow 1ng yariable 1s spattally dependent: ' See Paragraph 7.3.I'2 for a discussion l
~~1. Reactor coolant-temperature:~~
~~of thts var 1able spatial dependence. ( 7.2.1.2.4 Listts. Marains and Levels in the SNP~~
~~.The parameter values that will require reactor trip are given Chapter 15.~~
Technical Specifications, and in Chapter 15. Safety Analy are conservative. (Refer also to Subparagraph 7.1.2.1.9) The setpoints for the various functions in the Reactor Trip System have 1 been analyt1cally determined such that the operational l . Reactor Coolant System as a result of any Condition 11 incidentAs suc
~~-(anticipated malfunction).~~
following parameters to: ' 6 I. Minlaum DNBR - 1.30 1. ( 2. Maximum System Pressure - 2750 psia Fuel rod maximust linear power - maximum rated power. l6 3. The accident analyses described in Section 15.2 demonstrate that the-functional requirements as specified for the Reactor. Trip System are adeguate to meet the above considerations, even assuming, for I conservatism, adverse combinations of instrument errors (Refer to Table . , 15.1.3-1). Safety limits associated with the reactor core and Reactor I i Coolant System, plus the Limiting Safety System Setpoints, are presented- ' i [ in the SNP Technical Specifications. i 7.2.1.2.5 Abnormal Events
~~' The malfunctions, accidents or other unusual events which could '~~
physically damage Reactor Trip System components or could cause i ; environmental changes are as follows: *
1. Earthquake (discussed in Chapter 2 and Chapter 3).

2. Fire (See Section 9.5).- (See Section 6.2).

3. Explosion (Hydrogen butidup inside containment).
~~- 4. Missiles (See Sections 3.5 and 10.2.3). ,~~
~~! l 5 Flood (See Chapter 2 and 3). .~~
~~6. Mind and Tornadoes (See Section 3.3).~~
~~b All instrumer.tation, cotitrol and communication lines that~~
~~. F4 N~~
~~e 0068F/COC4 l 7.2-15 4~~
;- . l
~~y 4$ SQN-6~~
~~' 4 flood (DBF) or within a nonflooded structure or are designed for g submerged operation.~~
~~7.2.1.2.6 Minlaum performance Reauirements~~
~~. +~~
~~The performance requirements are as follows: ,~~
~~1. System response times:~~
The reactor trip system response time shall be the time interval from. 6 when the monitored parameter exceeds its trip setpoint at the channel sensor untti loss of stationary gripper cott voltage. Typical maximum a110wable time delays in generating the reactor trip- _ signal: Time (sec)
a. Power range nuclear power (High and low setpoint) 0.5 l

b. Neutron flux rates (positive ,
~~and negative) 0.5 ,~~
~~c. Overtemperature AT (Maximum) ii, oj;..; ;i.7,;prt tt= Of l d. Overpower AT (Maxinum) Unkdte; f.o~~
~~;7.r;p;t it; 0' * :::) 4,4-~~
[
e. Pressurtzer Pressure (low and high) 2.0 lb-

f. Pressurtzer high water level 2.0

g. Low reactor coolant flow 1.0

h. Reactor coolant pump bus under frequency 0.6-

1. Reactor _ coolant pump bus undervoltage , 1.2

3. ' . ;;;; ;;T;7;.%7 7;;i;t;- ??Z I.0
~~'3 M Low-low steam generator water level 2.0 4 k ?! Turbine trip- 1.0~~
f. af Steam generator water level high 12.5 g turbine trip-reactor trip A kees ass isef f $; %g Qf fg,,
7.2-16 0068F/COC4 +
NW , SQN-6 ; I* 2.. Reactor Trip accuracles are,given in Table 7.2.1-4. Protection system ranges: t 3. Ban 21 [!,
a. Power range nuclear power - I to 120% full power

b. Neutron flus rates' +5% to -5% of full-power for rapid 4 .
(positive and negative) I changes in power. t
~~' Overtemperature AT:~~
c.
-T... leg . 530 to 650*F-T,.. leg - 510 to 630*F t... 530 to 630'F Pressurizer pressure 1700 to 2500 pstg s F(A4) - 6 0 +. f g f,, .. .. .._
~~AT setpoint 0 to 150% power 6~~
~~d. Overpower AT (See Overtemperature~~
~~. AT) t t~~
e. Pressurizer Pressure 1700 to 2500 psig .

