ML20197B884

Proposed Tech Spec Changes Re Max Extended Operating Domain

ML20197B884
Person / Time
Site:	Grand Gulf
Issue date:	05/02/1986
From:	MISSISSIPPI POWER & LIGHT CO.
To:
Shared Package
ML20197B873	List: AECM-86-0129, Application for Amend to License NPF-29,changing Tech Specs Re Max Extended Operating Domain.Amend Required to Support Cycle 1 Operations & Cycle 2 Reload Licensing Effort.Fee Paid ML20197B884
References
TAC-61357, NUDOCS 8605130175
Download: ML20197B884 (120)

Text

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/3 /8' l

INDEX  !

l LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE.

3/4.0 APPLICABILITY................................................... 3/4 0-1  ;

3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.............................................. 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES......................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability...................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times.................... 3/4 1-6 Control Rod Scram Accumulators............................... 3/4 1-8 Control Rod Drive Coupling................................... 3/4 1-10 Control Rod Position Indication.............................. 3/4 1-12 Control Rod Drive Housing Support............................ 3/4 1-14 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Control Rod Withdrawal....................................... 3/4 1-15 Rod Pattern Control System................................... 3/4 1-16 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................ 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 3/4 2-1 DELETED ec 3/4.2.2 ^"""

SET"0!NTS............................................... 3/4 2-3 3/4.2.3 MINIMUM CRITICAL POWER RATI0................................. 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE.................................. 3/4 2-7 h

P GRAND GULF-UNIT 1 iv hmigbh\iXT b.

A /9 INDEX BASES SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN........ ........................... B 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES................................ B 3/4 1-1 3/4.1.3 CONTROL R0DS........................................ B 3/4 1-2 3/4.1.4 CONTROL ROD PROGRAM CONTR0LS........................ B 3/4 1-3 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM....................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.......... B 3/4 2-1 DELETED 3/4.2.2 APRM 0:!PO!NTS...................................... B 3/4 2-2 3/4.2.3 MINIMUM CRITICAL POWER RATI0........................ B 3/4 2-4 3/4.2.4 LINEAR HEAT GENERATION RATE......................... B 3/4 2-7 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION........... B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION................. B 3/4 3-1 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..................................... B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION..................................... B 3/4 3-2 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION........................... B 3/4 3-3 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION................... B 3/4 3-3 l

GRAND GULF-UNIT 1 xii i hmE@ MENT b-

TABLE 2.2.1-1 i

REACTOR PROTECil0N SYSTEM INSTRUMENTATION SETPo!1TS ALLOWA8tE c3 FUNCTIONAL UNIT TRIP SETPOINT VALUES c)

% 1. Intermediate Range Monitor, Neutron Flux-High i 120/125 divisions < 122/125 divisions p of full scale of full scale c: 2. Average Power Range Monitor:

z Z a. Neutron Flux-High, Setdown S 15% of RATED 1 20% of RATED g THERMAL POWER THERMAL, POWER

b. _ Flow Blased $1mulated Thermal Power-High 4; riv. G h ed -< 0.66 W+48%, with -<-& ?? .;;",5 , wTTh 5d""bl INSERT k a <-  ;; a maximum of ** Me Kl%

- f) High Flow Clamped < 111.ua oi M M O < 113.0E of RATED s,. inwel k<

TdRtf 2.2.1-1 THERMAL POWER i nt _ m EP_* g.a g

c. Neutron Flux-High 1 118% of RATED < 120E of RATED eL-f 5 'd THERMAL POWER THERMAL POWER %Wrequot

d. Inoperative NA NA

3. Reactor Vessel Steam Dome Pressure - High 5 1064.7 psig i 1079.7 psig to

• 4. Reactor Vossal Water Level - Low, Level 3 > 11.4 inches above -> 10.8 inches above

• - Instrument zero* Instrument zero"

5. Reactor Vessel Water Level-High, Level 8 5 53.5 inches above 1 54.1 inches above instrument zero* instrument zero" l 6. Main Steam Line Isolation Valve - Closure 3 6% closed 5 7% closed

, 7. Main Steam Line Radiation - High 1 3.0 x full power $ 3.6 x full power

background background

8. Drywell Pressure - High < 1.23 psig < 1,43 psig

9. . Scram Discharge Volume Water Level - High ST cf f;;; ;;;?; 0 % ;f f;? ,;;;.

l

10. ' Turbine Stop Valve - Closure 1 40 psig** 1 37 psig

11. Turbine Control Valve Fast Closure, 3 Trip 011 Pressure - Low 1 44.3 ps10** 1 42 psig a

'[ 12. Reactor Mode Switch Shutdown Position NA NA NA 3 13. Manual Scram NA

. rt

'h *See Bases Figure B 3/4 3-1.

• • Initial setpoint. Final setpoint to be determined during startup test program. Any regulied thange to i

this setpoint shall be submitted to the Commission within 90 days of test completion.

a. Transmitter / Trip Unit g 60% of full scale < 63% of full scale

b. Float Switch 5 64" }65" b

O

1 INSERT for Table 2.2.1-1, item 2b (MEOD) 2.b 1) During two recirculation loop operation i a) Flow Biased 3 0.66 W+64%, with 60.66 W+67%, with a r.aximum of a maximum of b) High Flow Clamped $111.0% of RATED $113.0% of RATED THERMAL POWER THERMAL POWER

2) During single recirculation loop operation:

a) Flow Biased f0.66 W+40% g0.66 W+43%

b) High Flow Clamped Not Required Not Required OPERABLE OPERABLE I

l i

i l

b g

J12 MISC 86030402 - 2 I

baQaQ

\

LIMITING SAffTY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

Average Power Range Monitor (Continued) amount, the rate of power rise is very slow. Generally the heat flux is in near equilibrium with the fission rate. In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit. The 15% neutron

- flux trip remains active until the mode switch is placed in the Run position.

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the sys-tem and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High 118% setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer

. associated with the fuel. For the Flow Biased Simulated Thermal Power-High I setpoint, a time constant of 6 1 1 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1. :Cvsagt'F The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating mar in that reduces the possibility of unneces-i sary shutdown. trip setpoint'aus specified formula in S ntain these margins n TP.

3. Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pres-sure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip set-ting is slightly higher than the opecating pressure to permit normal operation without spurious trips. The setting provides for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure seasurement compared to the highest pressure that occurs in the system during a transient. This trip setpoint is effective at low power / flow conditions when the turbine stop valve closure trip is bypassed. For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

B 2-7 GRAND GULF-UNIT 1 kmeQ

Q 7 2.Q INSERT F ,

flow In these,bi,ased equations, the variable v. is the loop recirculation flow as a percentage of the total loop recirculation flow which produces a rated core flow of 112.5 million Ibs/hr.

e

R .23 '

as mul%phad by iht smaller c.( tct(etr %e he.w- d epend en4 MOL uc,R (acier (MRTr AC 4) o( Tu3 ure 3. 2.1 - 2 , or h yewer dqe.ndeed

3. 2. s -3 MRPL4CR (acle.< (MAM Ach a( ~4%ure M.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLNGRs) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE s- '2  :.2. .shall not 3.g

-.S exceed e the limits shown in Figure 3. 2.1-1/ ndor.v . i- ._ z- a APPLICABILITY

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than er equal to 25% of RATED THERMAL POWER.

ACTION: p,,4 sambl en Mar 31,tB With an APLHGR exceeding th limits of Ti;.7; 2.2.1 1, initiate corrective action within 15 minutes and restore APLHGR to within the required limits within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREMENTS

{c. ppt.c.We s.tdite.1 en & 3i,196 4.2.1 All APLHGRs shall be verified to be equal to or less than the" limits :

det;,;i.;;d f.;- Ti ..; 0.0.1 L

a. At least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />,

b. Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and 4

c. Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

s

d. The provisions of Specification 4.0.4 are not applicable.

]

GRAND GULF-UNIT 1 3/4 2-1 AMKuomsvrNo. l

AM

/l

~

e a a en s 'g, s I , ,

l it Ig

- r:1 -

8 i

5 '

W l  :  :  :

. tE Ia li .l ...

8 Ig58 k

h.t !

- i n

="a w

I

-  ;< a  : -

! $ lg l

i!a g  ! =1-spg i -

i. ..  :. ,

tu" wa

. ~.

t, y y;gk-a

. 3 4

- t I,ve

• x E 3

1

. I "I a e a '. b

= 2 e3 *

(43/M) 31VW N011VW3N)0 AY3M WY3 Nil WVNVid 30VW3Av MontxvM ERNO GULF - INIT 1 3A 2-2A hootnr tb.

2 -

. A DS 1.1 l

//

/ 7-CLAMPED DURING _

/r/

0.9 ONE LOOP OPERATION I

E - FOR MAX FLOW = 102.5%

g 0.8

$ 2

/ FOR MAX FLOW e 107.0%

g E

0.7 f)

N i

O.6 40 60 80 100 120 O 20 CORE FLOW (% RATED), F FIGURE 3.2.1-2 MAPFACf GRAND GULF UNIT ONE 3/4 2-Ea

A A6 1.1 l.O

/

O.9 "a.

, o FOR P>70% ;

I O.8 - r DURING ONE LOOP OPERATION =

q FOR 25%iP140%;

CDRE FLOW FS50%

P FOR 40%<PS 100%;

4 N\ / ALL CORE FLOWS 0.7 s=

O.6

/

/

FOR 25%$ P$40%;

j O.5 CORE FLOW F > 50%

O.4 0 20 40 60 80 00 0 12 0 CORE THERMAL POWER (% RATED) P j

l l

FIGURE 3.2.1-3 MAPFACp l

1 l

GR AND GULF UNIT ONE 3/4 2-Eb AMENDMENT No. -

l _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . . _ _ _ _ - _ _ __ _ _ _ _ _ _ _ _ _ _ _ ___ _

M.27 b bw. ' ado rai.re wish % W

• T)E LET E D Tr., ew +p4.a now.w. V.Le _g o-POWER DISTRIBUTION LIMITS SS (o + Ao%)T SS *4v7.) T '~

,, s (au, w , 545)T Ses g(o.m w, T  ;;f 3/4.2.2 APRM SETPOINTS , '

i

! LIMITING CONDITION FOR OPERATION W

.2.2 The APRM flow.. biased simulated thermal power-hi h scram trip setpoint

( and flow biased neutron flux-upscale control rod b ock trip setpoint (S B) c sha1beestab]ishedaccordingtothefollowingrelationships: # Ud oEN Trip S ANowableValue 1

S < (0.66W + 51%)T 5 l 5 < (0.66W + 48%)T j S

S RB $ (0.66W + 4 5)T RB $ (0.66W + 45%)T 1 m #

where: 5 dS RB are in Percent of RATED THERMAL POWER.

W =L p recirculation flow as a percentage of th loop recirculation fl which produces a rated core flow of 11 million Ibs/hr.

T = Lowest alue of the ratio of FRACTION OF ATED THERMAL POWER LIMITING POWER DENSITY (FRTP) 'vided by the MAXIMUM FRACTION (MFLPD). is applied only if less t n or equal to 1.0.

APPLICABILITY: OPERATIONAL ONDITION 1, when TH L POWER is greater than or l

equal to 2 n of RATED THERMAL OWER.

ACTION:

hern power-high scram trip setpoint and/

With the APRM flow biased simulate l

or the flow biased neutron flux-upse e ontrol rod block trip setpoint less l

conservative than the value shown in allowable value-column for 5 or S etionwithin15minutesandrestNe, as above determined, initiate correc ve S and/or Se a to within the require imit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or reduce THERMAL j

POWER to 1Hs than 25% of RATED T . RMAL PO within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

j SURVEILLANCE REQUIREMENTS - x r each class of fuel s all be determined, the value 4.2.2 The FRTP AND MFLPD of T calculated, and the ost recent actual APRM fl biased simulated thermal power-high scram and f1 biased neutron flux-upscale ontrol rod block trip setpoints verified to e within the above limits or ad sted, as required:

a. At least ce per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, Within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after completion of a THERMAL PO increase of at b.

less 5% of RATED THERMAL POWER, and

c. In tally and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the rea or is operating th MFLPD greater than or equal to FRTP.

d. The provisions of Specification 4.0.4 are not applicable.

• W h MFLPD greater than the FRTP during power ascension up to 90% of RAT e ERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may adjusted such that APRM readings are greater than or equal to 100% times MFL D provided that the adjusted APRM reading does not exceed 100% of RATED THERMAL POWER, and a notice of adjustment is posted on the reactor control panel.

GRAND GULF-UNIT 1 3/4 2-3 MENDMur No.

A a:\!

SEPL6ct. wit 4 Ntw F t G,u RE 3.2.3- 1 1

los

/ l l ,, N /

\ '

/

I

\\

\\ [

Y\ /

,~,

W

,~.

^\

/ NX

\

X i.:

bted MC.PR of g Limit

[ \

\

I,

[ '

,, /

.0 20 40 60 to 10 0

\ 120 Core Flow, % of Roted Core Flow MCPR, nouac 3.2.3-1

__ 34 ,, mm

I Aa9

'Rt e uw e t. wira at w r , wet. s.2. s - 2.

l. ~

\ /

N /

'N N /

y x /

E4

.I N\ Y ,

E \> '

b

~

/ NN 1.2 g

/ \\  %

1.1

/ \ T j

l' O

/

20 40 50 80 10 0

\ 20 Thermot Power, % of Roted Thermal Power MCPR p

FIGURE 3.2.3-2 GRAND GULF-UNIT 1 3/4 2-6 gnygatm No,

k30 W6A  :

~

I.7

l. 6 15

\\ FOR MAX FLOW e 107.0%

( ,

\

FOR MAX FLOW = 102.5%

o;- '> s a

\\

l I2

\\ \N 1.1 RATED NCPR OPERATlNG LIMITal.18 1 1 1 1 1 0 20 40 60 00 10 0 12 0 CORE FLOW (% RATED), F FIGURE 3. 2.5-1 MCPR, GRAND GULF UNIT ONE 5/42-5 .

A3/

gM i

i l i i i THERMAL POWER 25% s.Pi40%

2.2 CORE FLOW > 50%

2.O ,

eTHERNA POWER 25% $ P$40%

CORE FLOW 5 50%

b i

l l.8 i

I. 6 THERNAL POWER 4 No < P i 70%

' ALL CORE FLOWS g

g I

i a N I I.4 THERNAL POWER P>70%

l.L CORE FLOWS N

f.2  %

1. 0 0 20 40 60 80 100 120

! CORE THERMAL POWER (% RATED) P l

l RGURE 3.2.3-2 NCPR p RELR-D

k3A TABLE 3.3.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUMENTATION

~

TABLE NOTATIONS (a) A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance without placing the trip system in the tripped condi-tion provided at least one OPERABLE channel in the same trip system is monitoring that parameter.

(b) The " shorting links" shall be removed from the RPS circuitry prior to and

~

during the time any control rod is withdrawn

• per Specification 3.9.2 and shutdown margin demonstrations performed per Specification 3.10.3.

(c) An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

(d) This function is not required to be OPERABLE when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

(e) This function shall be automatically bypassed when the reactor mode switch is not in the Run position.

(f) This function is not required to be OPERABLE when DRYWELL INTEGRITY is not required.

(g) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. qnuf-/sqc bs/ow thE appacpaink (h)

This function shall be automatically bypassed V when turbine first stage [ ea pressure is less than: 30%** of the value of turbine fint . stage-pressure ~ ~

in psia, at valves wHe open..(YWO)-steam TiowFequiv~alent to THERMAL POWER

_less .than -40%-of-1tATED THERMAL POWER.

N_ -

.l cOS 2L- h *4 N 14 valg e(4urbme f est sl y pren ure cd [

y.as aa< qu (t wo') sk- fic- u.k w c e d q - % l w (ua Ar h.p alu r< c4 cy a r tw er 42F F I

\c 9 sd -

@$22.5'/P e( the solue c4 4 urb Ost s4Se pr essur e cet VWOsl% (\cw u,Lan operokmj m N tcsId bd hpt u l ur( igk a c .' D c'F wel 42&h

" 31.Itauired for con _ttol rods lemovedJer Sgecification 3.9.10.1 or 3.9.10.2._

n iai .etp 9 + Final setpoint to be determined durin -test 770 gram.

i ed to the Commission *-

Any required change to ThTrset ~

within 10 thys . mp etion.

wcRoiu T:c.wrJ: te ss 1Lw @% 4 U A T c T> wcemet Poec.

se+po:st GRAND GULF-UNIT 1 3/4 3-5

[\muidmwr de_ ,

O 3.2CL.

INSERT G pressure setpoint of; e

1,2,5 U)

c. TrcIsaittcr/ Trip tlnit S M R IRI

~

b. Float Switch NA H R 1,2,5 I3}

cs TAOLE 4.3.1.1-1 (Continued)

REACTOR PROTECTION SYSTEM INSTRUNENTAfl0N SURVEILLMICf REQUIREMENTS a

CiANNEL OPERATIONAL E CHANNEL FUNCl10W. CHANNEL CON 0lil0NS FOR 1AllCH

! t;; CAllBRAil0N SURYEllLANCE R[QUIR(O ,

,

• FUNCTIONAL UNIT CHECK _ IE51 Scree Olscharge Volume ifater %Wibl on 9.

Level - High #.r - -M- -A M - 1, 2, [" y, g g 7

e.a

10. Turbine Stop Valve - Closure 5 M R I9I 1 l 11. TerMne Control Valve Fast

Closure Valve Trip System 011 I9) 4 Pressure - Low 5 M R 1 l 12. Reacter Mode $ witch 1,2,3,4,5 Shutdown Positicn MA R NA 1

13. Manual Scran MA M MA 1,2,3,4,5 3~

(a) feestron detectors may be escluded from CHANNEL Call 8RAfl04.

