Text

Proposed Tech Specs,Changing Setpoints and Limits Re Recirculation Loop Operations to Allow Single Loop Operation
	ML20148R648
Person / Time
Site:	River Bend
Issue date:	04/06/1988
From:	; GULF STATES UTILITIES CO.
To:	;
Shared Package
ML20148R646	List: ML20148R648; ML20148R664; RBG-27657, Application for Amend to License NPF-47,changing Tech Specs to Revise Setpoints and Limits Re Recirculation Loop Operations to Allow Single Loop Operation.Fee Paid;
References
	NUDOCS 8804150014
	Download: ML20148R648 (32)
	v • d • e

.

n c.u di v E D OCT 2 91987 2.0 SA/ETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SDC 2.1 SAFETY LIMITS THERMAL POWER, low Pressure or low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than '85 psig or core flow less than 10% of rated flow, be in at least HOT SHUTOOWN whoin 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1. with two recirculation loop operation and shall not be less than 1.08 with singl recirculation loop operation THERMAL POWER, High Pressure and High Flow 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.07"with the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

with two recirculation loop oper:Jtion or less than ACTION:

1.08 with single recirculation loop operation With MCPR~1ess than 1.0 and the reactor vessel steam dome pressure greater than or equal to 785 psig and core flow greater than or equal to 10% of rated flow, be in at least HOT SHUTOOWN within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the require-ments of Specification 6.7.1.

REACTOR COOLANT SYSTEM PRESSURE 2.1.3 The reactor coolant system pressure, as measured in the reactor vessel steam dome, shall not exceed 1325 psig.

APPLICABILITY: OPERATIONAL CONDITIONS 1, 2, 3 and 4.

ACTION

With the reactor coolant system pressura above 1325 psig, as measured in the reactor vessel steam dome, be in at least HOT SHUTDOWN with reactor coolant system pressure less than or equal to 1325 psig within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months and comply with the requirements of Specification 6.7.1.

l l

RIVER BEND - UNIT 1 2-1 Amendment No.

8804150014 800406 PDR P

ADOCK 05000458pcp

i

. TA3tE 2.2.1-1 E

E REACTOR PROTECTION SYSTEM INSTRtMENTATION SETPOINTS z

ALLOWABLE o FUNCTIONAL UNIT TRIP SETPOINT VALUES ,

E 1. Intermediate Range Monitor, Neutron Flux-High y $ 120/125 divisions < 122/125 divisions g 2. Average Power Range Monite :

of full scale of full scale 3 Neutron Flux-!'ign, Setdown < 15% of RATED < 20% of RATED THERMAL POWER THERMAL POWER

b. ilw Sie;ed Sisletu ik. el 7--er lligh

'*""^"

1) 71 w tiesed : 0.00 ":40%, with -' O.55 9 51v, eith 5 ='r= cf m -; w ef M Gh F10% El W ;d U.ntnrm rv-onll $ ^.I "TED ' i.,lM O' PJIE0 -

m m ,o-on m c. Neutron Flux-High 5 118% of RATED 1 $ 120% of RA"TED THERMAL POWER THERMAL POWER

d. Inoperative NA NA

3. Reactor Vessel Steam Dome Pressure - High $ 1064.7 psig $ 1079.7 psig

4. Reactor Vessel Water Level - Low, Level 3 > 9.7 inches above > 8.7 inches above instrument zero* instrument zero

5. Reactor Vessel Water Level-High, Level 8 < 51.0 inches above < 52.1 inches above instrument zero* instrument zero

6. Main Steam Line Isolation Valve Closure < 8% closed < 12% closed

7. Main Steam Line Radintion - Hi g.'- . < 3.0 x full power < 3.6 x full power background background E .

g 8. Drywell Pressure - H!gt $ 1.68 psig $ 1.88 psig

9. Scram Discharge Volume Wats level - High I a. Level Transkitter - LIS!;601A and B < 49 < 52" p LISN601C and D {49" { 51.7" g b. Float Switci:es - LSN013A and B < 48.76" < 53.50" LSN013C and D 346.88" 3 49.00" E !TijureB3/43-1.