f. Pressurizer water level Entire cylindrical i ,
~~- portion of-I pressurizer~~
g. Reactor coolant flow 0 to-110% of rated 6 i flow-

h. Reactor coolant pump bus 50 to 65 Hz !
~~under frequency~~
l. Reactor coolant pump bus 0 to 1001 rated undervoltage voltage

j. Steam generator feedwater 0 to 1201 max. calc.
- flow feedwater flow. .
~~k. Steam generator water --~~
+. 6 ft. from level nominal full load. ;
7.2.1.3 Final System Drawines The schematic diagrams for the systems discussed in this section are - ps KJs i contained in report SNP-1, submitted in support of this8.appitcatio - lt fra - titled " Instrumentation Drawings for the Sequoyah Nuclear Plant."- yg q,- L shila i., i 7.2 17 0068F/COC4 t y y, v- -
~~,-, .. ~ ,,- . . , ,...n , , _ .. , .. . . - . . . . . . . . . - . . . . . . - . - . . . ,- -4~~
SQN 7.2.2 Analyses , 7.2.2.1 Failure Mode and Effects Analyses A failure mode and effects analysts of the Reactor Trip System
~~performed.in NCAP-7706, "An Evaluation of Solid State Logic Reactor Protection In ,~~
Anticipated Transtants," (Reference 6). . 7.2.2.2 Evaluation of Deslan Limits While most setpoints used in the Reactor Protection System are fixed, there are variable setpoints, most notably the overtemperature AT and All setpoints in the Resctor Trip System have overpower AT setpoints.been selected on the basis of detailed safety analyses and e design studies. The capability of the Reactor Trip System to prevent loss of integrity of the fuel clad and/or Reactor Coolant Systes pressure boundary during Condition !! and !!! transients is demonstrated in the-T Safety Analysis, Chapter 15. , those setpoints determined from results of the engineering design ' studies. Setpoint listts are presented in the Technical' Specifications. A discussion of-the intent for each of the various reactor trips and the i accident analysts (where appropriate) which uttitres this trip is. It should be noted that the presented in Subparagraph 7.2.1.1.2. selected trip setpoints all provide for is actually required. to allow'for uncertainties and instrument errors. - The design meets the requirements of Criteria 16 and 22 of the 1971-G0C. 7.2.2.2.1 Trio Setootnt Discussion ' It has been noted in Subparagraph 7.2.1.2.4 that below a ONBR of 1.30 there is likely to be significant local fuel clad failure.. The DNBR l existing at any point in the core for a given core design can be l determined as a function of-the core inlet temperature, power output, operating pressure and flow. Consequently, core safety.11mits in terms of a DNBR equal to 1.30 for the- hot channel can be' developed as a and pressure for a specified flow as function of core AT, T.., lines 111ustrated by the solid in Figure 7.2.2-1. Also shown as solid Ifnes in Figure 7.2.2-1 are the lect'of conditions equivalent representing theto 118 percent of power as a function of AT and T.. overpower (kW/ft) listt on the fuel. The dashed Ilnes indicate the and pressure maximum permissible setpoint (AT) as a function of T., for the overtemperature and overpower reactor trip. . Actual setpoint i constants in the equation representing the dashed lines are as given in the SNP Technical Specifications. These values are conservative. to allow for instrument errors. The-design meets the requirements of Criteria 16, 20, 22 and 27 of the 1971 GDC. DNBR is not a directly measurable quantity 1 however, the processSmall isolated . variables that determine DNBR are sensed and evaluated. changes in various process variables may not individually result in i 0068F/C0C4 l 7.2-18 I
~~-~ .- -. _ _ _ _ _ _ _ . _ ___._ __ _.. _ .~~
~~SON-i . .~~
'; violation of a core safety Italt..whereas the combined variations, over sufficient tlas, may cause the overpower or overtemperature safety limit ) - to be esteeded. The design concept of the reactor trip system takes ;
cognizance of this situation by providing reactor trips associated with ; individual process variables in addition to the overpower /overtemperature ! , safety 11mit trips. The process variable trips prevent reactor operation .i whenever a change in the monitored value is such that a core or system 1 safety limit is in danger of being escoeded should operation continue, Basically, the high pressure, low pressure and overpower /overtemperature , jl AT trips provide sufficient protection for slow transients as opposed , to such trips as low flow or high flus which will trip the reactor for' rapid changes in flow or flus, respective)y, that would result in fuel
~~! damage before actuation of the slower responding At trips could be effected.~~