T (b) The IRM and SRM channels shall be determined to overlap for at least 1/2 decade during each CD startup after entering OPERATIONAL CON 0lil0N 2 and the IRM and APRM channels shall t,e deter-eined to overlap for at least 1/2 decade during each controlled shutdown, if not performed within the previous 7 days.

(c) [0ELETEO]

! (d) This calibration shall consist of the adjustment of the APIIM channel to confers to the power valves calculated by a heat balance during OPERAil0NAL CON 0lfl0N I when THERMAL POWER > 25% of RAT [0 g'

THERMAL POWER. Adiust the APRM channel if the_=*absolute difference in compliance is creater with (San 21 of Specification RATED 3.2.2 2

THERMAL POWER, _I Any Arm'CMannri v.; dj='--'

} l shall not be ' ncluded in determinina the absolute dif ference.

> (e) This Calibration shall consist of the adjustment of the APRM flow biased channel to conform to a callbrated flow signal.

, 3, (f) The LPWMs shall be calibrated at least once per 1000 MWD /T using the ilP system, j 3 (g) Calibrate trip unit at least once per 31 days.

I 51- (h) Verify measured drive flow to be less than or equal to estabilshed drive flow at the esisting flow con-3 . trol valve position.

80 (1) This calibration shall consist of verifying the 6 t I second simulated thermal power time constant.

(j) Blot appilcable when the reactor pressure vessel head is unbolted or removed per Specification 3.lft.l.

$ (k) Not appilcable when DRYWitt INT [GR11Y is not required.

D (1) Appilceble with any control rod withdrawn. Not appilcable to control rods removed per Systletra-D tion 3.9.10.1 or 3.9.10.2.

b ce lh

R34 INSTRUMENTATION TABLE 3.3.4.2-1 END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION MINIMUMOPERABLECHANg5PER TRIP FUNCTION TRIP SYSTEM

1. Turbine Stop Valve - Closure 2(b)

2. Turbine Control Valve - Fast Closure 2(b)

& v>:m m Oct W TnE. APM @ #

1

'wrw-SeTPO. (T" 7 (a) A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance provided that the other trip system is OPERABLE.

(b) This function shall be automatically bypassed when turbine first stage pressure is less thant30%* ef the value of turbin; fir;t t;;: pr ;;ure,

-in p;ia, et valvc; .cid ; pen ("WO) ;t::: ficw, ;;uivalent te THER"M P05'ER le;; then 40% of RATED T"EC"a NUER.

4 33 b Jb V EIbg I ,7 b V w Dv b bb w #* w w *w w Any required ch;n;; to thi; ;cip; int ;holl 5: ;ut;itted to the Cemi;;ien withi- 90 day; Of te:t comp! ticr cye. MC2.T' 3/43-46 2

1

- GRAND GULF-UNIT 1 3/4 3-43 NfESMWG NO.

A ss D5ERT 3/4 3-43 stage pressure at valves a)4 wide 26.91 of the value of turbine first open(WO) steam flow when operating wigh rated feedwater temperature of greater than or equal to 420 T; or b)f22.5% of the value of turbine first stage pressure at Wo steamflowghenoperagingwithratedfeedvatertemperature -

between 370 F and 420 F. Jefpoln +

These represent allowabl s of turbine first stage pressure equivalent to THER M. POhTR less than 40*. of RATED THER.M F0kTh.

a

TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS h

5 TRIP SETPOINT ALLOWABLE VALUE TRIP FUNCTION g

g 1. ROD PATTERN CONTROL SYSTEM 20+15,-0%ofRATEDTHERMfLPOWER

a. Low Power Setpoint 20 + 15 -0% of RATED THERMAL E POWER

b. High Power Setpoint 1 70% of RATED THERMAL POWER 170% of RATED TilERMAL POWER

2. APRM R EPt.act wit 4 let RT (or TA%c 3.3.6-2 sa,;g 3,, ng, }

~

a. Flow Biased Neutron Flux- \

Upscale - (0. 00 t! "2Tr)f * ' (0. 55 L' ' iS%)T *  %

RA w , m ,4

b. Inoperative RA

) c. Downscale 1 4% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER g w,%g a s g;,

d. Neutron Flux - Upscale 1 14% of RATED THERMAL POWER g ,

Startup i 12% of RATED THERMAL POWER 1

w 3. SOURCE RANGE MONITORS NA A a. Detector not full in NA

< 1.5 x 105 cp3

b. Upscale < 1 x 105 cps Y NA

$ c. Inoperative NA

d. Downscale 1 0.7 cps 1 0.5 cps

4. INTERMEDIATE RANGE MONITORS NA

a. Detector not full in NA

< 110/125 of full scale

b. Upscale < 108/125 of full scale NA

c. Inoperative NA

d. Downscale 1 5/125 of full scale 1 3/125 of full scale

5. SCRAM DISCHARGE VOLUME

a. Water level-High 1 32 inches 5 33.5 inches 4 p b w'

4

$ 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW

. ~

a. Upscale i 108% of rated flow 1 111% of rated flow

]

7. REACTOR MODE SWITCH SHUTDOWN NA k P05ITION NA

' = tie, vi M circulation loop flow

-t =

? *The Average Power Range moniiv. r:d 9~-k function is varied ac tk '"wl4]M FRACTION of LIMITING POWER DENSITY (W) and theThe ratio of FRACTI K_nf__ RATED T C Ji ruwtR ior p setting of this function must be main

_(TJ Med .

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

i I INSERT for Table 3.3.6-2, item 2a 2.a Flow Biased Neutron Flux - Upscale

1) During two recirculation loop operation i a) Flow Blased 30.66 W+58%, with & 0.66 W+61%, with j a maximum of a maximum of b) High Flow Clamped i108.0% of RATED 5110.0% of RATED THERMAL POWER THERMAL POWER l

I 2) During single recirculation loop operation:

a) Flow Biased 40.66 W+34% 40.66 W+37%

b) High Flow Clamped Not required Not required OPERABI,E OPERABLE l

i k

I J12 MISC 86030402 - I g.

t 9

A 38 INSTRUMENT ATION TABLE 4.3.6-1 (Continued)

CONTROL R0D BLOCK INSTRUMENTATION SURVEILLANCE REQUIREMEN NOTES:

a. Neutron detectors may be excluded from CHANNEL CALIBRATION.

b. Within 7 days prior to startup.

c. Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to control rod movement and as each power range above the RPCS low power setpoint is entered for the first time during any 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period during power increase or decrease. '

d. At least once per 31 days while operation continues within a given power range above the RPCS low power setpoint.

c. [ Deleted]

f.

This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to__

Adjust the APRM channel if the absolute 25% of RATED THERMAL POWER.

difference is greater than 2% of RATED THERMAL POWER. fan " % Cna5Hel gain ei.Uuite:mt -maNn7omfLNe n % specnication 3.2.2 shall not beg"d=% da;enng the aDsoiuir dif f:e"*

g. This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

h. This calibration shall consist of verifying the trip setpoint only.

• With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.

d 3/4 3-57 GRAND GULF-UN11 1 Annbow sT N e. -

P. 39 ,

3/4.4 REACTOR COOLANT SYSTEM 3/4.4.1 RECIRCULATION SYSTEM RECfRCULATION LOOPS -

LIMITING CONDITION FOR OPERATION

/

3. 1.1 Two reactor coolant system recirculation loops shall be in operatio APPLI ILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With one eactor coolant system recirculation loop not in o ration, immediatel initiate an orderly reduction of THERMAL POWE to less than +

of the 100% Rod Line as specified in Fi re B 3/4 2.3-1, g or equal to and be in at I st HOT SHUTDOWN within the next 12 ho s.

b. With no reactor to ' ant system recirculation loop in operation, )

immediately initiate n orderly reduction of TH L POWER to less .;

than or equal to 80% o the 100% Rod Line as ecified in figure B 3/4 2.3-1, and initiat measures to place he unit in at least y STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> an in HOT SHUT 00 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. e a

• 7 SURVEILLANCE REQUIREMENTS

[- 5 m;

- x Ei d 4.4.1.1.1 Both reactor coolant stem recirc ation loops shall be verified E to be in operation at least o per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. y 4.4.1.1.2 Each reactor co ant system recirculatio loop flow control valve shall be demonstrated OP ABLE at least once per 18 m ths by: 4ye

a. Verifying at the control valve fails "as is" n loss of hydraulic I$

j J, pressure t the hydraulic unit, and

b. Veri ing that the average rate of control valve mov ent is:

Less than or equal to 11% of stroke per second open g, and

2. Less than or equal to 11% of stroke per second closing.

/ "See Special Test Exception 3.10.4.

GRAND GULF-UNIT 1 3/4 4-1 ggggpffgr [p._

A 40 Insert 3/4.4-1 page 1 of 2 3.4.1.1 The_ reactor coolant recirculation system shall be in operation and not in Region IV as specified in Figure 3.4.1.1-1 with either:

a. Two recirculation loops operating with limits and setpoints per Specifications 2.1.2, 2.2.1, 3.2.1, 3.2.2 and 3.3.6, or

b. A eingle recirculation loop operating with:

1) a volumetric loop flow rate less than 44,600 gpm, and

2) the loop recirculation flow control in the manual mode, and

3) limits and setpoints per Specifications 2.1.2, 2.2.1, 3.2.1, 0.2.2, and 3.3.6.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*

ACTION:

a. During single loop operation, with the volumetric loop flow rate greater than the above limit, immediately initiate corrective action to reduce flow to within the above limit within 30 minutes.

b. During single loop operation, with the loop flow control not in the manual mode, place it in the manual mode within 15 minutes.

c. With no reactor coolant system recirculation loops in operation, ic=ediately initiate an orderly reduction of THERMAL POWER to within Region III as specified in Figure 3.4.1.1-1, and initiate measures to ~

place the unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN within che next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

d. During singA- loop operation, with temperature differences exceeding the limits of l'IRVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the THERMAL PO*.'ER or cecirevlation loop flow increase.

e. With operation in Region IV as specified in Figure 3.4.1.1-1, initiate corrective action within 15 minutes to either reduce power to within Region III of Figure 3.4.1.1-1 or increase flow to within Region I or Region II of Figure 3.4.1.1-1 within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

f. With a change in reactor operating conditions, from two recirculation loops operating to single loop operation, or restoration of two loop operation, the limits and setpoints of specifications 2.1.2, 2.2.1, 3.2.1, ? ' ?; and 3.3.6, shall be implemented within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> or declare the associated equipment inoperable. (or the limits to be "not satisfied") and take the ACTIONS required by the referenced specifications.

l l

• See Special Test Exception 3.10.4.

l l J14 MISC 86020502 - 9 L

/3 i//

Insert 3/4.4-1 page 2 of 2 SURVEILLANCE REQUIREMENTS 4.4.L.1.1 The reactor coolant recirculation system shall be verified to be in

~

operation and not in Region IV of Figure 3.4.1.1.-! at least once per-24 hours.

4.4.1.1.2 Each reactor coolant system recirculation loop flow control valve in an operating loop shall be demonstrated OPERABLE at least once per 18 months by:

a. Verifying that the control valve fails "as is" on loss of hydraulic pressure at the hydraulic unit, and t.

b. Verifying that the average rate of control valve movement is:

1. Less than or equal to !!% of stroke per second opening, and

2. Less than or equal to !!! of stroke per second closing.

4.4.1.1.3 During single loop operation, verify the loop recirculation flow control in the operating loop is in the manual mode at least once per 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

4.4.1.1.4 During single loop operation, verify the volumetric loop flow rate of the loop in operation is within the limit at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

4.4.1.1.5 During single loop operation, and with both THERMAL POWER less than 36% of RATED THERMAL POWER and the operating recirculation pump not on high speed, verify the followitg differential temperature require-ments are met within 15 minutes p rior to beginning either a THERMAL POWER increase or a recirculation loop flow increase and within every hour durinF THERMAL POWER or recirculation loop flow increase:

a) less than 100*F, between the reactor vessel steam space coolant and the bottom head drain line coolant, and b) less than 50*F, between the coolant of the loop not in ,

operation and the coolant in the reactor vessel, and c) less than 50*F, between the coolant in the operating Joop and the coolant in the loopanot in operation.

The differential temperature requirements 4.4.1.1.5.b and c do not apply when the loop not in operation is isolated from the reactor pressure vessel.

4.4.1.1.6 The limits and setpoints of specifications 2.2.1, 3.2.1, Dee,4 and 3.3.6 shall be verified to be within the appropriate limits within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of an operational change to either one or two loops operating.

J14 MISC 86020502 - 10

A %?

f i

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temper-ature following the postulated design basis loss-of-coolant accident will not exceed the 2200*F limit specified in 10 CFR 50.46. J 1 I i 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following i the postulated design basis loss-of-coolant accident will not exceed the limit specified in 10 CFR 50.46.

The peak cladding temperature (PCT) following a postulated loss-of-coolant i

• accident is primarily a function of the average heat generation rate of all ,&

i the rods of a fuel assembly at any axial location and is dependent only secondar- p*

$ ily on the rod to rod power distribution within an assembly. The peak clad tem- .'.,

J perature is calculated assuming a LHGR for the highest powered rod which is [f

_, equal to or less than the design LHGR corrected for densification. This LHGR g9' 7 times 1.02 is used in the heatup code along with the exposure dependent steady <

3 state gap conductance and rod-to-rod local peaking factor. The Technical Spect- l!

e fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this_LHGR of -;; j l _g the hiahest powered rod divided by_ its local peaking factor. l Thg_.1.iaiting' value' 3i i & I Gr t.?U:0^ i; eu.n ir. Ti sce 0.2.1-l R -

jd '

The daily requirement for calculating APLHGR when THERMAL POWER is greater M*

than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-2 tion sMfts are very slow when there have not been significant power or control i

rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the .

j completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER s ensures thermal limits are met after power distribution shifts while still t >

l a11otting time for the power distribution to stablize. The requirement for  ;:I

! calculating APLHGR after initially determining a LIMITING CONTROL R0D PATTERN s

, exists ensures that APLHGR will be known following a change in THERMAL POWER i or power shape, that could place operation exceeding a < thermal limit. ,

l The calculational procedure used to establish the APLHGR N b f;;rn d!'

i 3.2.1 ; is based on a loss-of-coolant accident analysis. The analysis was per- f formed using General Electric (GE) calculational models which are consistent f .

l with the requirements of Appendix K to 10 CFR 50. A complete discussir" of each code employed in the analysis is presented in Reference k Otfferences in this -jl t analysis compared to previous analyses can be broken down as follows. NMG

a. Input Channes

1. Corrected Vaporization Calculation - Coefficients in the vaporization correlation used in the REFLOOD code were corrected.

2. Incorporated more accurate bypass areas - The bypass areas in the i top guide were recalculated using a more accurate technique.

I 3. Corrected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

l

. 4r %. sar op.c ., _ e p n z.s z 4 e < - p = , ,,,,. +.m , n , 4 i 1 7*. -

ae m ~ 4. . + 4,,, w r.a m -3%w%

! GRAND GULF-UNIT 1 8 3/4 2-1 /NNoaENT M. l

hkb INSERT "H" The MAPLHGR limits of Figures 3.2.1-1 are multiplied by the smaller of either the flow depelident MAPLEGR factor (MAPFAC ) or the power dependent MAPLHGR factor (MAPFAC ) corresponding to existinh core flow and power state to assure the adherence Eo fuel mechanical design bases during the most limiting transient. The maximum factor for single loop operation is 0.86.

MAPFAC f

's are determined using the three-dimensional BWR simulator code to analyze slow flow runcut transients. Two curves are prnvided for use based on the existing setting of the core flow limiter in the Recirculation Flow Control System. The curve representative of a maximum core flow limit of 107.0% is more restrictive due to the larger potential flow runout transient.

MAPFAC 's are generated using the same data base as the MCPR P to protect the core fEom plant transients other than core flow increases.

i J12 MISC 86030401 - 2

AW l

POWER DISTRIBUTION LIMITS SASES _

WERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

b. Model Chance

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption

that flow from the bypass to lower plenum must overcome a 1 psi i pressure drop in core. l

2. Incoperate NRC pressure transfer assumption - The assumption used in I the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed. ,

A few of the changes affect the accident calculation irrespective of i CCFL. These changes are listed below.

a. Input Change i

, 1. Break Areas - The DBA break area was calculated more accurately.

b. Model Change

1. Improved Radiation and Conduction Calculation - Incorporation of  !

CHASTE D5 for heatup calculation.

l i

l A list of the significant plant input parameters to the loss-of-coolant .

! accident analysis is presented in Bases Table B 3.2.1-1.  !

I 3/4. 2.2 DE LE TED

\ 3/4.2.2 APRM SETPOINTS e fuel cladding integrity Safety Limits of Specification 2.1 wer ased

on a po distribution which would yield the design LHGR at RATED L t

POWER. The ow biased simulated thermal power-high scram setti and flow j biased simulate hermal power-upscale control rod block fun ons of the APRM 4 i n s.69 nstruments must b djusted to ensure that the MCPR does graded t become less than 4 A;r 4:4fr or that > IX plas strain does not occur in t situation. U,.N l '

The scram settings and r lock settings are adju d in accordance with the formula in this specification en the combina n of THERMAL POWER and MFLPD *M"@

indicates a peak power distribut to ens than an LNGR transient would not j be increased in degraded conditions.