"A" Insert

b. Flow Bissed Simulated Thermal Power-High

1) Two Recirculation Loop Operation ,

a) Flow Biased S 0.66 W+48%. with 1 0.66 W+51%. with a maximum of a maximum of b) High Flow Clamped 5 111.0% of RATED $ 113.0% of RATED THERMAL POWER THERMAL POWER

2) Single Recirculation Loop Operation a) Flow Biased < 0.66 W+42.7%, with < 0.66 W+45.7%. with a maximum of a maximum of b) High Flow Clamped < 111.0% of RATED < 113.0% of RATED THERMAL POWER THERMAL POWER J

'l

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

- a .- ..

RECEIVED 2.1 SAFETY LIMITS M 291987 ,

BASE 5 l

2.0 INTRODUCTION

The fuel cladding, reactor pressure vessel and primary system piping are the principal barriers to the release of radioactive materials to the environs. '

Safety Limits are established to protect the integrity of these barriers during normal plant operations and anticipated transients. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Because fuel damage is not directly observable, a step-back Insert B a)proach is used to establish a Safety Limit such that the MCPR is not less insert a ,

i tian 1.07 4 MCPR greater than 1.074 represents a conservative margin relative to the conditions required to maintain fuel cladding integrity. The fuel cladding is one of the physical barriers which separate the radioactive materials from the environs. The integrity of this cladding barrier is related to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission product migration from this source is incrementally cumulative and continuously measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design conditions and the Limiting Safety System Settings. While fission product ,

migration from cladding perforation is just as measurable as that from use related cracking, the thermally caused cladding perforations signal a thres-hold beyond',whi:h still greater thermal stretses may cause gross rather than incremental cladding deterioration. Therefore, t% fuel cladding Safety limit is defined with a margin to the conditions which (ould produce onset of transi- ,

tion boiling, MCPR of 1.0. These conditions reptu ent a significant departure

+

from the condition intended by design fer planned operation. r 2.1.1 THER'4AL POWER, Low Pressure or Low Flow 2 The use af the GE Critical Power correiation (Reference 1) is not valid for all critical power calculations at pres.:ures below 785 psig or core flows less than 10% of rated flow. Therefore, the fuel eiadding integrity Safety Limit is established by other me.*ns. This is done oy establishing a limiting condition on core THERMAL POWER with Lt.0 following basis. Since the pressure d.op in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater d an 4.5 psi. Analyses show that with a bundle flow of 28,000 lbs/hr, bundle pretsure drop is nearly independent of bundle power and has a value of 3.5 pss. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28, J00 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 806 psia indicate that the l l

fuel assembly critical power at this flow is approximately 3.35 MWt. With the '

1 design peaking factors, this corresponds to a THERMAL POWER of more than 50% of '

, RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER j for reactor pressure below 785 psig is conservative.

t B 2-1 Amendment No.

RIVER BEND - UNIT 1

I I

Insert "B" I

for two recirculation loop operation and 1.08 for single recirculation l loop operation.

(

e 1

l I

l l

I 1

I

tt h C E I V E D 3/4.2 POWER _ DISTRIBUTION LIMITS OCT 291987 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE SDC LIMITING CONDIT10'N FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type of fuel as a function 3. 2.1-1,of3.AVERAGE PLANAR 2.1-2, 3. 2.1-3, EXPOSURE

3. 2.1-4, 3. shall not 2.1-5, and 3. 2.1-6. exceed

<q__ l t shown in Figures OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or APPLICABILITY:

equal to 25% of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 or 3.2.1-6, initiate corrective action within 15 minutes and restore APLHG3 to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

The limits of figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1-6 shall be reduced to a value of 0.84 times the two recirculation loop operation limit when in single loop operation.

SURVEILLANCE REQUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal t'o or less than the limits

3. 2.1-1, 3. 2.1-2, 3. 2.1-3, 3. 2.1-4, 3. 2.1-5 and 3. 2.1-6 :

determined fr.om Figures

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

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

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

d. The provisions of Specification 4.0.4 are not applicable.

3/4 2-1 Amendment No. fq RIVER BEND - UNIT 1

POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 1

3.2.2 The APRM flow biased simulated thermal power-high scram trip setpoint (S) and flow biased neutron flux-upscale control rod block trip setpoint (SRB) shall be established according to the fo11 ewing relationships:

"tIF OC"0!'" n'.~dm4-VAWE.