Therefore, the Reactor Trip System has been designed to provide < protection for fuel clad and RCS pressure boundary integrity where: (a) A rapid change in a single variable of factor which wil' quickly result in esteeding a core'or a system safety Itait, and (b) A slow change in ons or more variables will have an integrated effect which will cause safety lletts to be eseeeded. Overall, the Reactor Trip System offers diverse and comprehensive protection against fuel clad failure and/or loss of Reactor Coolant System integrity for Condition !! and III accidents. This is demonstrated by Table 7.2.1-3 which itsts the various , trips of the Reactor Trip System, the corresponding Technical l Spectf tcation on Safety Limits and Safety System Settings and the . appropriate accident discussed in the Safety Analyses in which the trip. t could be utt112ed. I t It should be noted that the Reactor Trip System automatically providet, ! core protection during non-standard operating configuration, t.e., d operation with a loop out of service. Although operating with a-loop out ! of service over an entended time is considered to be an unilkely event, no protection system setpoints need to be reset. This ts because the nominal value of the power (P-8)' interlock setpoint restricts the i power levels such that DN8 ratios less than 1.30 will not be realized during any Condition II transients occurring during this-mode of.- operation. This restricted power level is considerably below the boundary of permisstole values as defined by the core safety limits for ; operation with a loop out of service.. Thus the P-8 Interlock acts essentially as a high nuclear power reactor trip when operating with one j , loop not in service. By first resetting the coefficient setpoints'in the overtemperature oT function to more restrictive values as Itsted in SNP Technical Speelfications, the P-8 setpoint can then be increased to the anstmum value conststent with maintaining DNBR above 1.30 for Condition II transients in the one loop shutdown mode. The resetting of the AT overtemperature trip and P-8 will be carried out under prescribed t
~~' Protection system Interlocks are given in Tables 7.2.1-2 and 7.2.1-4.~~
t 7.2-19 0068F/COC4 . 0 I
D , SQN administrative procedures and only under the direction of authorized ' supervision. Sefore resetting any protection system setpoints, the nuclear plant will At this time the p-8 and AT setpoints will be brought to hot shutdown. Testing (as necessary to assure proper actions have bee be reset. , completed) will be performed before plant power 1s increased.. The nuclear power plant Reactor Trip System design employed by , Nestinghouse was evaluated in detall The with designrespect meets the to common mode failure and is presented in References 6 and 7.
~~-requirements of Criterton 21 of the 1971 GOC.~~
Preoperational testing is performed on Reactor This # Trip System components and testing- . systems to determine equipment readiness for startup. - serves as a very real evaluation of the system design. I Analyses of the results of Condition I !!. I!! and IV Events, including considerations of instrumentation-installed to mitigate their consequences, are presented in Chapter 15. - The instrumentation installed < to mitigate the consequences of load rejection and turbine trip is-given $ in Section 7.7. 7.2.2.2.2 Reactor Coolant Flow Measurement I The elbow taps used on each loop in the primary coolant system are l instrument devices that indicate the status of the reactor coolant flow. The basic function of this device Is to provide information as to whether-i The correlation between flow or not a reduction in flow has occurred. and elbow tap signal is given by the following equation: LP . (y )', t l r AP. v. W., Where AP. Is the pressure differential at the reference flow, The l and 3P 15 the pressure differenttal at the corresponding flow, w. ' full flow reference point is established during initial plant startup. The low flow trip point'is then established by extrapolating along the The expected absolute accuracy of the channel is correlation curve. i within + 10 percent of full flow and field rasults have shown the repeataillity of the trip point to be within g 1 percent.> . . l 7.2.2.2.3 Evaluation of C =aliance to Anolicable Codes and Standards t ' 1 The Reactor Trip System meets the requirements of IEEE-Standard 279, i ' Reference 8, as indicated below.
~~1. -Single Failure'Criterton The protection system is designed to provide redundant (one out of two, two out of three or two out of four) instrumentation channels t~~
~~l 0068F/C0C4 ;; 7.2 20 i~~
}
R { i
j I SQN-6 J for circuits. each protective function and one out of two logic trainThese channel or train will not prevent protective action when Loss'of This meets the requirements of Criterion 22 of the GDC. input power, the most ititely sede of failure, to a channel or logic - train will result in a signal calling for a trip. This meets the-971 GDC. requirements of Criterion 23 of the , l*
!" To prevent the occurrence of common mode failures, such additional I' measures as functional diversity, physical separation, and testing as ,
well as administrative control during design, production. . installatton and operation ars employed, as discussed in References66 and 7. This meets t,h,e reqvfrements of Criterton 21--of the 1971 CDC.
~~2. Quaittyof.Componentsandliodules 1~~
a For a discussion of the quality of the components The quality and modules used meets used in ' the Reactor Trip System, refer to Chapter 17. the requirements of Criterton 1 of the 1971 GDC.- 4
~~3. Equipment Qualification For a discussion of the type of' tests made to verify the performance requirements, refer to Section 3.11.. The test results demonstrate ~~~
~~that the design meets the-requirements of Criterton 22 of the 1971 ' GDC. t i~~