The daily requirement to we y the A control rod block and scram setpoints when THERMAL POWER greater than or ual to 25% of RATED THERMAL l

POWER is sufficient since r distribution shift re very slow when there

! have not been signifi power or control rod changes. The requirement to verify the APRM se nts within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the comp 1 n of a THERMAL POWER increase at least 15% of RATED THERMAL POWER ensures real limits are met aft power distribution shifts while still allotting ti or the power d ibution to stabilize. The requirement to verify the AP tpoints  !

ence r 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initially determining MFLPD to be greater than ures that the consequences of an LHGR transient would not be increased j degraded conditions.

_ _ ,_ _ GRAND _ GULF-UNIT 1_ _ _ _ _ _ _ B_ 3/4 2-2 _

_ _ _ fwwemr Mr. __ .

b V8 i

POWER DISTRIBUTION LIMITS  ;

Bases Table 8 3.2.1-1 -

SIGNIFICANT INPUT PARAMETERS TO THE LOSS-OF-COOLANT ACCIDENT ANALYSIS Plant Parameters; Core THERMAL POWER .................... 3993 MWt* which corresponds to 105% of rated steam flow Vessel Steam Output ................... 17.3 x 106 lbm/hr which cor-responds to 105% of rated steam flow Vessel Steam Dome Pressure............. 1060 psia Design Basis Recirculation Line Break Area for:

a. Large Breaks 3.1 ft2.

b. Small Breaks 0.1 ft2.

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO 4

t.k smi Initial Core 8 x 8 RP 13.4 1.4 K on gu 31,i%

bMcPRg" A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 6.3.3 of the FSAR.

• This power level meets the Appendix requirement of 102%. The core heatup calculation assumes a bundle power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

8A g,*q ggis 14 oprah n, depst ka h-a o d d 'N'N^3 N en s

'^

%d 4. mu e 0l SM b" N ry,atus 4 in$d mc'PR.

GRAND GULF-UNIT 1 B 3/4 2-3 Amoer No._

A We POWER DISTRIBUTION LIMITS BASES -

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Speci'fication 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR ef 1.00, and an analysis of abnormai opera e u dl tional transients. For any abnormal operating transient analysis evaluation g2,19%

with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting tran-sients have been analyzed to determine which result in the largest reducthn in CRITICAL POWER RATIO (CPR). The type of *.ransients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant

' temperature decrease. The limiting transient yields the largest delta J1GPE C.PR. l l 5

,J When added to the Safety Limit MCPR O' IE, the required Mir operatinglimit .n MCPl 7E< The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for  % m,tm g generation of the MCPR operating limits. g g g.,,,,,

, 4 % 4 5 The evaluation of a given transient becins with the system initial s.bM

.! 8 parameters shown in FSAR Table 15.0-2/that are input to a GE-core dynamic mehavior en Mv 6 l

transient computer program > The code used to evaluate pressurization events i )". (*  !

) p'- is described in NEDO-24154(3) and the program used in non pressurization events is described in NEDO-10802 (2) . The outputs of this program along with the

'3l c  ; initial MCPR form the input for further analyses of the thermally limiting bundle g g with the single channel transient thermal hydraulic TASC code described in T - " NEDE-25149 I4) . The principal result of this evaluation is the reduction in l .; L MCPR caused by the transient.

< c ljh The purpose of the MCPR and MCPR p f ;f fi;r;; 3.2.3-1 :nd 3.2.3-2 is to 1 g c, T define operating limits at other than rated core___,a.s flow and power conditions.

n. , o.m._

, 1. i.,,.u- i na . , . . . a ,,_ __s __ __ . u w on n e n i.

l {g .gj qU E ED(5WJon[E M AM Qib5 E A @ r A*$The'MCPRs g l 0 ' j *. are established to protect the core from inadvertent core flow increases such

! g ,j that the 99.9% MCPR limit requirement can be assured.j fr,- IThe reference core flow increase event used to establish the MCPRg is a

1. g 3 j f

" lypothesized slow flow runout to maximum, that does not result in a scram from the APRM neutron flux-high level (Table 2.2.1-1

[ item 2).gith hfsbasisthe PR g curvesgnerated from a series of steady ffgneutronfluxovershootexceedi T.S state core t Ermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line. This ;n ce:;;;h to the 0-~ ;;t;. flu flg. cerar:1 tir; (rig = 3/4 2.3-1). In the actual calcula-

/o% ;tions a conservative highly steep generic representation of the ~~ " - " --

stse 4 flow control line has been used. Assumptions used in the original calculations fhv of this generic flow control line were consistent with a slow flow increase god /M transient duration of several minutes: (a) the plant heat balance was assumed GRAND GURF-UNIT 1 B 3/4 2-4 _ __

4NTMDMENr*

• k d--

INSERT I The maximuE runout flow value is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Tvo flow rates have been considered, 102.5% core flow and 107.01 core flow (for Increased Core Flow operation).

8

-  ?

3NI"I 2074 % 501 3 Nil 2074 %O01 a

9

~

o

• s,

% 4 N J 0

%# ~ g I

s k$ N 9 o-  ;

S.

+5 +b

% d

~

• gg  !

C E

= \?1 -

85 f'-

e 55)g 3 _ a=

=v T.

lE i k E*s 4 I Ig R W55 n i

WiS 3 *

• w

\ -

$05 xx SL - - s @

q 5

5 '5 5553 _a i

fEig og v sagr

• eYYo 9 hE*E z .a 3 a II1l

<suo i l o I i i i l i a 1 2 R R 2 9 2 2 S E g g W3*Od 7FMW3H1031VW Ao W'3J oo GRAND GULF-UNIT 1 B 3/4 2-5 Amsanmunr No. -

AW r

POWER DISTRIBUTION LIMITS BASES __

, MIN!EM CRITICAL POWER RATIO (Continued) to be in equilibrium, and (b) core menon concentration was assumed to be constant.

The generic flow control line is used to define several core power / flow states

, at which to perform steady-state core thermal-hydraulic evaluations. j.rne d UT

! Jo P8F nalyzed corresponded to the maximum core power at maxi-l A The

"'e'um core first state [of rated) after the flow runout.

flow (200.".. Several evaluations l were performed at this state iterating on the aersalized core power distribu-tion input until the limiting bundle MCPR just exceeded the safety limit Specification (2.1.2). Next, similar calculations of core MCPR perfomance were determined at other power / flow conditions cln the generic flow control line, assuming the same normalized core power distribution. The result is a i definition of the MCPRgperformance requirement such that a flow increase event to maximum (302."A) will not violate the safety limit. (Th? assumption of con-l stant power distribution during the runout power increase has been shown to be [t conservative. Increased negative reactivity feedback in the high power limiting

! bundle due to doppler and voids would reduce the limiting bundle relative power ,

j in an actual runout.)  !

The MCPR, is established to protect the core from plant transients other than core flow increase including the localized rod withdrawal error event. >

g Core power dependent setpoints are incorporated (incremental control rod with-a drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).

These setpoints allow greater control rod withdrawal at lower core powers where b

s core thermal margins are large. However, the increased rod withdrawal requires J higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is

,, not violated. The analyses that establish the power dependent MCPR require-

• ments that support the RWL system are cresented in GESSAR II. Appendix 158.

3 Wy of other (core-wide) transients at off-r ted diuvKTs t limited by the reyu b t o setdown the e simulated thermal J power-high scram trip setonk , ,,,m. ion (3.2.2), the rod withdrawal error is * 'S:ittTTransient and estabits 5 -aau_irements, -

• pgser j At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, IMu the reactor will be operating at minimum recirculation pump speed and the modera- cn tor void content will be very small. For all designated control rod patterns Mar 33, which may be employed at this point, operating plant experience indicates that 1%

the resulting MCPR value is in excess of requirements by a considerable margin.

During initial start-up testing of the plant, a MCPR evaluation will be made at 25% of RATED THERMAL POWER level with minimum recirculation pump speed.

The MCPR margin will thus be demonstrated such that future MCPR evaluation be-low this power level wl11 be shown to be unnecessary. The daily requirement for calculating MCPP when THERMAL POWER is greater than or equal to.25% of RATED THERMAL POWER is sufficient since power distribution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER in-crease of at least 15% of RATED THERMAL POWER ensures thermal ifaits are met

$ RAND GULF-UNIT 1 8 3/4 2-6 _ _ _]MENOMrur'_No.

f* SQ INSERT JT" (either 102.5% for Rated Core Flow operation of 107% of rated for

• Increased Core Flow operation).

l b WS l

INSERT "B2-6

The abnormal operating transients analyzed for single loop operation are

! discussed in <eference 5. The current MCPR P limits were found to be bounding.

f l

No change to the operating MCPR limit is required for single loop operation.

l l

INSERT "J"

.For core power below 40% of RATED THERMAL POWER, where the EOC-RPT and the reactor scrams on turbine stop valve closure and turbine control valve fast closure are bypassed, separate sets of MCPR limits are provided for high and low core flows to account for the significant sensitivity to initial core flows.

For core power above 40% of RATED THERMAL POWER, bounding power dependent MCPR l

limits were developed.

l I

l e

J12 MISC 86030401 - 3

/?So POWER DISTRIBUTION LIMITS BASES MINIMUM CRITICAL POWER RATIO (Co **.ved) after power distribution shifts while still a11otting time for the power dis-tribution to stabilize. The requirement for calculating MCPR af ter initially determining a LIMITING CONTROL ROD PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.  ;

i 3/4.2.4 LINEAR HEAT GENERATION RATE l 1

i This specification assures that the Linear Heat Generation Rate (LNGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated. ,

l The daily requirement for calculating LNGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER is sufficient since power distri-bution shif ts are very slow when there have not been significant power or control rod changes. The requirement to calculate LNGR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still allotting time for the power distribution to stabilize. The requirement for calculating LNGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LNGR will be known following a change in THERMAL POWER er power shape that could place operation exceeding a thermal limit.  ;

References:

1. General Electric Company Analytical Model for Loss-of-Coolant Analysis i in Accordance with 10 CFR 50, Appendix K, NEDE-20566, November 1975.

2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NE00-10802).

3. Qualification of the One Dimensional Core Transient Model for Boiling Water Reactors NEDO-24154, October 1978.

4. TASC 01-A Computer Program for The Transient Analysis of a Single Channel, Technical Description, NEDE-25149, January 1980.

5. CGd5 *it.u.4.c Tar 4 e ats 19t..s n4 T<. gram, sigk L.ey

\ openk 4 sos, so.4 uma. r;~Jbor+ febe % lb. w, n,I f

a on Mar 31, Ganar \ Else.{ric. C..speg AnalyG.a.1 %ds! (.c L n- (- 1%

6.

c lard Autph in Aesme Auu. w% tocsaso, AppuJh K.

i A%=,d 2, one Re.0r e.ul.l i..s L y 0 4 -.{. L e v A.e. ,

Nrbo 2c5w 2, TswW i . Tuly, i973 i 7. ccuroiunha cm , % mum utinaed e v % > + a j n#s;, , w, m.

t GRAND GULF-UNIT 1 B 3/4 2-7 /APENb ABrNT 80,

R S'1

\

i INSTRUMENTATION .

) BASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued) feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

A fast closure sensor from each of two turbine control valves provides input to the EOC-RPT system; a fast closure sensor from each of the other two turbine control valves provides input to the second EOC-RPT system. Similarly, a closure sensor for each of two turbine stop valves provides input to one EOC-RPT system; a closure sensor from each of the other two stop valves provides input to the other EOC-RPT system. For each EOC-RPT system, the sensor relay contacts are arranged to form a 2-out-of-2 logic for the fast closureThe of operation turbine control valves'and a 2-out-of-2 logic for the turbine stop valves.

of ejther logic will actuate the EOC-RPT system and trip both recirculation pumps.

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is administratively controlled. The manual bypasses and the automatic Operating Bypass at less than 40% of RATED THERMAL POWER are annunciated in the control room. y se n Q The EOC-RPT system response time is the time assumed in the analysis between initiation of valve motion and complete suppression of the electric arc, i.e., 190 ms, less the time allotted from start of motion of the stop valve or turbine control valve until the sensor relay contact supplying the input to the reactor protection system opens, i.e., 70 ms, and less the time allotted for breaker arc suppression determined by test, as correlated to manufacturer's test

) i.e., 50 as, and plant pre-operational test results.

/ results, Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip Setpoint and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATI The reactor core isolation cooling system actuation instrumentation is provided to initiate actions to assure adequate core cooling in the event of reactor isolation from its primary heat sink and the loss of feedwater flow to the reactor vessel without providing actuation of any of the emergency core cooling equipment.

Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the difference between each Trip 5etpoint ,and the Allowable Value is equal to or greater than the drift allowance assumed for each trip in the safety analyses.

3/4.3.6 CONTROL ROD BLOCV. INSTRUMENTATION The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits. The trip logic is arranged so that a trip in any or,e of the inputs will result in a control rod block.

The OPERABILITY of the control rod block instrumentation in OPERATIONAL CONDITION 5 is to provide diversity of rod block protection to the one-rod-out

) interlock.

GRAND GULF-UNIT 1 B 3/4 3-3

P.z z, 1

LIMITING SAFETY SYSTEM SETTING I

)  ;

8ASES -

REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

9. Scram Discharge Volume Water Level-Hi @

The scram discharge volume receives the water displaced by the ration of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient volume to accept the displaced Thewater reactor at pressures below 65 psig, control rod insertion would be hindered.

is therefore tripped when the water level has reached a point high enough to indicate that it is indeed filling up, but the volume is still great enough to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped. The trip setpoint for each scram discharge volume is equivalent to a contained volume of 26 gallons of water.

10. Turbine Stop Valve-Closure The turbine stop valve closure trip anticipates the pressure, neutron flux, and heat flux increases that would result from closure of the stop valves.

With a trip setting of 40 psig, the resultant increase in heat flux is such thatadequatethermalmarginsaremaintainedduring(_theworstcasetra assuming the turbine bypass valves fail to operate.

11. Turbine Control Valve Fast Closure, Trip Oil Pressure-Low

)

The turbine control valve fast closure trip anticipates the pressure, neutron flux, and heat flux increase that could result from f ast closure of the turbine control valves due to load rejection coincident with failure of the turbine bypass valves. The Reactor Protection System initiates a trip when fast closure of the control valves is initiated by a low EHC fluid pressure in the control valve and in less than 100 milliseconds after the start of control valve fast closure. This loss of pressure is sensed by pressure transmitters which output to trip units whose contacts form the one-out-of-two twice logThis c input to the Reactor Protection System. The trip setpoint is 43.3 psig.

trip setting and a different valve characteristic from that of the turbine stop valve combine to produce transients which are very similar to that for the stop valve. Relevant transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis Report. (T:tX9mtr. t m }

12. Reactor Mode Switch Shutdown Position The reactor mode switch Shutdown position provides trip signals into system trip channels which are redundant to the automatic protective instru-sentation channels and provides additional manual reactor trip capability.

13. Manual Scram The Manual Scram pushbutton switches introduce trip signals into system trip channels which are redundant to the automatic protective instrumentation

) channels and provides manual reactor trip capability.

GRAND GULF-UNIT 1 8 2-9

.~ _ . _ - _ . .

b ES INSERT K ,

The automatic bypass setpoint is feedwater temperature dependent due to the subcooling changes that af fcet the turbine first stage pressure-reactor power relationship. For RATED THERMAL POWER operation with feedwater temperature greater than or equal to 420'F, an allowable setpoint of ei 26.9% of control valve wide open turbine first stage pressure is provided for the bypass function. This setpoint is also applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature. The allowable setpoint is reduced to 6.22.5% of control valve wide open turbine first stage pressure for RATED THERMAL POWER operation with a feedwater temperature between 370*F and 420*F. Similarly, the reduced setpoint is applicable to operation at less that RATED THERMAL POWER with the correspondingly lower feedwater temperature.

i e

• r A

- -, - .. - < - , .,w , - , . , , . . .- - _ . , - - - - , . - - _ . , . - , , . .- -

_c. ., . - - -

Rn INSERT L As indicated in Table 3.3.1-1, this function is autetutically bypassed below the turbine first stage pressure valtee equivalent to THERMAL POWER less than 40% of RATED THERMAL POWER.

INSERT M As with the Turbine Stop Valve Closure, this function is also bypassed below 40% of RATED THERMAL POWER. The basis for the setpoint is identical to ' chat described for the Turbine Stop Valve Closure.

O

i GGNS MAXIMUM EXTENDED OPERATING DOMAIN ANALYSIS March 1986 .