Insert C 4 : : (;, :g : ;); ;: 0,0:e - g);

'0. J ^ i2)T

&{_:'O.?!U-431T '?](:

where: S and S RB are in percent of RATED THERMAL POWER, W = Loop recirculation flow as a percentage of the loop recirculation flow which produces a rated core flow of 84.5 million 1bs/hr.

T = The ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the CORE MAXIMUM FRACTION OF LIMITING POWER DENSITY (CMFLPD). T is applied only if less than or equal to 1.0.

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

ACTION:

Mith the APRM flow biased simulated thereal power-high scram trip setpoint and/or the flow biased neutron flux-upscale control rod block trip setpoint less conser-vative than the value shown in the Allowable Value column for 5 mr SRB' ** * **

determined, initiate cor'rective action

within 15 minutes and adjust 5 and/or S RB to be consistent with the Trip Setpoint value

within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months or reduce THERMAL.

POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.2.2 The FRTP and CMFLPD shall be determined, the value of T calculated, and the rest recent actual APRM flow biased staulated thermal power-high scram and flow biased neutron flux-upscale control red block trip setpoints verified to be within the above limits or adjusted, as required:

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

^

b. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after complet1on of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operat-ing with CMFLPD gre u er than or equal to FRTP.

d. The provisions of Specification 4.0.4 are not applicable.

With CMFLPD greater than the FRTP, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that the APRM readings are greater than or equal to 100% times CMFLPD, provided that the adjusted APRM. reading does not exceed 100% of RATED THERMAL POWER, and a notice of the adjustment is posted on the reacto,r control panel.

RIVER BEND - UNIT 1 3/4 2-7

a Insert "C" !

i

a. Two Recirculation Icop Operation Trip Setpoint Allowable Value !

S '6- (0.66W + 484)T S 6 (0.66W + Slt)T [

Sg 5 (0.66W + 42%)T Sg 6 (0.66W + 45%)T

b. Single Recirculation Icop Operation Trip Seteoint Allowable value S $ (0.66W + 42.7%)T S 6 (0.66W + 45.7%)T ;

S RB f (0.66W + 36.7%)T S RB (0.66W + 39.7%)T :

P i

, i t

l l

l 1

e I

i j J

r 1

J l

__ . ____- - - - . _ = _. _--

f TABLE 3.3.6-2 CONTROL ROD 8 LOCK'INSTRUNENTATION SETPOINTS es TRIP FUNCTION TRIP SETPOINT ALLOWA8LE VALUE E 1. 1100 ?ATTERN CONTROL SYSTEM e a. Low Power Setpoint 27.513% of RATED THERMAL POWER 27.5 1 7.5% of RATED THERMAL POWER

b. High Powe.- Setpoint 62.513% of, RATED THERMAL PCMER 62.5 1 7.5% of RATED THEllMAL POWER

.- 2. APRM

. T h ",'Z d %Z U ;- f?

Iasert D -> - 4 ;;'; 0.% W : 47 4 0.Z " : 0'.f* -

b. Inoperative RA NA

c. Downscale 15% of RATED THERMAL POWER 13% of RATED THERMAL POWER

d. Neutron Flux - Upscale Startup $ 12% of RATED THERMAL POWER $ 14% of RATED 1HEIPIAL POWER

3. SOURCE RANGE MONITORS w a. Detector not full in MA 5

NA 1 b. Upscale $ 1 x 10 cps < 1.6 x 10 cps w c. Inoperative NA HA h d. Downscale 3 0.7 cps 3 0 5 cps **

4. INTEllMEDIATE RANGE MDNITORS

a. Detector not full in MA NA

b. Upscale $ 108/125 division of full i 110/125 division of full scale scale i c. Inoperative NA NA

d. Downscale -> 5/125 division of full -> 3/125 division of full scale sca,le

5. SCRAM DISCHARGE VOLUME

a. Water Level-High - LISN602A $ 18.00" $ 21.12" LISN6028 $ 18.00" $ 21.60"

6. REACTOR COOLANT SYSTEN RECIRCULATION FLOW

a. Upscale $ 108% of rated flow $ 111% of rated flow "The Average Power Range Monitor rod h1ock function is varied as a functit,n of recirculation loop flow (W).