~~4. Independence Each Individual channel is asstuned See to one Figure of7.2.2-2.~~
four channel ' designations, e.g., Channel I. 11. !!I IV. Channel independence is carried throughout the system, extending from. the sensor through to the devices actuating the protective function, physical' separation is used to achieve. separation of redundant , transmitters. Separation of wiring is achieved using separate m wireways, cable trays, conduit runsggntainment .. equipment penetrations is separated by for a' each redu nnel. Redundant-
' n different protE _tton rack sets. Each redundant locating channel 4s energtred from a separate AC power feed. This meets the- i reqbtrements of Criterton 22'of the 1971-GDC. ,
~~4' x~~
~~5. Control and protection System Ihteraction~~
~~- n~~
~~The protection system is" designed to be Independent of the control ,~~
. system. In certain appilcations the control signals and other non-protective functions argged fromThe individual-isolationiampM protective Mcas esesAare channels through isolau g._ , ..;;;.
classified as part of the protection system and are located in the lion-protective functions include those ppcar ;n N protective racks. signals used for control, remote non-!E process g- in l 0068F/COC4' l 7.2-21 l i
N SQN-6
# I computer sonttoring. The isolation are designed such that a short circuit, open circuit, or t e application of 440V AC or 250V- lg I
DC on the isolated output portion of the circuit (i.e., the non-protective side of the circuit) will not affect the input (protective) side of the circuit. du. ~
}(
~~The s1gnals obtained through.the isolation,;: . : 4 are~~
~~-"ysnever i returned to the protective racks. This meets the requirements of l Criterton 24 of the 1971 GDC.~~
~~A detailed discussion of the design and testing of the isolation dwten,s~~
--": : Is given in References 4 I C These reports include the '
results of app'ytng vari as 1 function conditions-on the output portion of the iso'att . The re its show that no- , stgnifIcant disturbance o the isolation input signal , occurred. , - Where failure of a protection system component can cause a process excursion whict) requires protective action, the protection system can < i withstand another, independent failure without loss of protective
~~$6( action.. This meets the requirements of Criterlon 25 of the 1971 GDC.~~
~~g4 j 6. Capability for Testing l The Reactor Trip System is capable of being tested during power~~
operation. Wher6 only parts of the system are tested at any one i time, the testing sequence provides the necessary overlap between the parts to assure complete system operation. .
l The protection system is designed to permit periodic testing of the proces i ;;d :; channel portion of-the Reactor Trip Systes during reactor power operation without inttisting's protective action unless a-trip l condition actually entsts. This is because of the "ANO" logic 1 required for reactor trip. Note ths.t the source and intermediate ! i range high neutron flus trips must be bypassed during testing.- ! The operability of the process sonsters is ascertained by comparison with redundant channels monitoring the saae' process variables or. j i-gg those wtth a f1med known relationship to the' parameter being~ N3
, checked. The sensors can be calibrated during plant shutdewn. - _ ___._. _,n ~E"7;%c.c.;; ::::(;; ": ;;n'..n ' :t tr.:': i::; ::::: ::::::::
~~O ""~~
?!1?^^5N!$$$N Ni!NEiIO '.2?N.I'"YE. . _ . . ". .' .' ."..,,w_ '2 ' ' e N "i _ .'? ! a 1 .' "y r . .y .... .' I .r r '.'. .".i' . .I'.i".'"1 .._ .M"! , . . "". _..v......,....,, ' ','J" ' L 1 ' " '"l ' ?f '"" ""'J"'_I'_'I ' !!; - d '"*""I ' _ '!!' . "M '? ? 'J""* = 9 '
xn:_::1:-- ___..__ ..= 1: n: nn:J: _n:..1 - mmn.; :;. h; u. . ;.i.G..JJ.R;i._7_L. in 3L,G ,i;;i.is;'. K;r;;; ,;;;;;' m;;'" 3_ ' m _
. .i ',*", ,.'". ...U I..'. " L '.",I ' ' .'"" ""r_... ? '.'.T <1 r,.I '. ." ."."."'.," . . . .MI. .' ?".'I...*- ' .".'.&. _. """1"."
I
~
1 7.2-22 0068F/COC4 l' .__._____.__________________u_____ _ . _ . . _ _ _ _ _ _. _ :
[' () i lEEE.Tft s t This is normally achieved by means of two-out-of four (2/4) trip logic for t each of the protective functions except Steam Generator Protection. The ( Steam Generator Low Water Level-protective function relies upon g two-out-of-three (2/3) trip logic and a control system Median Signal Selector (MSS). The use of a control system MSS prevents any protection . system failure from causing a control. system reaction resulting in a need l L for subsequent protective action. l l
~~- t l~~
~~i I I i s. O L i i L e, ese e . d~~
~~? --++w - --~~
m w - ws+ . ,e+-ee e w w.-m,=r1- -%wew ae - .eera. h _,. --g. ,-g-,- y +*w+> awa - e-we--
M INSERT 7 The Process Protection System performs automatic surveillance testing of the digital process protection racks via a portable Man Machine Interface (MMI) test cart. The MMI test cart is connected to a process rack by inserting a connector into the process rack test panel. Using the MMI, the ' Surveillance Test" option is then selected. Following instructions entered through the MI, the rack test processor automatically performs the iollowing operations: '
l. Selection of the individual process channel to be tested.