Prepared for MIS $1SSIPPI POWER & LIGHT COMPANY GRAND GULF 1 & 2 NUCLEAR STATIONS Prepared by General Electric Company Eclear Energy Business Operations ,

San Jose, California 95125 5

0

APPIND!r 15.D TABLE OF C0KTEWTS

.P.*E 5

Maximum Extended Sperating Domain 15.D.1-1 15.D Definition of Cufrent Power / Flow operating 15.D.1 15.D.1-1 Domain 15.0.2 Dbjectives of the Maximum Extended 15.D.2-1 Operating Domain 15.D.3-1 Introduction and Summary ,

15.D.3 15.D.4-1 MCPR Operating Limit 15.D.4 25.0.4-1 Abnormal Operating Transient 15.D.4.1 15.D.4 2 15.D.4.2 Rod Withdrawal Error 15.0.4-3 15.D.4. 3 Flow Runout Transient 15.0.4-3 15.D.4.4 Operating Limit MCPR 15.D.5-1 15.D.5 Stability Analysis 15.D.6-1 Loss of Coolant Accident Analysis 15.D.6 15.0.7-1 ,

Containment Analysis 15.D.7 15.D.8-1 Loads Impact on Internal 15.D.8 15.0.5-1 15.D. 8.1 Acoustic and Flow Induced Loads 15.D.8-1 15.D.8.2 Reactor Internal Pressure Differences 15.D.8-2 15.D.8.3 Sundle Lif t Evaluation 15.D.8-2 15.D.S.4 Impact on Reactor Internals ,

15.D.9 1 15.D.9 Flow Induced Vibration 15.D.1D Impact on Anticipated Transient without 15.0.10-1

• Scran 15.D.11-1 15.0.11 Fuel Mechanical Performance Average Power Range Monitor (APRH) Sisulated i 15.D.12 15.0.12-1 Thermal Power Scrae and Rod Block 5etpoints 15.D.13 Feedwater Heater (s)'Out of Service in the 15.D.13-1 Maximas Extended Operating Domain 15.D.13-2 Abnormal Operating Transients 15.D.13.1 15.D.13 7 Other Impact Evaluations 15.D.13.2 .

l 15.D.i

, l TCL:ge/F07231*-2 ,

7/23/64

TABLE OF CONTENTS (Continued) e N

15.D.14 Elimination of sty APRM Trip setdown

~

15.D.14-1 Requirement 15.D.14-2 15.D.14.1 Transient Evaluation 15.D.14-3 15.D.14.2 Loss of Coolant Accident Plant Operating Limit 15.D.14-3 15.D.14.3 Flow Dependent MCPR Limit 15.D.14-3 15.D.14.3-1 '

Power Dependent MCPR Limit 15.D.14-3 15.D.14.3-2 15.D.14-4 15.D.14.3-3 Flow Dependent MAPLHGR Limit Power Dependent MAPLHGR Limit 15.D.14-4 15.D.14.3-4 15.D.14-4 15.D.14.3-5 Governing Dwerall Limit 15.D.15-1

~

15.D.15 Refarences 1

i i

I 15.D.i1 TCL:ge:ra/F07231*-3 7/21/84 1

LIST OF TABLES WUPSER -

M PAGE 15.D.4-1 Analysis Power - Flow Points and CPR

~

Results for BWR/6 Soun' ding Transient 15.D.4-4 Evaluation j 15.D.4-5,6,7 15.D.4-2 Input Parameters and Initial Conditions for Transients for ME0D at 104.3 Power /

73. 8 Core Flow 15.D.4-8,9,20 15.D.4-3 Input Parameters and Initial Condi.tions for Transients for Mf0D at 104.3 Power /

108% Core Flow Sumary of Transient Peak Values Results - 15.D.4-11 15.D.4-4 MIDD Operation Summary of Critical Power Ratio Results - 15.D.4-12

. 15.D.4-5 MEOD Operation APRM Instrumentation Setpoint for ME00 15.D.12-1 15.D.12-1 15.D.13-1 Sumary of Transient Peak Values Results FWOS in F.EOD, EOC1 15.D.13-3 15.D.13-2 Summary of Transient Peak Values Results FWHOS in ME00, 2000 MWD /T before EDC1 15.D.13-4 15.D.13-3 Summary of Critical Power Ratio Results 15'.D.13-5 FWHOS in ME00, E0t1 15.D.13-4 Summary of Critical Power Ratio Results FWOS in ME0D, 2000 IWD/T before E001 15.0.13-6 15.D.14-5 15.0.14-1 Sounding SWR /6 Transient Analysis Result for Elimination of APRM Trip 5etdown Grand Gulf Transient Analysis Results for 15.D.14-6 15.D.14-2 Elimination of APRM Setdown 15.D.14-7 15.D.14-3 Rated Operating Limit MCPR Values 15.D-111 .

TCL:ge:rm/F07231*-4 1/23/S4

LIST OF FIGURES

~

TITLE M ,

Maximum Extended operating Domain Power / Flow Map 15.D.3-1 15.D.4-1 Load Rejection With S'ypass Failure 104. 3 Power /

108% Flow 15.D.4-2 Load Rejection With Bypass Failure 104.2% Power /

73.8% Flow 15.D.4-3 Feedwater Controller Failure 104.2% Power /

108% Flow 15.D.4-4 Feedwater Controller Failure 104.d Power /

73.8% Flow 15.D.4-5 Flow Dependent MCPR Limit for ME00 15.D.12-1 APRM Configuration Setpoint for MEDD 15.D.13-1 Load Rejection With Bypass Failure 104.2% Power /

109.6% flow 370'F feedwater temperature, E0C1 15.D.13-2 Load Rejection with typass Failure 104. 3 Power /

110.9% flow 320'F feedwater temperature. E0C1 15.D.13-3 Feedwater Controller Failure 104.3 Power /

109.6% flow 370*F feedwater temperature, E0C1 .

15.D.13-4 Feedwater Controller Failure 104.2% Power /

110.9% flow 320*F feedwater temperature, E0C1 15.D.13-5 Load Rejection With 5ypass Failure 104.3 Power /

~

109.6% flow 370*F feedwater 1,emperature, EOC1 -

2000 WD/T 15.D.13-6 Load Reject 1'on With Bypass Failure 104.5 Power /

110.9% flow 320'F.feedwater te nerature, 10C1-

• 2000 MWD /T 15.D.13-7 Feedwater Controller Failure 104.2% Power /

109.6% flow 370*Ffeedwatertemperature,E0Cl-2000 WD/T .

s 15.D-iv l

TCL:gc:rm/F07231*-5 '

7/23/84

L151 0F FIGURES NUMBER TITLE 15.D.13-8 Feedwater Controller~ failure 104.2% Power /

110.9% flow 320'F feedwater temperature. EDC1-2000 MC/T 15.D.14-1 Power Dependent ICPR Limit 15.D.14-2 Flow Dependent MAPLNGR Limit 15.D.14-3 Power Dependent MAPLNGR Limit .

15.D-v l

TCL:ge/F07231*-6 7/23/84 i

\

15.0 Maximum Extended Operating Domain

. 25.0.1 Definition of current Pover/ Flow Operating Domain The current power / flow operating domain as given in Figure 4.4.5 of Chapter 4 can be' regarded as a map bounded by the following restrictions:

(1) The 100% rated power condition.

(2) The 105% rated steam flow rod line.

(3) The 100% rated core flow condition.

(4) Low power recirculation system component ca,vitation restriction.

Minimum core flow restrictions on pump speed and FCV position.

(5)

Chapter 6 and 15 justify safe operation of the Grand Gulf Nuclear Stations (GGNS) in this defined region by evaluating all normal and abnormal tran and all design basis accidents to prove that all requirements established by the Code of Federal Regulations are met.

15.D.1-1 TCL:ge:ra/F07232*

7/23/84

15.D.2 Objectives of the Maximum Extended Power / Flow Operatino Domain An extended power / flow operating domain is defined relative to the normal operating map in Figure 4.4.5 for a region satisfying the following:

(1) The additional region is operationally beneficial and achievable.

(2) It is safe to operate in this additional region and all requirements in the code of Federal Regulations are set.

(3) The Technical Specifications required to coyer operation in the extended region do not restrict operation unduly.

(4) The hardware changes required in this region are not major changes to the GE Boiling Water Reactor (BWR) 6 standard plant definition.

This appendix will show that the operating domain is currently defined in Figure 4.4.5 can be safety extended to meet all the above objectives and specifically all requirements in the Code of Federal Regulation.

15.D.2-1 TCL:rm/A072511 7/23/84

15.D.3 Introduction & Sunenary 1

• This appendix presents the results of a safety and impact evaluation for operation of the Grand Gulf 1 & 2 Nuclear Stations in a modified operating envelop called the Maximum Extended Operating Domain (ME00). The ME00 region can be utilized to improve the operating flexibility and capacity factor for the Grand Gulf Nuclear Stations.

Two factors which restrict the flexibility of a boiling water reactor (BPR) during power ascension in proceeding from the low-power / low-core-flow condition to the high-power /high-core-flow condition are: (1) the Final Safety Analysis Report (FSAR) power / flow curve, and (2) Preconditioning Interim Operating Management Recommendations (PCIOMRs) for core without the PCI resistant fuel.

If the rated load line control rod pattern is maintained as core flow is increased, changing equilibrium xenon concentrations will result in less than rated power at rated core flow. In addition, fuel pellet-cladding interaction The considerations inhibit withdrawal of control rods at high power levels.

combination of these two factors can result in the inability to attain rated core power directly. .

The maximum extended operating domain as defined and illustrated in Figure 15.D.3-1 permits improved power ascension capability to full power and provides additional flow range at rated power including an increased flow region to compensate for This expanded reactivity reduction due to exposure during an operating cycle.

One is the expanded operation power flow map can be separated into two regions.

in the lower than 100% core flow region which is termed Extended Load Line region (ELLR) and the other is the expanded region in the higher than 2005 core The combined flow region which is termed Increased Core Flow Region (ICFR).

ELLR and ICFR is termed Maximum Extended Operating Domain (ME00) in the remain-ing content of this appendix. ,

The ME00 analysis consists of three features:

! a) Operation in the ELLR -

b) Operation in the ICFR ~

i TCL:ge:rf/F07233* 15.0.3-1 3/14/85

c) Elimination of the APRM Total Peaking Factor Setdown Technical Specification Requirement. . .

The extended load line region (ELLR) boundary is limited by 75% core flow at 100% power' and its corresponding power / flow constant roi line. This is deter-cined based on a safety and impact evaluation as well as a feasibility study that indicates that this is the highest rod line that is operationally feasible in meeting thermal and reactivity margins. The Increased Core Flow Region

' (ICFR) is bounded by the 105% core flow line. This ICFR boundary is limited by plant recirculation system capability, acceptable flow induced vibration limit and reactor internal pressure differences plus the impact of fuel bundle lif t

! forces on the vessel internal components.

The MEOD evaluation is perfomed for the GE fueled Grand Gulf Nuclear Stations for initial Cycle 1. It is applicable to the nomal annual 12 month cycle operation for Cycle I with the current GE6 fuel design.

The limiting normal and abr.ormal operating transients in Chapter 15 were reevaluated in the MEOD. No change in power dependent operating limit MCPR, l

needs to be made when operating in the MEOD. However, a new set of power dependent operating limit MCPR p is necessary as a result of elimination of the APRM trip setdown requirement. A new set of flow dependent operating limit MCPR is also developed for operation in the ME0D. It is also detemined that f

the fuel mechanical limits are met for all transients occurring in the MEOD.

l It is Overpressure protection analyses were also performed in the ME00.

concluded that peak vessel pressures for the MOD conditions are below the ASME code limit. Therefore, adequate pressure protection is present in the ME00.

I i

The Loss of Coolant Accident and Containment responses as described in Chapter 6 were retyaluated in the MEOD. It is found that the responses are bounded by the current design analysis. ,

7 Y

TCL:gc:rf/F07233* 15.D.3-2

• 3/14/85 l

Thermal hydraulic stability was evaluated for its adequacy ,with respect to the ,

General Design Criterion 12 (10CFR50, Appendix A). It is shown that DE00 operation, satisfies this stability criterion.

The effect of increased Reactor Internal Pressure Differences (RIPD), acoustic loads, flow induced loads and fuel bundle lift forces on the reactor internal components and fuel channels due to increased core flow were evaluated to show that the design limits are not exceeded. The effect of increased flow rate on the flow-induced vibration responses of the reactor internals was monitored during startup testing and evaluated to ensure the " responses are within acceptable limits for Grand Gulf Stations.

Several impact evaluations were also performed to justify operation in the MEOD.

This includes impact of the ME00 on Anticipated Transient Without Scram (ATWS), fuel assemblies, fuel channel bypass flow, creep and control blade interference. It was found that all acceptance criteria and design limits are met.

The Average Power Range Monitor (APRM) simulated therral power scram and rod The block configuration are redefined to accommodate operation in the MEOD. -

same protection margin as the current configuration is maintained in the new definition.

This appendix also justifies operation of Feedwater Heater (s) Dut of Service All evalua-(FWHOS) as described in Section 15B in the MEOD (ELLR and ICFR).

tions described in Appendix B were reevaluated or reviewed in the MEOD to ensurethatFWHOSoperatibninthisregionissafeandfeasiblewiththerequired additional modifications to the technical specification MCPR limits. ,

TCt:ge:rf/F07233* 15.D.3-3 ,

3/14/85

N N. -% ~ ~

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Finally, this appendix justifies the replacement of the APRM trip setdown requirements by more meaningful power and flow dependent Mayimum Average Planar ,

Linear Heat Generation Rate (MAPLHGR) limits to reduce the need for manual New setpoint adjustment and to allow for more direct administration of limits.

power dependent MCPR p limit requirements are also established.

Lastly, due to the different MCPR limits required as discussed in the main text, Appendix 15C for single loop operation and Appendix 158 for feedwater heater (s) out of service, a new power dependent MCPR multiplier (K p ) is developed to simplify the MCPRp implementation .

d TCL:gc/F07233* 15.D.3-4 7/23/84 v . -- ----r-, ---- - _,-.r

A. NATVRAL CIRCULATION 7051710N QI OPEN15C) *

3. W W RICIRC. PUHF SPED VALVE MININUM 7051710N

,C .

D. 1DW REC 1RC. FUMP SPEED TALVE NAXINUMRA E. ANALYTICAL 1ATA LIMIT OF AUTOMATIC 10AD FOLLOWIWC -

.F . AXALYTICAL UPPER LIMIT OF AUTOMATIC LOAD FOLLOW

' F0WER FWW .

C. LOWEST ALLOWABLE FIDW AT'RATD F0WER (2001 F. 75tF) -

,. E. 31 CHEST AllbWABLE FLOW AT RATO POWER (2001 F.1951 F KAC7X: EXTDiDED LOAD LINE RIC10N e FW1Jr: INCREASED CORI F1&'** RIC10N 120 -

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MAXIMUM EXTENDED OPERATING DOMIN POWER / FLOW MP lM1551551PPI

• PWER & LIGHT k

l

15.D.4 MCPR Operating Limit 15.D.4.1 Abnonnal Operating Transients All abnormal operating transients described in Chapter 15 were exami">ed for Maximum Extended Operating Domain (MEOD) operation. A bounding analysis is performed using a standard BWR/6 238 size 748 bundle pitat with the highest enriched GE-6 fuel type as a basis. The core average power density of this standard BWR 6 plant is almost identical to the 251 size 800 bundle GGNS. The fuel type used in this analysis represents the bounding nature of this analysis.

This bounding analysis was performed at various MEOD boundary power / flow conditions of Figure 15.D.3-1 at the end of the 18 month equilibrium cycle using both the computer models described in reference 15.D.15.1 and 15.D.15.2.

The various power / flow conditions and transients analyzed are tabulated in Table 15.D.4-1. The CPR resalts of this analysis are also tabulated in Table 15.D.4-1. This bounding evaluation concluded that the CPR results for all the cases analyzed in the MEOD are bounded by the current Technical Speci-fication power dependent MCPR p limits.

The following limiting pressurization and cold water injection abnonnal operat-ing transients were reevaluated in detail for the Grand Gulf Stations to confirm l

! that the bounding analysis performed for equilibrium cycle is bounding for GGNS at Cycle 1:

(1) Generator Load Rejection With Bypass Failure (LRNBP) l (2) Feedwater Flow Controller Failure (FWCF) i The transients were analyzed at the end of Cycle 1. An initial power of 104.2%

of rated was used. The core flow condition chosen for the analysis were the minimum f1pw of 73.8% of rated and maximum achievable flow of 108% of rated.

l Plant heat balance, core coolant hydraulics and nuclear transient parameter data were developed and used in the above transient analysis. The initial condition for these lowest and highest flow analysis point at rated power is l

presented in Tables 15.D.4-2 and 15.D.4-3. The computer model described in TCL:ge:rf/F07234* 15.D-4-1

• 3/14/85

l ate these event s. The transient peak

. Reference 15.D.15-1 was used to simu values results and critical power ratio (CPR) results for the cases analyzed at 104.25 power (lowest and highest flow) are sumarized in Table 15.D.4-4 and 15.D.4-5 respectively. The transient responses are presented in Figure 15.D.4-1 to 15.D.4-4. The results of this Grand Gulf unique evaluation show that the ACPR results for all the cases analyzed are bounded by the bounding BWR/6 Standard Plant analysis presented in Table 15.D.4-1. Therefore, no change in MCPR limits are required for operation in the ME00. Section 15.D.14 provides a p

new set of MCPR limits to support the elimination of APRM trip setdown p

) requirement.

The BWR/6 standard plant bounding overpressure protection transient analysis using the computer model described in Reference 15.D.15-1 is performed at the power flow conditions described in Table 15.D.4-1. The bounding MSIV closure flux scram event resulted in a peak pressure of 1273 psi at 1101 core flow condition. This result is verified for the GGNS which results in a peak pressure of 1262 psig at the 108% ICF condition. Therefore, it is shown that the peak vessel pressures for the ME00 conditions are below the ASME code limit l

of 1375 psig. Hence, adequate pressure margin is present in the MEOD.

The 100'F Loss of Feedwater Heater (LFWH) Transient results described in the main text of this chapter are applicable to the MEOD. A generic statistical LFWH analysis using the computer model described in Reference 15.0.15-3 and methodologies described in Reference 15.D.15-4 was perforined utilizing a large data base throughout the power / flow map. It is found that the LFWH responses initiated from all off-rated power / flow conditions (including MEOD) are bounded l by the generic rated condition 95% probability 951 confidence values. This generic conservative ACPR val n n.10 t$iich is bounded by the values described j

in Chapter 15.