The trip setting of this function must be maintained in accordance inith Specification 3.2.2.

- h;,@ led signal to noise ratio is 1 2, otherwise setrf..',qt of 3 cps and a110wable 1.8 cps. ,

_ _ . _ . _ _ _ _ _ _ _ - _ _ _ _ _ _ ___ -_ __ _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ . _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ __ +- __ __

4 k Insert "D" ;

j !

c i

a. Flow Biased Neutron Flux Upscale i

1) Two Recirculation Loop Operation 5 0.66W + 421* 10.66W + 45%* l l

t

! 2) Single Reciredation Loop Operation 4 0.66W + 36.7%* 10.66W + 39.?!* j 4 l 4

. I i

e g

i d f

)

I l . l i i i

i A

i a

4 I

1 l

l i

s i

i l

1 1

l 1

1

3/4.4 REACTOR COOLANT SYSTEM 3/4,4.1 RECIFCULATION SYSTEM RECIRCULATION l.00PS r

LIMITING CONDITION FOR OPERATION

, , , v... _ . _ _ ___, _a ..._a_ _ 4 ..i. 4.. i.... - ' - -.6.11

w. 4n an.P a nn

. ,.e.a inw i = = w ww i = ww w wii s ey e --> .www. wwww-- -- ,- r- -- r - - -

wMt

. ,ma., ____ ,, . ____ __ u._ __ _ _ _ . _ , ._ ,,, ., __ _; ____ n. ..

bliult Ws wqwui ww 7 eig v3 I w ww w twuu i ywn p --

W. wwwwI wws u u s wit y e WW bMI J e2.....

_- . v . 9, A- -.

n i___

2.A2..w J...

_m._

t i, 1, 2. A.

v. v. .i i P. au - a1 .n A.1 1P. n iW . .L._._.

b. vi bw A.L.._

. er . , , s,.. .

Replace with a . , . 4 4 4.

Insert E ......w. m. .. __> ..

.... ...... ._ .....ww.....

Flr r b 4 bN & L & I I e VT E M i & VIM b bVIW 4 5 4 W el & WIIw de 'e

.. .4. 1..- ..a 4. ..... 4..

r__._-'s wi u-v ,. k.v

.,...a..

...ww. ...i.w

... . . a. .ey. 4ea. -- _.....ivuo

.www. .w i.. 3 .-

> _ _ as__i._ n_ sis _ _ __i>__ i _ __ t__ viiP au m e nn .v. .' N. _ ' g .. iL __

b 4,1 5W55 %5i=le w5 I 3ew&* = Iss I E b W i g I 3 5 I b I E b5 5h b I W54 bW I Ewlulh W 5 I I b &l IF1 b I i ....i n. at. ss_sa .__.sss_; s_ e s .. . . 9 A .1 s 1w Tv bR

..zaks. e km....

. a.

a si e i ywi = wi 1eeee - i s vi e a v = . . . . . ..

w5==1 ww wiww a uri s = ey=wie i w is

a. mi ._. AL_ .._fa J_ .a 1 __A a TAT r t ts evnAt ..

1.ded.a. . . . . . . _ _ .

5 sa ww IWW5b I n,F 5 isFI IV 4 brw v,.1 si vu v vv v v v w ww wris is a w i we ww yiWbu ossw wsi n =

&L. __ma L...__ a

-,i. i v = rs w s..e w., we

.a .._m a_m_

b we b, ibbwi i .. r iv

.a.__h.

_ms = = = ww i

_.. -A w ww s w e i , ey e = = rri

-..-O-_ -m i == ii _m.a==wwwiwir avvyw u si m

wr=s ww a... wie g

a. --A1.- a. --2.... Tursu A t MAlle n a. 1... ..

2 22

.._w =.awu ig i.. 2 2nd i e s i w v w wis a im o i wi e ww . .... rirui.-.. . w n a vi -w i=-. e.k. . ... ..

_....1 a ._ -1a2.2 A- 22... 9A 1 9.1 mdaki. 9 wqw=v -a. akm ww wiiw = eywinsssww vir i sy= w w vi esa e w T DTTT v .k.m...... .m . A.

a. 1--- ...Ja 2- .a 9...a r* A attin J, e ,6. d . f. .