2. Calibration of the test reference signals and verification of the tester time base.
~~, 3. Placement of the individual channel trip outputs in either~~
~~' Channel Trip" or ' Bypass' (password protected) mode.~~
A. Bypass Mode disables the individual channel bistable trip circuitry which forces the associated logic input relays to i remain in the non tripped state until the " bypass" is removed. 1 B. Channel Trip Mode - Interrupts the individual channel bistable outputs to the logic circuitry to de energize the associated logic input relay (s).
4. Activation of the test injection signal.

5. Performance of Analog to Digital (A/0) converter test, and engineering unit values conversion test.

6. Performance of bistable setpoint tests.

7. Performance of channel time response test.

8. Completion of test cycle and automatically remove " Channel Trips".

g. Verify calibration of the test injection signals.

10. Display of test results on the MI screen.
Interruption of the bistable output to the logic circuitry for any reason !. (test, maintenance purposes, or removed from service of the logic to be actuated and accompanied by a chan)nel trip alarm andcauses channel status light in the control room. Status lights on the process rack test panel indicate when the associated bistables have tripped. Each 1 channel is fully testable via the portable MMI test cart. l-l l l l I
h t
~~' a ;~~
partial trip stars and channel status light actuation in the controlpoints. 44l > roam. Each channel contains those switches. additional necessary to test the channel. See Referenc lif
; information. . , [
The power range channels of the Nuclear Instrumentation The output of Sy l be1ng received by the channel at the tise of tosting.the 61 stables ;
Also, since the power range channel logic is two out of four, bypass of this reactor trip function is not required. l
+
To test a power range channel, a ' TEST-CPERAft' switch is provided to require dellberate operator action and operation of which dfil initiate the 'CHANNtt, TEST".annunctator in the control room. gtstable operetton ts. tested by increasing the test signal level up to its trip setpoint and verifying 61 stab' e relay operation by ' control board annunciator and trip status lights. It should be noted that a valid trip signal would cause the channel under test to trip at a lower actual reactor power level. A reactor trip would occur when a second blstable trips. No provillon has been - made in the channel test circuit for reducing the channel signal ; level below that signal being received from the Nuclear Instrumentation System detector. A Nuclear Instrumentation System channel which can cause a reactor
- trip through one of two protection logic (source or latermediate range) is provided with a bypass function which prevents the ;
initiation of a reactor trip from that particular channel during the ' short perted that it is undergoing test. These bypasses initiate an alare in the control room. For a detailed description of the Nuclear Instrumentation System see ' t Reference 2. l The logic trains of the Reactor Trip System are designed to be capable of complete testing at power. except for those trips listed
~~, la subsection 7.2.3. Annunciation is provided in the control room to Indicate when a train is in test, when a reactor trip is bypassed and~~
~~when a reactor trip breaker is bypassed. Details of the logic system testing are given in Reference 3.~~
The reactor coolant pump breakers cannot be tripped at power without However, causing a plant upset by loss of power to a coolant pump. the reactor coolant pump breaker open trip logic and continuity through the shunt trip coli can be tested at power. Manual trip ~
~~'cannot be tested at power without causing a reactor trip since '~~
operation of either annual trip switch actuates both Train A and Train 8. Note however, that manual trip could also be initiated ; from outside the control room by manually tripping one of the reactor . 7.2-23 0068F/COC4 ,
j [} l i, !
~~$@-6 trip breakers.~~
Initiating safety injection or opening the turbine ; trip breakers cannot be done at power without upsetting j normal pl l operation. However, the Topic for these trips is testable at power. I l I Testing of the logic trains of the Reactor Trip System check of the $$pS input relays and a logic matrix check. The includes .fam ( following sequence is used to test the system: i ' i
~~a. Deck of input relays l~~
During testing of the process instrumentation systes and nuclear . l i instrumentation systen bistables, oath channel % l
~~,s Train 5 to de-energine. A contact This of each card performs relay isboth connected to .~~
~~a Universal p and loute printed conttoring cf reaf functions. The t card.thecontact thatreactor tr'~~
~~creates the reactor trip also causes a status lamp and an$1ther the Train A :~~
annunciator on the control board to operate. i or Train B input relay operation will light the status leap and . l annunciator. l Each train contains a multipleutng test switch, one The A of+ which 8 j (etther train) normally remains in the A + 8 posttion. position alternately allows taformation to be transattted fromDuring 4 i the two trains to the control board.instrumentatibn testing, a ste indicates that input relays inContact de-energised. both trales have beenA flashing le inputs to the logic two tra' as did not both de-energtre. protection systen such as reactor coolant pose hus under { frequency relays operate input relays which are tested by l operating the remote contacts as described above and estag the same type of indications as those provided for blstable input . relays. f Actuation of the $$p$ taput relays provides the overlap between % l the testing of the logic protection systen and the testing of those systems supplying the inputs to the logic protection i ' ( systes. Test Indications are status lamps and annunctators on Inputs to the logic protection systes are i
~~i. the control board.~~
checked one channel at a time, leaving the other c ' out of four channels trip becomes a and out of three i protection system remain la service during this portion of the
test. , -