15.0.4.2 Rod Withdrawal Error The rod withdrawal error (RWE) transient documentet; in Chapter 15 is analyzed f

l vsing a statistical evaluation of the minimum critical power ratio (MCPR) and

• 11near heat generation rate (LHGR) responses to the withdrawal of ganged TCL:ge:rf/F07234* 15.D-4-2 -

3/14/85

control rods throughout the operating power / flow map including the ME00 region (see Figure 15.8.-9 of GE55AR !! 238 Nuclear Island). Therefore, the current Technical Specification MCPR limit is adequate to protect the RWE in the ME00. ,

15.D.4.3 Flow Runout Transient l

The current Technical Specification flow dependent MCPR operating limit (MCPRf )

was determined based on the slow recirculation flow runout transient event.

This curve was generated with some contingent conservative margins in the original design process. This event was reanalyzed,as part of the MEDO program to include the highest rod line for the ELLR, up to 102.51 maximum flow and the '

ICFR, up to 1071 maximum core flow. This analysis utilized the latest design 1

procedure in which some of the unnecessary original contingent design conserva-tism were removed. The new flow dependent MCPRf curves are presented in Figure 15.D.4-5. It is also shown that these curves still bound the other flow dependent abnormal transients which are considered to establish the flow dependent MCPRf limits.

1 15.D.4.4 Operating Limit MCPR i The analyses presented in the above subsections concluded that the current Technical Specification power dependent MCPR, limits are adequate for operation in the ELLR and ICFR of ME00. However, new MCPR, limits are required to eliminate the APRM trip setdown Tech Spec requirement. The flow dependent MCPRf curves are to be modified as described in Figure 15.D.5-5 for operation in the ME00. Section 15.D.14 describes the set of MCPR, limits to support the f

elimination of APRM trip setdown requirements.

d 4TCL:ge:rf/F07234* 15.D-4-3

, 3/14/85 .

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

~' _ _

Table 15.D.4-1 Analysis Power-Flow Points and CPR Results for o

~

BWR/6 Sounding' Transient Evaluation Transients ACPR(a)

Power (5)/Flov (%)

LRNBP 0.071 70/40 0.095 FV F CLDLP 0.087(D}

FCVO 0.208 LRNBP . 0.066 83/55 LRNBP 0.076 104.2/75 FWCF 0.084 FCVO 0.072 LRNBP 0.110

- 104.2/200 FWCF 0.095 LRNSP 0.114 104.2/110 0.097 FW:F LRNSP 0.125 53.5/116 0.284 FW:F (a) Option A adders included for the Reference 15.D.15-1 analysis. ,

(b) This transient is covered by M PR g, not MCPRp .

NOTE:

LRMBP Gent rator Load Rjection With Bypass Failure FWCF Feedwater Controller Failure (saximum demand)

CLDLP Cold Loop Startup FCVO Flow Control Valve Opening 15.0.15-1 and the The LRt.BP and FW F transients are analyzed using Reference CLDLP and FCVD transients are analyzed using Reference 15.D.15-2.

15.D.4-4 TCL:gc/F07234*

7/23/8A

- j l

Table 15.D.4-2

- Input Parameters and Initial Conditions for

• Transients and Accidents for MEOD,104.21 Power, 73.85 Flow l'. Thermal Power Level, MWt 3994 (104.2% rated)

Analysis Value

2. Steam Flow, alb per hr 17.22 Analysis Value .

3. Core Flow, alb per hr 83.0

4. Feedwater Flow Rate, alb per hr 17.22 Analysis Value .

Feedwater Tempereture. *F 425 5.

Vessel dome pressure, psig 1045 6.

Core exit pressure, psig 1053 7.

Turbine Bypass Capacity, 1 NBR 35 8.

Core Coolant Inlet Enthalpy 522 9.

Btu per Ib

20. Turbine Inlet Pressure, psig 961 8x8R

11. Fuel Lattice

12. Core Leakage Flow, % 10.65I ")

13. Required MCPR Operating Limit 1.27 First Core

14. MCPR Safety Limit for Ir.cidents of Moderate Frequency First Core 1.06 Reload Cores 1.07 t

15. Doppler coefficient (-)t/*F' O.132(a)

Analysis Data I

TCL:gc:rf/F07234* 15.D-4-5 3/14/85

I l

Table 15.0.4 (Continued) l

16. Void Coefficient (-)t/s Rated Voids Analysis Data for Power increase Events 14.0(a)

Analysis Data for Power Decrease Events 4.0(a)

17. Core Average Rated Void Fraction, 1 48

18. Jet Pump Ratio. M 2.32

19. Safety / Relief Valve Capacity, 1 NBR

1. 1145 psig 102.4 Manufacturer . Dikker Quantity Installed 20

20. Relief Function Delay, seconds 0.4 ,

21. Relief Function Response Time Constant, sec. 0.1

22. Setpoints for Safety / Relief Valves 1175,1185,1195,1205,1215 Safety Function, psig 1145,1155,1165,1175 Relief Function, psig

23. Number of Valve Groupings Simulated 5

Safety Function, No.

Relief Function, No. 4

24. High Flux Trip,1 NBR 127.1 AnalysisSetpoint(122x1.042),1MBR 1095

25. High Pressure Scram Setpoint, psig

26. Yessel Level Trips, Feet Above Separator Skirt Bottom 5.88 Level 8 - L8), feet Level 4 - L4), feet 4.03 Level 3 - L3), feet 2.16 Level 2 - (L2), feet (-)2.182

27. APRM Thermal Trip Setpoint, 1 NBR 118.8 0.19 28, RPT Delay, seconds TCL:ge:rf/F07234* 15.D-4-6 l

! 3/14/85 .

Table 15.D.4 (Continued)

29. RPT Inertia Time Ccnstant for Analysis, sec. ~5

30. Total Steamline Volume, ft* 4358 (a) These values for Reference 15.D.15-2 analysis only.

Reference 15.D.15-1 values are calculated within the code.

TCL:ge:rf/F07234* 15.D-4-7 .

3/14/85

l j

Table 15.D.4-3

~

Input Parameters and Initial Conditions for -

Transients and Accidents for MEOD, ID4.21 Power 108.05 Flow Thermal Power Level. MWt 3994 (104.2% rated) 1.

Analysis Value Steam Flow, alb per hr 17.29 2.

Analysis Value Core Flow, alb per hr 121.5 3.

Feedwater Flow Rate, alb per hr 17.29 4.

Analysis Value Feedwater Temperature. *F 425 5.

Vessel dome pressure, psig 1045 6.

Core exit pressure, psig 1056 7.

35

8. Turbine Bypass Capacity, 1 NBR Core Coolant Inlet Enthalpy 532 9.

Stu per Ib 959

10. Turbine Inlet Pressure, psig 8x8R

11. Fuel Lattice I

10.65*)

12. Core Leakage Flow, %

1.18

13. Required MCPR Operating Limit First Core

14. MCPR Safety Limit for Incidents of Moderate Frequency 1.06 First Core 1.07 Reload Cores 0.132(a)

15. Doppler Coefficient (-)t/*F Analysis Data TCL:ge:rf/F07234* 15.D-4-8 3/14/85

Table 15.D.4 (Continued)

O

16. Void Coefficient (-)t/1 Rated Voids Analysis Data for Power increase Events 14.0(a)

- Analysis Data for Power Decrease Events 4.0 (a)

17. Core Average Rated Void Fraction, % 41.0

18. Jet Pump Ratio. M 2.32

19. Safety / Relief Valve Capacity, 1 NBR

1. 1145 psig 102.4 Manufacturer ,

Dikker Quantity Installed 20

20. Relief Function Delay, seconds 0.4

21. Relief Function Response Time Constant, sec. 0.1

22. Setpoints for Safety / Relief Valves 1175,1185.1195,1205,1215 Safety Function, psig 1145, 1155, 1165, 1175 Relief Function, psig

23. Number of Valve Groupings Simulated 5

Safety Function. No. 4 Relief Function, No.

24. High Flux Trip, % NBR 127.1 Analysis Setpoint (122x1.042),1 NBR 1095

25. High Pressure Scram Setpoint, psig

26. Vessel Level Trips. Feet Above Separator Skirt Bottom 5.88 Level 8 - L8), feet 4.03 Level 4 - L4), feet 2.16 Level 3 - L3), feet Level 2 - L2), feet (-) 2.182

27. APRM Thermal Trip ,

118.8 Setpoint,1 NBR 0.19

28. RPT Delay, seconds TCL:ge:rf/F07234* 15.D-4-9 3/14/85

Table 15.0.4 (Continued)

29. RPT Inertia Time Constant for Analysis, sec. - 5 i

30. Total Steamline Volume, ft8 4358 (a) These values for Reference 15.D.15-2 analysis only.

Reference 15.D.15-1 values are calculated within the code.

1 1

0 l

TCL:ge:rf/F07234* - 15.D-4-10 3/14/85

1

' E Table 15.D.4-4 I'I Sassaary of Transient Peat Values Results - 104.2% Power MEOD Peak Peak Peak Peak Neutron Dome Vessel Steamitne Core Flow Flux Pressure Pressure Pressure Transient (1 NBR) (5 MBR) (psig) (psig) (psig) Figure Load Rejection 108 II 135 1200 1236 1209 15.D.4-1 with Bypass Failure

' 73.8 104 1205 1230 1209 15.D.4-2 Feedwater 108 (b) 111 1163 1195 1162 15.D.4-3 ,

Contro11er .,

Failure, Max. .

Demand ~.

• 1190 1167 15.D.4-4 i 73.8 111 1169 ,

(al Feedwater is 425'F.

(b) Maxless achelvable core flow with 425'r feedwater.

M 15.D.4-11

TABLE 15.D.4-5 .

Sumary of CPR Results - 104.21 Power ME00 I *I Core Flow -

ICPR CPR MCPR Transient (1 NBR) 108.0(b) 1.18 0.05 1.13 Load Rejection ,

. with Bypass Failure

" 100.0 1.18 0.05 1.13 73.8 1.27 0.05 1.22 108.0(b) 1.18 0.09 1.09 Feedwater Controller Failure, Max.

Demand 100.0 1.18 0.09 1.09

" 73.8 1.27 0.09 1.18 (a) Feedwater Temperature is 425'F. Option A adders included. ,

(b) Maximum Achievable core flow with 425' Feedwater.

15.0.4-12 TCL:gc/F07241 7/24/84

I NEUTRON WLUX 2 PERK FUEL CENTER TEMP 3 RVE SURFf CE Ef4T FLUX 150* -

4 FEEDWRIEF FLOW.

5 VESSEL $1 ERM FLOW

~

5 103.

W \

E g

y -

E w 50.

5 E

~

T f i>

I D-

v. s. s.

- 0. 2.

TIME ISEC)

~

1LEVELIINN-MEF-5EP-5K1RT 2 W R SENS 3 LEVEL (INCHES) 3 N R SENS 0 LEVEL (INCHESI 203. W EORE IN.E l FLOW (PCT) 5 DRIVE FLC.4 1 (PCT) 100 d --

3 !8, u f  %

0. ,

7 100. 1

g. 6. ' 8.

2.

= istC>

FIGUE GENERATOR LOAD REJECTION WITH SYPA55 FAILURE,104.2: 15.D.4-1 POWER M1551551PPI POWER 8 LIGHT 1081 FLOW EOC1 L

l 1 VESSEL PP ES RIE (P511 '

2 STM LINE PRES MIK LP511 l

, ' 3_SAFETT Vf LVE FLOW t.) )

300. 4 RELIEF VF LVE FLOW fel 5 BTPASS Vf LVE FLOW t.)

- 6 TURB STEf M FLOW iPCTI

~

200. .

1

?

8 100- ,

g x F

j

0. -3@ . 635 E3 4 [ 35 4 8.

0. 2. 4. 6.

TIME ISEC) t 1 VOID READ 11VITT ,

2 000PLER F fACTIV1TT /

3 SCRA". Rf8 :11VITT

[

3* -

j y 10lR RU C11v111 7

, 0- 1 - -

'u -

I E C - .

,y l '

b

'd -

E .

2,~...,s,,,,

2. 3. 4.

0. 1.

I TIME ISEC)

FIGURE M1551551PPI GENERATOR LOAD REJECTION WITH BYPASS FAILURE,104.21 POWER, 15.0.4-1 POWER & LisNT 1081 FLOW E0C1 (Cont.)

i

1 NEUTRON WLUX  !

2 PERK FUEL EENTER TDP 3 RVE SURFf CE HER1 FLUX 150. 4 FEEDH41EF FLOW

~ * - 5 VESSEL 51 ERM FLOW ,

5 - k ~

< 100. 3 A 4 g L e -

g sD. -

g 5

l [ '

~

L- '- "' --

1 D. 6. 8.

O. 2. 4.

TIME ISEC) 1 VESSEL PpES R15E (P511

2 SIM LINE PRES RISE (P511

! 3 SW ETT vf LVE FLOW (.)

300. y RELIEF VF LVE FLOW (el 5 BTPR55 VF LVE FLOW (.)

6 TURB STEf M FLOW (PCTI l

200.

u 9/NN w 100.

i E35 83 5

v. s. s.

0~o'. 3 6 - 2. TIME ISEC)

I FIGURE GENERATOR LOAD REJECT]DN SYPA55 FAILURE,104.21 POWER.

MI551551PPI 15.0.4-2

[

POWER & LIGHT 73.8% FLOW E0C1 r

b ._ _ _ . . _ ~ _ _ _ _ _ _ _ _ _ _ , _ _ _ __ _ _ _ , _ _ . . _ _ . . . _ . _ _ _ _ _ . . , _ _ _ _ . _ _ . . _ . _ . . . _ _ _ _ _ _ _ _ . _ _ _ . _ - . _ . _ _ _ . .

. I LEVELIIN N-REF-SEP-5KIRT

~ 2 N R SENSE D LEVELiINCE st 3 N R SENSE D LEVELIINCES) 200* 4 CORE int.E 1 FLDH (PCT!

. - 5 DRIVE FLCW l (PCTI 100.

4E _A y 0.

my ' W ,, L-100, ,,,,s,,,, _

2. 6. B.

0. W.

TIME ISEC) ,

1 VOID RERf TlVITT 1, 2 COPPLER T ERCTIVITY f 3 5(RnN REr :TIVITT 1- iT T 0Tr m r :11Vili 7

l 0. 11  :

G Y

C -l.

! F  :

i 1 -

g _

W 5 *-

2. 3. 4.

-2.O. 1. ~

TIME ISEC) l l

F 1jGjRj-2

, . M1551551PPI PWER & U GHT GENERATOR 3.8% FLOW. EOC1 LOAD REJECT 10r SYPASS FAIL (Cont.)

i 1 NEU1RDN f LUM 2 PERn Fift r. ENTER TEMP '

i 3 RVF. $LRTI M tt.f1T FLUX

, 150. ;77g7g;73),;, ,ggg 4 u 5 VE5SEL 51 CRM FLOW

)

. 1[ 5 - ::==E:==== (

t i

g i n j '

50' gL .

L%

4

~

l r

i-U U

• O.

~

5. 10. 15. 20.

D.

TIME ISEC) l

' ' 1 LEVEL llNN-MTF-5EP-5KirT 2 w n SEN5! D LEVELl1N:4 51 i

3 N R SENN OLEYFdINET53

~

i t 150. EtCN. t If LO=4 tFt11 i 1 (PCTI I

~ 5 DRIVE l'LL y u 5 5 100. -

l i

i My- \

so. g W

~

i a

n, D.'"'''5.

gg, TIME ISEC) 35, i

FIGURE FEEDWATER CONTROLLER FAILURE.104.2% PWER,108% FLOW EOC1 M1551551PPI 15. D .4 -3 POWER S LIGHT 1

3

. I VE$SEL PF CS RISE IPSI)

? STM LINE PRES filSE IPS1' 3 il RINI,Of % $ RISF (P$]g N 1 $U5 W u/Lgi

. 125. - h ?r44)31EF iiT VI LVE FLOW IPti) g l g 6 luN. %TLI FI TLON (PCT 1

~

l '

e

- g 3, t u- ~

25. '

1_A.'L 33 -: (g i h a

23. ....i.... 20.

5. 10. 15.

D.

TIME ISEC1 n

i 1 untr. r rfj'i.'19 9' Ii* 11T It.t i 1.4 . i.

l*  : *.: ti:': ..* 1 11V1 M i.),fT,7: . . *. );]Y.~

t

~

I*

~

D.

== 5 h  !

]

1 E-7 xi-g .

f V -

W 5 n i f -

.... .... 15. 20.

5. 10.

i 2.O. TIME 15ECl .

i I E0C1 FIGUPE FEE 0 WATER CONTROLLER FAILURE.104.2% POWER,1081 FLOW 15.D.4-3 M1551551PPI  ; cont.) __

POWER 8 LIGHT i

l - - - - ~ , - - . . _ , , _ _ "N* '--P- - . , _ . _ _ _ _ _ - _ _ ,

1 s

e 1 NEUTRON Y LUX 2 PEAK FtlEL CENTER TEMP 3 RVE RtJRr6 CF EAT FLUX 150.! GIENT.iUTC5it

u 5 VE5SEL 51 ERM FLOW 100. ~2 b

l 50-  :

m' ' x' .s b N -

D.' '5. 10. 15. 20.

0.

TIME 15EC) 1 LEVEL TIN <-REF-SEP-5K!rtT 2 W R SEN:.f .) LEVEL IIN' HE'.'I l 150. ,$ gofg h'tl:3 hk l 5 D31VE Ft( 4 1 IPC13 103.

us J I us

~

33' 22#- N _ -A

. [, - x- W ,P %

H N

10 15. 20.

o*D. 5.

TIME ISEC)

I EOC1 r1G P.E M1551551PPI FEE 0 WATER f,0NTROLLER FAILURE,104.21 POWER, 73.8% FLOW POWER 8 LIGHT f -

'~

l YESSEL N ES RISE (PS11 2 SIM LINE PRES RISE IPSI)

[ 3 TURDINT T RT$ R E IPSI) y ColCTAt t i W (DIU/LDI c' 23' 5 RTLIEF Vt LVE FLOW IPCTI E s 1N RD S1U M FLOW LPCTI I

w. 7

[F ,

2s.