b. Oeck of Logic MattIses '
Input relays ,. Logic matrises are checked one train at at time. Reactor trips are not operated during this portion of the test. j . j 0068F/COC4 , 7.t-24 .
-..e--.,, a,,,,,, w . , , - - .,,,,---+--,w ,-,,,,,r-.wn,._~n-.w,~c,,m,,m.-~.--.u-,.,-,,,-.-. .--a,,n--~-- - - > - , - - - - - - ~
~~i 0~~
~~.. g~~
~~l l~~
~~from the train being tested are inhibited utth the use of the J' input error inhibtt witch on the seat automatic test panel in~~
~~' the train. Detatis of sont-automatic tester operation are given :~~
in Reference 3. At the comp 14 tion of the logic estrix testt, one i' htstable in each channel of process instrumentation or nuclear instrumentation is tripped to check closure of the input error ~ Inhibit switch contacts. l The logic test scheme uses pulse techniques to check the i coincidence logic. All possible trip and non-trip combinettons ! i: are checked. Pulses from the tester are applied to the inputs of 4 the universal. logic card at the same terminals that connect to the input relay contacts. Thus there is an overlap between the input relay check and the logic matria check. Pulses are fed > i back from the reactor trip breaker undervoltage cell to the tester. The pulses are of such short duratton that the reactor 'j trip breaker undervoltage cott armature cannot respond , mechanically. - j l Test indications that are provided are an annunciator in the control room indicating that reactor trips from the train have been blocked and that the train is 16etng tested, and green and red lamps on the sont-automatic tester to indicate a good or bad logic attrix test. protection capability provided during this portion of the test is from the train not being tested. The general design features and details of the testability of the logic system are described in Reference 3 thus this testing , capability meets the requirements of Criterion 21 of the 1971 GDC. i
7. Testing'of Reactor Trip Breakers Normally, reactor trip breakers 52/RTA and 52/RTB are in service, and In bypass breakers 52/8YA and 52/8YB are utthdrawn (out of service). .
testing the protection logic, pulse techniques are used to avoid ; tripping the reactor trip breakers thereby ettminating the need to bypass them during this testing. The following procedure describes the method used for testing the trip breakers: '
a. With bypass breaker 52/8YA racked out, annually close and tr*1 pit to verify its operation.