5' b 17 - ->

(6 s5 E M5 5- E '

25.'- '-

5. 10. 15. 20.

0. TIME ISEC1 1 VnlD Fail"PY

• 0,U 1 lJ ."** 1 1 C11 T

. ;; r 3, , 1611. ~

it',1;llit: 1. .I  ; I V l'I'.

I-

. 0. J*-5 .

\ ~

f E -1. ,

~

5 i u "

h

' 20.

5. 10. 15.

! -2'o. TIME (SEC)

FIGURE MI551551PPI FEEDWATER CONTROLLER FAILURE,104.21 POWER, 73.8% FLOW E0C1 15. D.4 -4 POWER & LIGHT (Cont.)

1 1.7 . , ,

. i ia I' i

7 .1 1 _!

i

-}

3i.. .

.. -, , ?, , I. ,. i

.. i .. .

. . . i.

.s. .

.,i_ ,,,

i ,,

i,,, .... , . ,

.. . ,. . i. .

... . ,,. i. , ... 1 ~*

,i,, .. . .

ii i a .

- 1.8 ,,

1 _- .

~ ' ~ ~

x_ .

FOR MAX FLOW

• 107.01

.-A-

/7  ;

g_ __ 1 MCPRg0.006948F = MAX (1.18} 1.8114

.ii x .

,N !.L '...,,

u - ..;

.1 .

C x -

1.5 ,

-s,.

1 ,

i3< . .

m si .

., I"'

,s . , .

e A elt. i * ,

_. 1

s. .

f

• e

.\ \  ;

'% .\' y

I 4 1 .

e

- n ' +

FOR MAX FLOW = 102.5% _

s 1 1.4 \ ' '.'1- M PRg = MAX (1.18,1.7458. ,

O.006593r) L

' M1 '11 - ,-

3 g 1 I -

-.5 .

. x .x -

s s .

~~ \ %-'\1 -

1.3 6

b J 1 s _- ,

y g

= \ Q, ___,l_ -

-J <- ,

._ ._ g X _; .

3 1 .

l x s __

s s.

i __

4

.s l . .

i j j ,) - , , .

i e

- , , . >>. i.

ie t , i .

I J

' I i 1.0 - i 120 i

60 80 100 0 20 40 i l

CORE FLOW (1 RATED). F l

I FIGURE FLOW DEPENDENT MCPR LIMIT FOR MEOD 15.D.4-5 MI551551PPI POWER & LIGHT l l .

l l

15.D.5 stability Analysis The General Electric Company has established stability criteria to demonstrate

- compliance to requirements set forth in 10CFR50 Appendix A General D These stability compliance criteria consider potential Criterion (GDC) 12.

Itait cycle response within the limits of safety system intervention and assure that for GE BWR fuel designs this.eprating mode does not result in specif Furthermore, the onset of power acceptable fuel design limits bein's exceeded.

esci11ations for which corrective actions are necessary is reliably and re detected and suppressed by operator actions and/or automatic systes fun The fuel performance during limit cycle oscillations is characteristically ,

' dependent on the fuel design and certain fixed systes features (high neu It is therefore flux scram setpoint, channel inlet erifice diaeeter, etc.).

possible to determine the acceptability of fuel designs independent The stability compliance of those GE SWR fuel designs and cycle parameters.

contained in the General Electric Standard Application for Reactor F Reference 15.D.15-6) is demonstrated on a generic basis in Refere

~ and therefore a specific analysis for each cycle is not required.

For operation in the Maximum Extended Operating Domain (MEOD margin (defined by the core decay ratio) is reduced at higher powe Therefore at the limiting condition for stability, natural given core flow.

circulation flow, operation at the maximum However, extended load the norm 1 line wil higher decay ratio and therefore lower stability margin.

realistic operating region has the lowest stability margin at the max speed /sinimum valve position flow (minimumThisforced increasedcircu to 43% core flow for Grand Gulf ((11ustrated in Figure 15.D.3-1).

l I

flow n1stive to natural circulation results in a significant increas stability margin for the maximum extended operating domain Operation below sinf aum forced circulation Operation in flows can tests at operating SWRs.

only occur during transients, e.g., two recirculation pump trip.

this region is addressed in a set of GE operating recommenda 15.D.15-7).

15.D.5-1

~

TCL:gc/F07236*

7/23/84

For demonstration of compliance with CDC 12, the generic stability ana Reference 15.D.15-5 is independent of stability margin sinceInthe reactor ,

- assumed to be operating in a limit cycle condition (no stability margin) addition, analyses are performed at various power / flow conditions to de strate that fuel design liaf ts are not exceeded during 11 alt cycle ope svy region of the power / flow map. ,These snelyses have shown th perforsance is a function of the oscillation amplitude defined by th ence between the high neutron flux scras setpoint and the operating pow level.

Therefore, higher power levels (MEDD) result in smaller oscillations u to the scram setpoint and subsequently less limiting fuel performance.

The analyses of Reference 15.D.15-5 therefore support15.0.15-6. operation in Extended Operating Dosain for these fuel designs contained in Referenc As discussed above, in addition to these analyses, GE has issued a 15.0.15-7) which inform the reactor operating reconcendations (Reference

- operator how to recognize and suppress unanticipated oscillations w Together, the analyses and operator encountered during plant operation. liance recommendations support operation in the MIDD region and demonstra to GDC-12.

l

\ .

I 15.0.5-2 TCL:ge/F07236*

7/23/84

15.D.6 _Loss of Coolant Accident Analyses .

A bounding BWR/6 Loss of Coolant Actident (LOCA) analysis was performed Maximum Extended operating Domain (ME00) boundary defined in Figure The results were reviewed for GGNS., .It is found that the current MAPLHGR limits presented in Chapter 6 are adequate to coverThe theflow entire maximum dependent extended operating domain as defined in Figure 15.D.3-1 MCPR operating limits used in the LOCA analysis bound the new flow depend There-MCPR limits required for MEOD which is illustrated in Figure 15.D.4-5.

fore, the LOCA analysis presented in Chapter 6 are applicable in the pow domain defined in this appendix with the current MAPLHGR limits.

The effect of potential limit cycle oscillations on LOCA analyses was exam A bounding BWR/6 evaluation was performed for a limit cycle oscillation The maximum power oscillation at this power and core power and 60% core flow. To bound this oscil-flow was found to be s71 higher than the initial power.

lation, LOCA calculations were performed starting at 100% power and 60%

flow.

The results of this analysis showed no early boiling transition prior to Consequently, these conditions produce less severe LOCA jet pump uncovery.

analysis results than are obtained at the 105% steam flow /100% core fl condition in Chapter 6.

i 1

i 15.D.6-3 l

TCL:ge:rf/F07237*

3/14/85 .

15.D.7 Containment Analysis '

A containment analysis is performed in the MEOD region ~for Grand Gulf Stations. The peak drywell and wetwell pressures, peak suppression pool temperatures, chugging loads, condensation oscillation and pool swe~.1 containment responses were eveluated to bound the entire ME00 region with reduced rated feedwater temperature operation at 320*F to provide a bounding analysis to justify feedwater heater (s) out of service operation (described in Appendix B and Section 15.D.14) in the ME00.

The analysis shows peak drywell pressure for the worst ME00 combined with FWH05 operation condition of 38.0 psia. The corresponding peak drywell pressure is 23.3 psig which is 1.3 psi above the Chapter 6 value of 22.0 psi However, this value is still below the design limit of 30 psig reported in Chapter 6. The limiting break is switched from main steam line break to the The peak suppression pool temperatures, chugging recirculation line break.

loads, condensation oscillation and pool swell boundary loads are all found to be bounded by the rated power analysis in Chapter 6.

15.D.7-1 TCL:gc/F07238 j

3/14/85 l

i

~

15.D.8 Load Impact on Internals l

15.D.8.1 ~ Acoustic and Flow Induced Loads The scoustic loads are lateral loads on the vessel internals that result fres propagation of the decompression wave created by a sudden recirculation-suction-line break. The acoustic loading on the vessel internals is proportional to the total pressure wave amplitude in the vessel recircula-tion outlet nozzle. The total pressure amplitude is the sum of the initial pressure subcooling plus the experimentally determined pressure undershoot below saturation pressure. A larger downconer subcooling results in a lower saturation pressure, thereby, having a larger total pressure agli-tude and resulting in large acoustic loads. The maximum subcooling in the MEOD was found to be 48.68 BTU /LB.

The high-velocity flow patterns in the downcomer resulting from a recircu-lation- suction-line break create lateral loads on the shroud and the jet

, pump. These loads are proportional to the square of the critical mass flux rate out of the break. The additional subcooling in the downcomer resulting from operating in the MEOD 1eads to an increase in critical flow and, therefore, in flow induced loads.

l The reactor internal most impacted by acoustic and flow induced loads are the shroud, shroud support and jet pump. The impact on these components were generically analyzed with a maximum subcooling of 83.6 STU/LB It is associated with final feedwater temperature reduction operation.

found that these components have enough design margin to handle these loadings.

15.D. 8. 2 Reactor Internal Pressure Difference Loads A reactor internals pressure difference analysis is performed for the ICFR of ME00. The increased reactor internal pressure differences across the reactor internals are generated for the maximum core flow o.t normal, upset, emergency and faulted conditions as inpu; data for the fuel lift and the reactor internals impact evaluation to ensure the GGNS reactor internals ~

j can withstand the increased pressure differences.

l 15.D.8-1 l TCL:ge/F07239 3/24/85

15.D.8.3 Sundle U ft Evaluation

- The margin to fuel bundle lifIt was

~

t was reevaluated for the ICFR op shown that there is enough net normal, upset and faulted conditions. A probabilistic fuel fuel lift margin during the worst case faulted event.

lift analysis is performed which concluded the fuel lif t criteria increased core flow operation.

Iepact on Reactor Internals 15.D.8.4 The impact of increased core flow on the various reactor intern 15.0.8.2 are evaluated using the differential pressures discussed in Sectio i including and the forces generated by the probabilistic load combination an The reactor internals most fuel lift data discussed in Section 15.D.8.3. plate, affected by pressure under increased core flow conditions h d are the co guide tube, shroud head, upper shroud, shroud support ring a These compo-shroud top guide, fuel channel wall, steam dryer and jet pump.It ditions.

nents are evaluated under normal, upset, emergencynents and faulted co is concluded that the pressure differences for theseiand the other comp during increased core flow operation produce stresses that are allowable design limits.

~

I l

)

I 15.0.8-2 TCL:gc/FD7239*

7/23/84

35.D.9 Flow Induced Vibrations -

internals is To ensure that the flow-induced vibration response of the reactor acceptable, a single reactor of each product line and Af ter analyzing the size und

, during initial plant startup. ll within acceptable cxtensive vibration test results of such tests and assuring that all responses fais calssified as a v the established criteria, the reactor limits of All other reactors of the prototype in accordance with Regulatory Guide 1.20. t to assure same product line and size undergo a less vigorous conf similarity to the base test.The vessel internal vitiration startup test has been prototype BWR/6 251 plant. The results of the vibration completed for Grand Gulf 1 Station in the ICFR.The Grand Gulf 2 Station test results are documented in reference 15.D.15.8. similarity to the will undergo a less vigoruous confirmatory flow test to ensure base test.

l 1

i l

r t

l i

i e

1 15.0.9-1 TCL:ge:rf/F072310* '

Jl 3/14/85

Impac*. on Anticipated Transient Without scram (ATVS1 15.D.10 -

, in the ME00. The An ATVS perforinance impact evaluation was performed l d in for GGNS ATWS the ME00 region.

limiting Main Steam Isolation t Valve All peak (MS

)

l which resulted in a higher power cqndition following an ATWS eve Therefore, j

I pressures are below the emergency stress limit t i

it is concluded th:t MIDD operation is acceptable from ATVS r standpoint including ATVS stability considerations.

4 i

1 I

l

)

• I i l

! l t

I l

15.0.10-1 f TCt.:ge:rs/F072311*

7/23/84

_ _ _ - - _ _ _ _ _ _ _ ' "-"v-m ___ . , , _ _

15.D.11 fuel Mechanical Performance Evaluations were performed to determine the acceptability Compo- of operation on GE fuel red and assembly thetsal/sechanical fuel rod perfo ment pressure differentials (described in Section 15.D.8-4) and everpower values were determined for anticipated ope GNS MIDD initiated from MEOD conditions.

applied as the fuel rod and assembly design bases and there speration is acceptable and consistent with fuel design bases.

An evaluation was also performed which concluded i that fuel ch flow, creep and control blade interference are not tapacted by o in the MIDD.

15.D.11-1 TCL:ge/F072312*

7/23/84

Averane Power Range Monitor (APRM) 5teulated Thereal Po 15.D.12 Scram and Rod Block setpoint .

~~

In order to allow operation in the Maximus Power Extended Monitor Opera the current Average Power Range Monitor (APRM) Simulated (STPM) scram and rod block configuration and setpoints ar

~

date this region.

This consists of:

Raising the current APRM red block and STP scram line to hi (1) setpoints.

(2)

Clipping the APRM rod block at high core flow.

Increasing the High Flow Rod Block setpoint.

~

(3) llustrated in The new APRM rod block and STPM scram setpoints for th Figure 15.D.12-1 and tabulated in Table 15.D.12-1.

l (0.66),

The new setpoints presented in Table 15.D.121 maintain t d power the same clip setpoint (111% for STPM scram and l k setpoints 105 for as rod condition and same sargin (6%) between the STPM scram an the current 100% rod line Technical Specification. when f

block warning or scram protection exists due t operating in the MIDD. The increase of the high flow red block operation described in Appendix 15C. This is to setpoint from 10R to 111% core flow is one of operational y rod block c ensure operation in the ICFR will not result in too many unn alares.

15.D.12-1 TCL:ge/F072313*

7/23/84

l Table 15.0.12-1 APRP. Instrumentation Setpoint for ME00 Analytical Trip- Allowable Limit Setpoint Values _ _

Functional Unit Flow liased Simulated Thermal Power-Hign 0.66W + 70% with 0.66V + 67% a maximue of 0.66W

• 64% with a max.

a) Flow biased with a max. 114% of ratec of 111% of of.113% of thermal power b) Migh flow rated thermal rated thersal power clamped power 122% of rated 120% of thermal power 118% of rated thermal c) Neutron flux- rated thersal power high power Flow liased Rod Block 0.66W + 61%

0.66W

• 58% with a max.

a) Flow biased with a man. of 110% of of 108% of rated thermal b) High flow rated therr.a1 power clamped power Reatter Coolant System Recirculation Flow Rod Block i 114% of a) Upscale i 111% of rated flow rated flow l

l l

15.0.12-2 1CL:ge/F072313*

l' 7/23/84

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

F 1

. )gg 120 .

atu1 ROM Flux utstAtt SCRAM s

STP Scram ggg :

J e s stark 108'j 110 - ,

g FM

.66w

• 64 -

D g 100 . $

- J'

s0 .

.66v ,* 58 l _

80 . [ g 100: ROD LINE l

1.@

6 h

p h '.  !

l .

E 70 . 3 h E 1 l E-

" s 60 .

de y t 75: ROD LINE E

g /

/

{50 .  !

J i

g 40 .

[

J 30 .

20 .

CAVITATION REGION ~

~

10 .

- 'I a i

110 120 80 90 100 40 .50 60 70 10 20 30 .

CORI FLOW (t RATED) I ' J '"

l REFER 10 FIGURI 15531 FOR 0(FINITION O

\

II # ,,

15.D.1, MI55155]PPI APRM SETPOINT CONFIGURATION FOR MCOD ._

POWER 4 LIGHT

Feedwater Heater (s) Dut of Service in the Maximum 15.D.13 _ Operating Domain d

The section presents the results of the safety (FWHOS) evaluation during an for th operation of GGNS with feedwater heater (s) out of iservice domain annual initial cycle with GE-6 fuel in theinmaxim (MEOD) .

This section supplements the description in1 basis Appendix 158.

Figure 15.D.3-1.

The evaluation is performed for the GE fueled GGNS i on initial C The and is applicable to 12 month cycle GE6 fueled initial cycle opera l power conditions of operation are those of unlimited continued d t 1005 ther operation during the standard operation cycle with a maximu temperature reduction of 100'T in the MEOD. h ME00 between unlimited continued operation during an operation cycle in t e 420'F to 320'F feedwater temperature at rated power.

i d or All the inpact evaluations describedConclusions in Appendix made in 158 were reexam reevaluated in the MEOD for FWHOS operation. An operating MCPR limit l

Appendix 15B are directly applicable in the MEOD. feedwater of 1.18 is adequate for the range of 420'F to 370'F 1. Anrated operating temperature FWH05 operation in the ME0D for GGNS temperature at cycle MCPR limit of 1.19 is required for the range of rated feedwater of 370*F to 320*F FWHOS operation in the ME00.

in It is concluded that with the power dependent MCPR rate limits Appendix 15B and those described 1. in thh5 appendix for with FWH05 down to 320'F in the MCOD during cycle 15.D.13-1 TCL:ge:rf/F072314*

3/14/85

,gg,3,g, 39,,,

Mom am NOJ Abnormal coerating Transient 15.D.13.1 _

Two of the limiting abnormal operating transients (Load Reje Fal. lure and Feedwater bounding SWR 6 233 size 748 bundle Standard plant at both 7 Controller Failure) 1 % core were flow and verified for GGNS at about 110% core flow As described in Section 15.D.4.1. the core average power 0 condition.

density of the Standard BWR 6 plant is aleost identical to th The fuel. type used in the standard plant analysis represents bundle GGNS. The GGNS specific verification analysis was perforned the bounding nature. bounding for GGNS l

to cycle 1.

confirm the bounding analysis for equilibrium The GGNS evaluation was performed at 370*F andd 320'F this analysis.

feedwater torperatures at end of cycle 1 and 2000 MWD /T ex of cycle.