b. Rack in and close 52/BYA. Manually trip 52/RTA through a >
~~protection system logic matrix.~~
~~c. Reset 52/RTA.~~
~~Trip and rackout 52/8YA. d.~~
~~e. Repeat above steps to test trip breaker 52/Rild using bypass ;~~
breaker 52/8Y8. , e ( . 7.2-25 .- - 0068F/C0C4 ,
~~- . . _ _ - _ _ - . _ _ _ _ _ _ _ - _ _ _.,_. _ _. _ _ ,. _ _ _ . _ ___.._ _ .. _ _ _ . _ _ _ _ _ . _ . ~ . .~~
N l l i ! . sow ! i Aust11ary contacts of the bypass breakers are connected into the ; alarm system of their respective trains, as described in Reference 3 , such that if eHher train is placed in test while the bypass breaker i of the other * " is closed, both reactor trip breakers and both bypass breaker. .i ' " tomatically trip. ; Auxillary contacts of the bypass breakers are also connected in such a way that if an attempt is made to close the bypass breaker in one - train while the bypass breaker of the other train is already closed, i both bypass breakers will automatically trip. The Train A and Train 8 alarm systems operate separate annunctators j in the control room. The two bypass breakers also. operate an' annunciator in the control room. Sypassing of a protection train with either the bypass breaker or with the test switches will result i in avelble and visual indications. , The complete Reactor Trip System is normally required to be in , I service. However, to permit onitne testing of the various protection , channels or to peralt continued operation ' n the ev:;nt of a subsystes * ' j instrumentation channel failure, SNP Technical Specification, 3/4.3.1 defining the etnimum number of operable channels and the minimum degree of channel redundancy has been formulated. This Technical t Spectftcation alko defines the required restriction to operation in ' the event that t1e channel operability and degree to redundancy requirements canhot be met. l The Reactor Trip Systes is designed in such a way that response time i tests can only be performed dur' ng shutdown. However, the safety analyses uttilre conservative numbers for trip channel response time. The esasured channel response times are compared with those , used in the safety evaluations. On the basis of startup tests conducted on several plants, the actual response flees asesured are ; less than the times used in the safety analyses. Refer to Table ; l 15.1.3-1. , ,
8. Sypasses i ip Nher's operating requirements necessitate automatic or manual bypass J of a protective function, the design is such that the bypass is ,
removed autcastically whenever peratssive conditions are not met. Devices used to achieve automatic removal of the bypass of a - protective function are considered part of the protective systes and are designed in accordance with the criteria of this section. Table i 7.2.1-2 'dentiftes and discusses bypasses associated with the Reactor Trip System. Indication is provided in the control room if some part of the systen has been administratively bypassed or taken out of . service. 5 fincas5 telechbs Jffh> j1, )n.tyed $s. if aA JA. d ,d k (a, /ge) l^
~~a.5>~~
~~\ &oQ 4 ro v U s. $hy ae pidic 1~~
ks+ e,f a be rhJA+ pr yc% c Ad. 7.2-26 0068F/C0Ca l
b ! SQN s- .
g. Multiple Setpotats y gf For sonttorino neutron flus 4 aultiple set re used. When a j more restrict 1ve trip setting becomes necessary to provide adequate ,
protection for a particular mode of operation or set of operating ! f conditions. the protective system circuits are designed to provide _, ~ positive means or adelnistrative control to assure that the more restrictive trip setpoint is used. The devices used to prevent ! leproper use of less restrictive trip settings are considered part of , , the protective system and are designed in accordance with the l . criteria of this section. ! i .: !
~~10. Completion of protective Action l~~
The protection system is so designed that, once inttlated, a ! j protective action goes to completion. Retura to normal operation l requires action by the operator. ll l 11. Manual Initiation ' I switches are provided on the Control Soard for manual initiation of ,
protective action. Fallure in the automatic systes does not prevent