Section 15.B.2.1 describes the justification of these analy conditiens. The 75% core flow case was not evaluated fo t l

bounding BWR 6 analysis performed at both 320*F and 370 i

significant margin existed in this condition.

The GGNS plant specific transient peak value results are su The critical power ratio (CPR) results are Table 15.D.13.1 and 15.D.13.2. Tbv transient responses are sumarized in Table 15.D.13.3 and 15.0.13.4.

presented in Figure 15.D.13-1 to 15.0.13 8.

13-4 The GGNS specific analysis results shown in Table 15.D.1 l

controller failure transient at indicate that the ACPR for the feedwater 320*F rated feedwater temperature exceeds the current 0.01.

For GGNS cycle 1 appliceLtion, the technical specification i

described in Section 15.8.2-1 are to be applied to FWH E00.

That is, a rated operating limit MCPR value of 1.18 is to for the range of rated feedwater temperature down ature to 37 required for FWH05 operation in the range of rated f from 370*F to 320'F in the MOD.

rated MCPR , limits to support the elimination of AP requirement including operation with FWH05 in the ME00.

15.D.13-2

• TCL:ge:rf/F072314*

3/14/85

- - ~ _ _ _ *--~-r..-,, '---"-m-, - _ __

1 Table 15.0.13-1 I*I Sunumery of GCNS Transtent Peak Velves Results - FWHOS in ME00 - t0CI '

Peak Peak Peak Fdwtr J Peek vessel Steemline Dow Pressure Temp.

Neutron Pressure Pressure (psfg) ,[F)

Core Flow Flux (psig)

(psIg)

(% NBR] (X N8R} 373 Trenstent 1224 1195 I 162 1194 Lead Rejectien 109.6 with Bypass 373

' Failure 1175 1149 120.3 1150 Feedwater 109.6 Controller Failure, Men.

322 Demand 1218 1189 166 1188 lood Rejection liO.9 with Bypass

322 i

Feifure 1152 1127 II 125 1128 Feedwater 110.9 Controller Failure, Mem.

j Demand

, t e (e) Inittet power is 104.25 MBR for enelysts.(b) Meelman schleveble core flow .

I i

l i

15.D.13-3 ,

1CL: ge:rf /F072315*

S#14/A5

l . .

Table 15.0.13-2 II MWO/T Sefore EOCI

• Suomeer of Transient Peak Velve Results - FWHOS In ME00 2000 Peak Peak Peak Peak Fdwtr Dome Vessel Steamilne Neutron Pressure Tede.

Pressure Pressure Core Flow Flus {psig) {'FJ

(% NMR)(* (pstg) {pp}

Transient (% NBR)

-l 1189 373 l

1185 1216 Load Rejection 109.6 (b) 104.3 i

with Bypass Failure 373 1160 1135 109.6 II 118.2 1136 Feedwater Controller Failure, Mem.

Demand 322 1205 1184 l'10.9 II 104.3 1177 Leed Rejection f with typass l

' Failure 1111 322 IUI 1112 1137 110.9 124.3 Feedwater l

Controller Failure, Mem.

! Demend (a) Inittet power is 104.2% NOR for enelysts.

j 1

(b) Meuleum achievable core flow for the given feedwater temperature. .

1 15.D.13-4 .

TCL:gc:ef/Tb72315*

1/14/85

Table 15.0.13-3 Summary of CPR Results - FWHOS in ME00 - EOC1_

Fdwtr.

Temp Core Flow ICPR(h} ACPR MCPR Q ,

Transient (% NBR) 0.05 1.13 373 I 1.18 load mejection 109.'6 *I 1

with Bypass Fatture n.11 1.07 373 Feedwater 109.6I *I 1.18 Controller Failure, Max. Demand l 0.05 1.13 322 110.95 'I 1.18 Lood Rejection

^

with Bypass Fallure s 0.12 1.06 322 Feedwoter

- 110.9I *I 1.18 I Controller Fativre, 1 Max. Demand l

t be added.

(a) Manteum achievable core flow for the given feedwater temperature.(b) B NOTE: Option A adders included.

l t

\

15.D.13-5 TCL:gc:ef/F072315*

i 3/1a/95

i Table 15.3.13-4 i MWD /T Before EOC1 Susumery of CPR Results - FWHOS in ME00 - 2000 Feedwater Temperature Core Flow g MCPR (*F) i l

ICPR ACPR l (% MBR)

Transient 1.13 373 l I*I 1.18 0.05 109.6 f load Rejection with Bypass Failure 1.07 373 t

I*I 1.18 0.11 109.6 Feedwater Controller Failure Max.

Demand 1.13 322 I*I 1.18 0.05 e 110.9  ?

toed Rejection -

with typass l 322 Failure 0.13 1.06 l

110.9 (*} 1.19**

Feedweter Controller Failure, Max.

Demand temperature.

0.01 must be added.

2 (a) Montmum echieveble cere flow for the given feedwater(b) Desed o

• • Requires operating Ilmit CPR change.

i MOTE: Option A edders ineluded.

i i

15.D.13-6 .

TCL:ge: rf /F072315*

-...a.s

i 1

I NEUTRON PLUX l 2 PEm FUEL CENTER TEMP l

lI 3 RVE SURFF CE E RT FLUX y T [FwT1TP TEDH 150t  ;

5 VESSEL 51 ERM FLOW I

l

\

- 4:1-x & '

{' KR g N ~

b 5 -

- i I M l .

L' Y- ~ ' ,'

4. 6. 6.

0'O' 2. -

TIME ISEC) l . 1 LEVEll1NdH-REF-SEP-SKIRT .

2 W R SENSED 1.EVELilNCHE51 3 N R SENSE D i. EVE'. (INCHE51 200. 4 CORE INL F FL N E II 5 DRIVE FLtd 1 (PCT) 100.

E -

A a 1, y l

- x  !> 3 N ,3 i

0. ,

g. 6. B.

100. ''''2*

TIME ISEC)

FG LOAD REJECTION WITH BYPASS FAILURE.104.21 POW MISSISSIPPI POWER & LIGHT 373*F TFW E0*1 .

~

E

1 YESSEL PpES RISE (PSil .

2 STM LINE PRES RISE (PSIl

- 3 SAFETT VFLjrE FLOW (el 300. 4 RELIEF Vf LVE FLOW t.1 S BTPRSS VF LVE FLOW f.1 6 TURB STEF M FLON (PCTI 200.

~

A E s 100.

4 s '; u Et(5 4

0. n .3 m . , , I E35 8.

4. 6.

0. 2.

TIME ISEC1

. I V010 RERfTIVITY 2 DOPPLER F ERC1IV!TT 3 SCROM REICT1VITT I*

& iiT01h N fCTIvili

/

/

0. I O ' -

~

• (

n \

-1. ,

' ~

C .

W <

E  ! ( i

" ,,,,s... \

3. 4.

2,O. 1. 2.

r 11ME ISEC1 "m . _

FIGW LOAD REJECTION WITH BYPASS FAILURE,104.21 15.0PO MI551551PPI (Con POWER & LIGHT 373*F TFil EOC1 t ,

t

j 1 IEUTRON FLUX 1 j 2 PEAK FUEL CENTER TEMP 3 AVE SURF (CE WRT FLUX 150. iiTEEMIEr TE5E -

t 5 VE5SEL 51 ERM FLOW 100' 4 Y

h D E 9%

I s

- k fy 50. ,

l .  %

' ~

^

4. 6. B.

0 'O

• 2.

TIME (SEC) q 1 LEVELiINdH-REF-5EP-5KIRT ,;

2 W R SENSE D LEVELilNCHESI ' ,

3 N R SEN_S{ 0 LEVELIINCHESI 200. y (58T INLE l FLOW (PCT) 5 DRIVE FLC+4 1 (PCTI 100

d. - '

i 4 y

- w-5 i

12

0. ,

7

' 4

-100.' f'

2. 4. L. 8.

0* TIME ISEC)

FIGU LOAD REJECTION WITH BYPASS FAILURE.104.21 POWER,110.9% FLO M1551551PPI POWER & LIGHT 322*F TFW E0C1

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

. I VESSEL PP ES RISE (PSIl 2 SIM LINE PRES RISE (PSil 3 SAFETT Vi tVE FLOW - t.1 -

EO* -

it RCIEF M CVE FLOW 1 1 S BfPA55 VF LVE FLOW t.)

6 1 URB STEfM FLON (PC11 ,

200.

\- n 100.

3@.... 635 6 c4 63[ 4 O. 6. B.

0. 2. 4.

TIME ISEE)

/ 1 VnlD Rinfi lV11T

thll'11111 l 1W liVlli

I '.f itslH lifi l'11Vili -

I" vs - -

in fultit tti CITvTri V

/

O. ~ _- .

U -1.

b  :

W -

E  : *

'- 4.

1. 2. 3.

-2 0. TIME ISEC1 .

r!GURE LOAD REJECTION WITH BYPASS FAILURE,104.2% POWER,110.9

- 15.D.13-2 MI551551PPI 1 POWER & LIGHT 322 *F TFW E0C1 (Cont.) _

a

1 NEUTRON FLUX 2 PERh Flft CLNTER TEMP ,

3 AVE St#~s CC % AT E UX 150* -

u ii'lIC5iRil:r YCfC  ;

4 S VESSEL SI EnN FLOW .

1 g-- \

E 100.

t 6

E w 50.

E

~ ~

h 3 5 I

4 u .

1 u 1

0. ...#.... 20.

5. 10 . 15.

O.

TIME ISEC)

, I LEVELilNf H-RET-SEP-SKIRT 2 W R SfN5t 0 LEVfLilNCHE51 3 N R SENSI O l.fVilllNCHEbl 150* 4 00HE INil iTi.tW TPCT1 5 DRIVE FLt H I (PCTI J

u 100, 5 N -

/

,0 W .

N N" x

a

>j v s ~

a f

0. -

10. 15, 20.

D. 5.

! TIME ISEC)

FIGURE MISSI5SIPPI FEEDWATER CONTROLLER FAILURE.104.21 POWER.109.61 '

15.D.13-: FL POWER & LIGHT 377F TFW EOCl t

O 1 VESSEL Ptf.S R15E IPS11

[ j 2 SIM LINF Pff.S RISE IPSil

/2 3 TURBINE F AlS lil5E IPSIi -

125* -

3 s  % :on o un rsanarun.si S LIEF VIi.Yr. FLOW IPCTI 6T STD M FLOM IPCT) .

g _

)

~15.

N \

~

,y  % *

• 2S. ;

(t"2 -

i ;5 i

\1 .

10. 15. 20.

25.D. S.

TIME ISEC)

1V0ln,DCdTT1TiTT

.' lish f't t it I i lif'l l VIT T t'Si 1u

11 tti e i eiVili it 10id['htit 1TVID -

~

I* g J

/

" * - 3 .

0 ss

,-L- - ,'-

[ .l. ,

b .

W -

Y 4

_n t,,,a 1S. 20.

2,0. S. ID.

TIME ISEC)

FIGURI FEEDWATER CONTROLLER FAILUPI,104.25 POWER,109.61

• FLO 15.D.13 M1551551PP) -

.(Contd POWER & LIGHT 37 rF TFW E0C) .

I NEUTRON PLUX .

. 2 PERK FLCl CENTER TEMP 150. ,

y3 RVE 7 g *Infi

,ggCE-

-glERT FLUX S VESSEL $f CRM FLOW ia 5

w 100. _

5 --

L b

I5' I

s A s s,

h*

\ 1 u 1 11 4

D.

0. ,,,,,,,,,5. 10 15. 20.

TIME ISEC) ,

1 LEVELtINCH-REF-SEP-SKIRT 2 W R SENM O LEVELilNCHESI 3 N R SfN'tIl LEVEL. IINCHE51 150. iFCORE 1 il T TD w'~TPCT) 5 DRIVE ILtd 1 (PC1) 8J S

100.

(

50 l W

• s 5

~

10. 15. 20.

0'O. 5.

TIME ISEC) r!GURE FEEDWATER CONTROLLER FAILURE,104.2% POWER,110.9% FLOW. 15.0.13-4 MIS $155]PP]

POWER & LIGHT 322 *F TFW EOC1

s 1 - f

' I N,1 VESSEL Pf CS niSE (PS11'RES RISE  :

IPSI)

STM L INE i -

> 1 UI M' ' 'uS nlF IPSIi i 4 ' O kti f WTil llTU/LBI 125* ~

5 IEF W t.YE FLOW IPCT) J

'l 61 STf.itt FLOW IPCTI o

_E --

M. N 25.

6 5 s sS

- l2L" 3 5 5

-25. ~ - 1.- 15. 20.

S. 10.

D. TIME ISEC) 1 vn1TT.TJ 6 G lTY N ini t i t.i; i ffir I iv1T T

/  ; ..i I:tif t I:t i - 1Ivi11 -

I 1t11:1L"l;:: c1TVIY1 I* ,/ l I

sy .

mA_s g* _1 A17 - Tlf G '-

- /

U g

-1* *

( p l -

...'.... 1 15, 20.

5. 10.

2.D. TIME (SEC) .

~

FIGURE FEE 0 WATER CONTROLLER FAILURE,104.2% POWER,110.9% 15.D.13-4 F MI553551PPI .(Cont.) _

POWER & L194T 322 *F TFW E0C1

~

1 NEUIRON PLUX 2 PERK FUfL CENTER TEMP 3 RVE SUR'f CE HER1 FLUX ,

150. 4 FEEDWAlO D

. 5 VESSEL 51 ERM FLOW .

100.

e .

E s -

w g 'I  %

s- -

50. ,

e-

' ( t 1 de

• E- ' '

y* 6. 8.

O 'O .' 2-TIME ISEC) 1 LEVELtINtH-REF-SEP-SKIRT 2 W R SENSE D LEVELtINCHE51 3 N R SENSE D LEVEL tlNEHE51 203. y CORE INT ( i FLO.4 (PCT) 5 ORIVE FLtu 1 (PCTI d

2 -

RM . <

1 , s u

s _.

f r

& q 11 D. '

~

, _1 100. y* 6. 0-D. 2-TIME ISEC) l F1 GENERATOR LOAD REJECTION WITH BYPASS FAILURE.104.2% POWER ,

hfR k GHT 109.6% FLOW. 373*F TFW EOC1-2K W D/T

1 VESSEL PfES RISE IPSil 1 2 SIM LINE PRES RISE (PS11 J

. 3 stwrTi vr LVE FLOW tel _.

300' -

ii MDU75 cvt rLoW I.)

5 DTPRSS Vf t.YE FLOW tel 61 URB 51LF M FLOW iPC11 .,

200.

100. 4 Y E Ru E 5 .

3 t& . . . E3 5 D.

2. 4. 6. 8.

O.

TIME ISEC) i l Vnlit ntndTIVITT

.'th1111.111 I i, lit.11 VI T T i- E11071 tirl J11Vili -

I* ,

ii lo H1. liLI i 11V'11'T Af

0. 12 __

T' E

t M'

-1. ,

E y _

E i *

-2. i -

4.

1. 2. 3.

0. TIME ISEC)

O FIGUR MI551551PPI mtR i u=T GENERATOR LOAD REJECTION WITH SYPASS rAILURE.104.21 io...s rt0W. 373 r Tru E0Ci.rx o,T g;;;.p,PO

l O

l 1 NEUTRON PLUX 2 PEAK FUEL CENTER TEW .

. 3 RVE SURFG CE K RT FLUX 1 150. 4 FEEDWRIEF FLOW 5 VESSEL S1 ERM FLOW ,

100.

E

• [ M N

b  %

• 5 5 50. ,

E  : Il  %

i [

~

, 3 - 44 M =

0. - -

4. s e.

O. 2.

TIME ISEC)'

1 LEVELi1N H-REF-SEP-SKIRT 2 H R SENS 0 LEVELIINCHESI 3 N R SENS 0 LEVELIINCHES) 200* y EORE int.E l FLOW (PCT) 5 DRIVE FLtd 1 (PCTI d -

100' N' 'M N n u y

.5 3 1

m _M M

= - v D. ,

L

. a

~

-100. 4. 6. 8. .

D. 2.

TIME ISEC)

.C.

FIGURE MIS $1551PPI LOAD REJECTION WITH BYPASS FAILURE,104.2% POWER,110.9% FLO 15.D.13-6 POWER & LIGHT 322*F TFW EOC1-2KIWD/T _

,w ,

,, - - -, ,,- _- - - - y- ,c -w e

I VESSEL PPES RISE (PS11 i

. 2 STM LINE PES RISE (PSI) . I 3 SRFETY vf LVE FLOW tel --

300. 4 E LIEF Vf LVE FLOW (*I 5 BTPASS vf LVE FLtM (*)

6 TURB STEF M FLOW (PCT) 200.

100.

x v_r #,

m3 kU 8 5

0. M ' ... E3 5 8.

2. 4. 6.

O. ,

TIME ISEC) 1 VOID REAC T1VITT T'

2 DDPPLER F CACTIVITY 3 M HnN IVf CTivlTT 3* 1 4 10TAMICT 1V11 T 1

0.

J 3

I

{ -1. .

E 7

4

....f. .. ) 4.

2,0. 1. 2. 3.

TIE ISEC)

FIGURE LOAD REJECTION WITH BYPASS FAILURE.104.2% POWER,110.9% 15.D.13-6 MI551551PP3  ;

POWER 8 LIGHT 322*F TFW EOCl-2K MWD /T (Cont.) __

I NEUTRON f LUt 2 PERK FliEL EENTER TEMP 3 RVE $tlRFG CE MERT FLUX 150, IITTEDwelEP FLDW -

4 g 5 VESSEL 51 EAM FLOW

- b 100.

\

g*

E I ,

w 50.

W  :

N '

Y -

N k1 -

ul ut.

0.'

5. 10. 15. 20.

O.

TIME ISEC) 1 LEVELIIN H-PLEF-SEP-5MIRT I 2 W R SENSED LEVELilNCHESI 3 N R SEN;L D LEVFL f!N:MES) l

150. 4 CORE INR 1 FLOW IPCT) l 5 DRIVE FLl w 1 (PCT) l 5

100.

\ A m

50.

%w M.

s a

D. l 1 -

15. 20.

D. 5. 10.

TIPE ISEC) .

FIGURE FEEDWATER CONTROLLER FAILURE,104.21 POWER,109.61 FLOW,

• l M1551551PPI 15.D.13-7 POWER & LIGHT 373*F TFW EOC1-2K MWD /T

~ SSEL NES RISE (PSit 2 LINE PRES RISE (PSit 31 BlNE f RES RISE (PS11 -

125.

-. yC ~~TT 1N F50tT'IETu/LB1

- 5 REL) F Vf LVE FLOW (PCT)

M FLOW (PCTI E -

5 N 6 1 URB , TEF

- h ., (\

%- [

W I h

[f f N 25.

!r 12 -As a s e \ \

25.'''''5.

O.

10.

TIME ISEC)

15. 20.

r 1VD1DREN TIV

? DITT 7.l;4 i. 6 IV1TT 7 %rn 1ITtu11VIT1 A- i lNi)T~.iiL h' 6 .')Tv'llT 1.

[

, g-l,jf -

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I

{

w

( e 1 #

I E

j w  :

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-2. * *

• 1 * ' ' ' 5- 15. 20.

10.

u nE iSEci .

FIGURE MI553551PP1 FEEDWATER EONTROLLER FAILURE.104.21 POWER,109.6% FLOW, 15.D.13-1 POWER & LIGHT 37 3*F TFW EOC1-2K PWD/T (cont.)-

L

I 1 NEUTRDN f LUX 2 PERK FLEI CENTER TEMr

' 3 RVE Surfs cE Mtni FLUX -

150* u- 4 F EEDMID TGM 5 VE55EL $1 tan rLow

~~ 100.

.B La '

5 E

b '

E "'

d lig  :

! N ,

s 9

~ '

i 4

ut ui.

,,..,, \1 D.~ 10. 15. 20.

O. 5.

U ME.15EC3 1 LEVELi1NT4-REF-5EP-5KInT 2 H R SEN!lL1LEVELIINCHUit 3 N R SEN7.I in I.FVFL litCHESI iFtont Itli ITLWMT )

150.

5 DRIVE FL( W l (PCTI

^

100. .

~

M PXND%

"' - l l

' ' E p

l

* 'I

- 15. 20. l D'O. 5. 10. l

-U ME.15EC)  ;

I I

l .

l i

FIGURE FEEDWATER CONTROLLER FAILURE,104.2% POWER,110.9% FLOW, M155155]PPI 15.D.13-1 I

POWER & LIGHT 32 TF TFW E0tl-2K WD/T

i -

! VESSEL PfES PISE (P511 2 STM LINI PRCS MISE (PSI) 3 TURDlNI I aW.5 R15 IPSII

~ I * -

r cone li(1 1~!iGt"~t U/LDI '

f LIEF VF LVE FLOW IPCil

- I 6 STEntFLOW (PCT 1 E .

i I

'75. k h u

D. ,

g

~

l

wn s a  !

N' s

~

'- 20.

5. 10. 15.

45 0. TIME ISEC) g yn;p yee' :ui.Ti -

. g. 1 i.: , i il *1TT

. .: .l:: ..)

~

I e .i i!VI**

. t ' '. 1". "'

I+ 9 / ]" 1 ]!i'i "l l

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0 N ,L -- f

• 1 u #

h -1. ,

C  :

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10. 15. 20.

4'O. 5.

TIME 15EC)

FIGURE FEEDWATER CONTROLLER FAILURE,104.21 POWER,110.9% FLOW,15.D.13-8 M1551551PP]

, POWER & LIGHT 3gg.*FTFW E0C1-2K M/T (Cont.)

g_

t

~

15.D.13.2 Other Evaluations All other evaluations described in Appendix 158 are directly applicable to The 100*F Loss of Feedwater Heating is not affected by ,

FWHOS in the MEOD.

The MEOD because the generic study applies to all power flow conditions.

rod withdrawal error analysis described in Section 15.B.2.2 is directly applicable to the MEOD because the generic RWE analysis is performed based on the ME00. As described in Section 15.B.3, the compliance to stability General Design Criterion (GDC) 12 is demonstrated because the generic analyses of reference 15.D.15.5 are independent of stability margin since This the reactor is already assumed to be in limit. cycle oscillation.

implicity covers any variations in stability ma,gtns caused by FWHOS operation. Section 15.D.5 also provided ju;tification for compliance with GDC 12 in the ME'D0 region. It is therefore concluded that stability criteria are met for FWHOS in the MEOD because con 61 nation of FWH MEOD does not result in any changes in parameters affecting the analysis of Reference 15.D.15.5 as the analyses already assumed limit cycle oscilla-tion. Loss of Coolant Accident responses are shown to be bounded by Chapter 6 analysis. Containnent analysis and acoustic / flow induced load analysis described in Section 15.D.7 and 15.0.8 are analyzed with reduced feedwater temperature associated with final feedwater temperature reduction Feedwater (FFWTR) operation in the MEOD and are shown to be acceptable.

nozzle, sparger fatigue and system piping are all independent of power / flow condition. Impact on ATWS. annulus pressurization loads and fuel duty are '

t all shown to be acceptable for FWHOS operation in the ME00.

l TCL:ge:rm/F072317* 15.D.13-7 7/23/84 .

f

\

Elimination of the APRM Trip setdown Reautrement 15.D.14 l .

The GGN5 Technical Specifications require that the flow biased APRM trips be Iowered (setdown) when the core maxh.an total peaking factor exceeds the design The *APRM setdown" requirement oriG inated from the total peaking factor.

absolete Hench-Levy Minimum Criti?al Heat Flux Ratio (MOHFR) thermal limit celtarion.

The change to GETAB/GEXL and the move to secondary reliance on flux scram Itcensing transient evaluations (for transients terminated by anticipatory or direct scram) has provided more effective and operationally acceptable alterna tives to the setdown requirement. The GGNS ME0D evaluation uses transient analyses to define thermal limits initial conditions (operating limits) which conservatively assure that all licensing criteria are satisfied without setdown of the APR*. scram and flow biased rod block trips.

The objective of this evaluation is to , justify removal of the peaking facter Those licensing areas which might be affected by the setdown requirement.

elieination of the setdown requirement are:

i

a. fuel thersal-rechanical integrity, and

b. loss-of-ccolant accident.

The following criteria assure satisfaction of the applicable licensing ments and were applied to demonstrate the acceptability of eliminati

. setdown requirement.

s.

MCPR safety limit shall not be violated as a result of any abnorma operating transient, b.

All fuel thermal mechanical perforsance shall remain within th licensing bases, and l

c. peak cladding temperature and saximum cladding oxidatio ing a LOCA shall ressin within the limits defined by the applic regulations.

15.D.14-1 TCL:gc/F072318*

7/23/84

- - - -~

The safety evaluations therefore include abnormal operational transients and LOCA analysis. -

a 15.D.14.1 Transient Evaluation ,

l A large data base was used to stu6 the trend of transient severity without the '

This data base average power range monitor (APRM) core peaking factor setdown.

was established by analyzing limiting t;ansients over a range of power and flow l conditions and was used to develop plant operating limits (MCPR and MAPLNGR) which will assure that margins to fuel integrity limits are equal to or larger than those in existence at the present t!ae. .

Results from the above transient analyses were used to establish the MAPLHCR versus power and flow and to verify or establish the MCPR versus power and flow Technical Specification. A variety of OGN$ specific Feedwater Controller

~

Failure and Load Rejection With Bypass Failure Transients with and without Feedwater Heater (s) Dut of Service (FVHDS) and Final Feedwater Temperature Reduction (FFV7R) together with a bounding BWR/6 analysis and a large data ba of operating plants results were used to assure that suitable conservatise.

exists for operation in the M10D with FFVTR and FWH35 mode of operations.

Results of the bounding BWR/6 analyses are tabulated in Table 15.D.14-1 Results of the GGNS specific analyses are tabulated in Table 15.D.14-2.

15.D.14.2 Less of Coolant Accident The impact of the elimination of the APRM setdown requirement on LOCA It is found that the

- examined in the maximum extended operating domain.

current MAPLNGR limits are adequate to protect against a Loss of Coolant Accident even without APRM set 6wn to assure that peak cladding temperat This is because the LOCA remain below the standard LOCA result;s in Chapter 6.

analysis documented in Chapter 5 is performed without taking credit APRM setdown.

e 15.D.14-2 TCL:ge:rs/F072318*

1 7/23/84 w ~ - - - - - - - , - - - - - , - - - - ---

15.D.14.3 Plant Operatina Limit

- power and flow dependent limits on local peak power and EPR are imposed that fuel design and safety criteria'are satisfied without the peaking factor setdown.

Flow Dependent MCPR Lielt 15.D.14.3.1 i

The current flow dependent MCPR limits remain unchanged because the basis flow runout event is a slow flow / power increase event which is not terstnated by scram. .

15.D.14.3.2 Power Dependent M:PR Liett The current power dependent M PA 1f aits are modified based on resu

~ GGNS specific analysis and prior bounding BWR/6 analysis as well as a large operating plant data base to include:

A new set of limits including a new limit format for core power be (a)

(power level where reactor scram on turbine control valve fast cl bypassed) which consists of both high and low core flow depende dependent EPR Limit due toThis theset significant sensitivity to initial c of limits also apply to flew below this bypass power level.

Feeewater heater (s) Dut of service (FWH05) and final fe

• reduction (FNTR) operation.

. (b)

A new KPR sultiplier limit p (K ) is established for core power a to replace the absolute power dependent E PR, limit to eliminate i complication of having several EPR, curves for different m A generic Kp curve is to be applied to different rated operating limits above 40% power. , 1 These new power dependent EPR limits p and multiplier (E )

Figure 15.D.14-1. .

15.D.14-3 TCL:gc/F072318*

7/23/84

4 15.D.14.3.3 Flow Dependent MAPLHGR Lisit

~

The flow depeMant MAPLNGR limits were determined using the three-dimensional f

l SWR simulator (Reference 15.D.15-3) to analyze the slow flow runout transients.

These factors are derived such that the peak transient MAPLHGR during these The flow dependent events is not increased above the fuel design basis values.

MAPLHGR factor (MAPFAC ) limit is shown in Figure 15.0.14-2. The actual flow p l dependent MAPLHGR limits are equal to this MAPFAC p multiplied by the rated MAPLNGR limit.

15.D.14.3.4 Power Dependent MAPLMGR Limit -

In the absence of the APRM scrar. setdown requirement, special limits are l substituted to assure adherence to the fuel thermal mechanical design bases.

Power dependent limits are generated using the same data base as the power

~ dependent E PR Lietts. As previously discussed under MCPR, p (K ) limit, a significant sensitivity to initial core flow exists below 40% core power.

These Therefore, a set of both high and low core flow limits is provided.

liatts are derived to assure that the peak transient MAPLHGR is not increased above the fuel design basis transient values. Appropriate MAPLHCR(p) limits are selected based on GCNS specific and bounding BWR/6 transient analyses The new power dependent and trends observed in the operating plant data base.

The actual MAPLNGR factor (MAPFAC,) liett is presented in Figure 15.D.14-3.

power dependent MAPLHGR limits are equal to this MAPFAC, multiplied by the For single loop operation (SLO), the most restrictive of rated MAPLHGR limit.

the 5LO MAPLHCR factor and this MAPLNGR factor will define the limiting

. of operation.

15.D.14.3.5 Governino Overall Limit ,

The most At any given power / flow state, all four limits must be determined.

limiting MCPR and the most lietting MAPLHGR (maximum of MCPR, (X, and minimum of MAPLNGR,and MAPLMGR 9 ) will be rated MCPR limit) and MCPR g The rated operating limit MCPR value for the different modes governing limit.

of operation are presented in Table 15.D.14-3.

15.D.14-4 TCL:gc/F072318*

_ _ ___7/23/84- ---- - - _ ._ __ _ __ _ _

w

Tabl4 15.D.14-1 Bounding BWR/6 Transtent Analysis Results

~

~ For Elimination'of APRM Trip 5etdown d

Power. Flow Scree" gPR*

Case e Transient' A- E MCPR, LS .15 1.21 FW:F - W/RPT' 100 111.7 1

L8 .39 1.45 53.5 116 2 FW:F - W/RPT' L8 44 1.50 40 92 3a FW:F - W/RPT' L8 .47 1.53 40 92 Sb FWCF - N/RPT' PR .99 2.05 40 92 3c LRNBP - N/RPT L8 .19 1.25 25 64 4a FWCF - N/RPT' PR 1.04 2.10 25 64

~

4b LRNEP - N/RPT LB .13 1.19 40 50 5a FV:F - N/RP1' PR .71 1.77 40 50 Sb LRNBP - N/RPT L8 .16 1.22 25 50

&a FW:F - N/RPT' PR .92 1.98 25 50

- 6b LRNEP - N/RPT FOOTNOTES:

a W/RPT = With Recirculation Pump Trip N/RPT = No Recirculation Pump Trip b L8 = High Water Level 8 Scram PR = Pressure Scram c ODYN Option A adder included j

! d based on SLMCPR = 1.06 NCPg = SLMOPR

• ACPR e

with feedwater temperature reduqtion of 170'F, i.e. , TFW = 250'F 15.D.14-5 TCL:ge/F072318*

7/23/84

Table 15.D.14-2 Grand Gulf Transient Analysis Results

~

For Elimination of APRM Trip Setdown d

Case Powed A

Flow E Scree b g ,C,,

Transient

• 2 108.6 L8 .13 1.19 M F - W/RPT' 100 1 1.23 110 LS .17 FWCF - W/RPT' 85 2 1.29 LB .23 70 112 3 M F - W/RPT' '

LS .38 1.44 46 114 4 M F - W/RPT' L8 .39 1.45 40 105 5a FVCF - W/RPT' 40 1.45 105 LS 40 5b FWCF - N/RPT'

.89 1.95 105 PR

~ 40 5c LRNSP - N/RPT 1.01 2.07 72 PR 25 6 LRNBP - N/RPT

.62 1.68 50 PR 40 7 LRNBP - N/RPT

.83 1.89 50 PR 25 8 LRNBP - N/RPT FOOTNDTES:

a W/RPT = With Recirculation Pump Trip N/RPT = No Recirculation Pump Trip b LS = High Water Level 8 Scram PR = Pressure Scram c ODYN Option A adder included d based on SLMCPR = 1.06 MCPg = SLMCPR + ACPR 250*F e

with feedwater temperature reduction of 170*F, i.e. , rated TFW =

d m

15.0.14-6 TCL:ge/F072318*

7/23/84

Table 15.D.14-3 .

Rated Operating Limit MCPR Values Rated

• P>de of OLMCPR byeration Current FSAP. P/F Map (Fig. 4.4.5) 1.18 FWHOS (Appendix 15B) 1.18 (420*F to 370'T up to 100% flow) 1.19 (370'T to 320'T up to 1001 flow)

SLO (Appendix 15C) 1.18 MEOD (Appendix 15D) 1.18 FW"5 in MEOD (Appendix 15D) 1.18 (420'F to 370'F.

up to 1051 flow) 1.19 (370*F to 320'F. up to 1051 flow)

• These values are for cycle 1 only. These values are to be applied with the X curve for off-rated conditions above 401 power.

p MCPR,(p) = Kp (p)

• OLMCPR rated.

• • All evaluations and results are limited to GE6 fuel used in operating strategies with the target Haling end of cycle exposure distribution.

Various modes of spectral shift operation in which the cycle average void distribution significantly exceeds that obtained with a Haling strategy can violate the validity of these limits.

TCL:gc:rf/F072318* 15.D.14-7 10/15/84

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. 15.D.15 References .

15.D.15-1 ~" Qualification of the One-Dimensional Core' Transient Model for Boiling Water Reactors" MEDO-24154 Oct. 1978.

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15.D.15-2 R.B. Linford " Analytical Methods of Plant Transients Evaluations for the General Electric Boiling Water Reactor" NEDO-10802 Apirl 1973.

15.D.15-3 "Three Dimensional BWR Core Simulator" NEDO-20953-A, January 1977.

15.D.15-4 Letter, J.S. Charnley (GE) to F.J. Miraglia (NRC), " Loss of Feedwater Heating Analysis", N1y 5, 1983 (MFN-125-83).

15.D.15-5 " Compliance of the General Electric Boiling Water Reacator Fuel Designs to Stability Licensing Criterin' NEDE-22277-P, December 1982.

15.D.15-6 "Ger.e ral Electric Standard Application for Reactor Fuel" NEDE-20411-P-A, January 1982.

15.D.15-7 "BWR Core Themal Hydraulic Stability" Service Information Letter, SIL 380 Revision 1. February 1984.

15.D.15.8 " Grand Gulf-1 Reactor Internals Vibration Measurements Sussnary i Report" NEDE 31148P February 1986.

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• 7/24/84

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