the annual actuation of the protective functions. Manual actuation ,
' relles on the operation of a etnieus of equipment t i 12. Access . ,i - l The design provides for administrative control of access to all set >
~~point adjustments, module calibration adjustments, test points, and l~~
the means for annually bypassing channels or protective functions. \ Fordotatisrefer.toReference'] . i 13. Information Read Out ~ The protective system provides the operator with complete information j pertinent to systes status and safety. All transmitted signals (flow, pressure temperature, etc.) which can cause a reactor trip is i either indicated or recorded for every channel, including all neutron flus power range currents (top detector, bottom detector, algebraic difference and average of botton and top detector currents). ( ' Any reactor trip will actuate an alarm and an annunctator. Such
~~protective actions are indicated and identified down to the channel~~
! level. ! Alaras and annunctators are also used to alert the operator of j deviations from normal operating conditions so that he any take appropriate corrective action to avoid a reactor trip. Actuation of l . any rod stop or trip of any reactor trip channel will actuate an - altra.
! l ' 7.2-27 0068F/C0C4 i j c , .....-.-___.__.--_._....__.~,__,,.-._..-,.___...._..__.....,.....-,....._.',..,_.... i
~~h SQN~~
14. Identification The identtf tcation described in Section 7.1 provides lamediate and ;
unambiguous identification of the protection equipment. : 7.2.2.3 Soetific control and Protection Internettons ) 7.2.2.3.1 lieutron Flus Four power range neutron fles channels are provided for overpower l prateet1on. An add 1tIonal auctioneered hIth signal is derived by ; auctioneering of the four channels for automatic rod control. If any ' ekannel falls in such a way as to produce a low output; that channel is : incapable of overpower protection Dut will not cause control red movement because of the auctioneer. Two out of four overpower trip logic will ! ensure an overpower trip if needed even with an independent failure in ; l another channel. j
^
l In addition, channel deviation signals in the control s tea will give an alare if any neutron flus channel deviates stentftcanti from any of.the other channels. Also, the control system wl1I respond Iy to rapid , changes in indicated neutron flus; slow changes or Wrtfts are compensated by the temperature control signals. Finally, an overpower signal from any nuclear channel will block automatic rod withdrawal. The setpoint for this rod stop is below the reactor trip setpoint. ' 7.2.2.3.2 Coolant Teneerature , The accuracy of the resistance temperature detector O -- ' :; 4emposetens measurements is demonstrated during plant startup tests by
'::; resistance comparingdetectors ,paperajug, temperature with onemeasurement.from anothu "g all[M:::gg j 7 7 -- 2 ni stm tury rzeswu g -" r- . '/ r- 'rt swg wiew owww swy F'F'"y v' m --=" '- "rrn r "-" ~~-' -
r' The comparisons are done with the Reactor Trip Systes in an isothermal condt f,lon. The Itnearity of the AT asasurements obtained from the hot i
, leg and cold leg erpWWUuGWup resistance temperature detectors as a function of plant power is also checked during plant startup tests. The absolute value of AT versus plant power is not taportant as far as reactor protection is concerned. Reactor Trip Systes setpoints are based ,
upon percentages of the Indicated AT at nostnal full power rather than on absolute values of AT. For this reason, the linearity of the AT' , signals as a function of power is of taportance rather than the absolute values of the AT. As part of the plant startup tests, the loop . res1 stance temperature detottor si nals wt11 be compared vtth the core ! entt thermocouple signals. Note a so that reactor temperature control is 6 i based upon slynals dert g isolation by solatio., y d g: .. protection
~~such that system channels after~~
_ , no feedback effect can i perturb the protection, channels. State control is based on the highest ' l average temperatu:n of the loops, the control rods are always moved based F l
l 7.2-28 0068F/COC4 l
~~-----.v-.~..o__, , , , _ _ , _ _ , .,,m,,_,.,. , o w e m, m , ,.m.-..,, _.,,,..._e.,~~
p, ,.w.m7,,,-,,,,.-y,m_,wyvw-,~.,-p,,,,,#,-e
l
~~$ON N~~
~~k' . Upon the most pesslaistic temperature seasurement with respect to margins~~
\
j to DNS. A spurious low average temperature sensurement from any loop temperature control channel w'11 cause no control action. A spurious high average temperature anaturement v11) cause rod insertion (safe j direction). s ._ . .,_. . -._ _._ .. _, ., . . . . .. .. ,__ .._ __.. __
~~": n ' _ _ ' :'" M :_' ': ": !" :',~ ',':: : ' _ ' ":" : : T":" L". _'" " ' ""~~
~~I" 7; L:' 7 l l l- L !'i n _""_:Z' - i'a Imi J; i"L . " a .~~
~i = :- ~~=~--

i. :__ ' _:::u '. :'.;. .:... n : ' .,-_: :: . z ..
~~73 7_ _ - . . _ _ _ . - . __,. l~~
~~u: ::":"::" '_T_r' . r:=" '- . ::" .::=LL '1' : :=. ::- -.~~
~~1: Z: z nr c:-' : :n ?"...m "._r z '... '.".-.'. ..' .- . .'1 :.. . . . . . ' . . . " , , . . -. _. 3...}}~~

ML20006B389: Difference between revisions

Revision as of 15:46, 17 February 2020

Text

7.1 INTRODUCTION

Navigation menu

ML20006B389: Difference between revisions

Revision as of 15:46, 17 February 2020

Text

7.1 INTRODUCTION

Navigation menu

Search