L Ie4.a. . ,wwwwv

. . . . . .w d' w w awi L,. .ie iw yvv ww wwwww e i r.r1 i w i

- v rvv v w i w ww yr 1www w L.... .J 4. L1MT f ul ..lakl. _*me f L..._.

TT .T W .WW TUi 5Ew T wi u .fT vw MA,L.'A. .i. vW v wy iWWW a.k .

3ru TUwr9w w 4tv.3 WO i ***. d. ., ., .. .. & sf .

p._.1. . 4. d .a .

4 .ta e.., m, 4, a,.m,.. t i. .& .J

_8. h__ k. h.La. . ...m. . _. A. p ..$. . e. ,4 4,m_m _. , 6. I. . .. . . ..7, Ji- -a vi nP au m a . .. r p m di .. i... AEW 8 --a 2 ----

g. 9..m i. m...

a.ka. I W vs b aT y v v vv suww eswui www wy v w w w us mwI g I 5 W vs W 5 Iv i i ib .w Ir1.,

. ...w

%2_A. AAA.A A- PE__.-- m 1 % _.1 _

,,A WubW4

__ AL_-

41I555

.L_

b43W I 43115 b 5pWh4I IWu Ist ( lyw s W %.7 w dee'he annu -_; ammune __J-. 1____t _ se ._ __sgg____ m a e e .s a ma._k_.

w is wu a sus vitu at.

wiiw riv iwa =aiw by su i t rw I e g a5yw55 gdviYW4 . I uiIg 5 7e7, e , 4 e (, f ,

eu us mi s___i ____ --- a L_.._- __J Wg fi b I455b we gh W yu5 w i sw w i eg wi vu u

a. k.. . _ _ _ ,- 1i - m.v i . i . ,. .s,= v.11.e s. .u.m. e.nAur.s 6 Ag yJa.v. g g ,' i n.

. . . a. ,.

4v... d. e.

i I w- , =

kI.

S

.U.,i.

--Mw-

. . .v . w. h. ....h.

... - .. ...ww i .u -== ,

.. N k. $.h. . 8. k. . A D D_W. ..

t.D. D_u.a n

_ .._._.+A_.

. .. . f l. i.n.w- . . .... .

.. . i. . <, . i. ..

... .,...+.s..

..... + k. .m .

1l_. ..J.. 1..mi.

= , a.w . i, J.L.J L..

u.. aA_.. akm2.

view e u

. .k i-..ww ww wis i i i vis e rw i e u a is v = v w g E,--....ww . Jg .a ,.1.g..

wT i s u u o a ver= e i

A _a_ .-__..m2... . al.. a_ __.a__. mL. _.d.. 1,=-. . . .- . .i .-. o d e.k. t .

! w= =e-= wiew 1w -w Tfh> u== wwiiww=sTw ww w w wu n 1 u_ ____.s__2 ss_ma. ..a,ska. 9 6... u.. e.. .

4..., .... . . . . s i. .m ._.

wiew i s qw i i w is e i == v w w - wii e vi e iv... . wp .w. ... .. -- _

A Uf . # _ m .a t..J. m.. di a Ws -- k.- - J.. 4-. Tue nu a i MAi_rc B _

ewme,n wg Ww .. ierg i v . v w w i um u w v's m (1

6. h. . . ... vn w . . . . . v.

e e t-9J_J. __-_JE1_a J_ PJ__.-

L. , 1 ..

ww swee ak._

wi v= i s v. wwww -=w

_ _... a.

A L

=<rs ,. s i ssi a b syWwaiiWw usi r s yvi W e 4 d. . 4 - 4 4.

_ = g.

a t ._ ._ e_r

.,4._ .

1 T. ... .- a, e . w- p a. 2 w_ i , v..w. m. 9..- __..__ a v.A.._..,

.1f.....

A

...J. P. .-...

.8 t.e. n.u. .wwa , 2,11,__ -.yo, _

.www yiv. .a. .W,.w., r, N _ & $_ _ P h_ ._ 9 L. __1A__ 2 J

._J

__ f.t

.g

. . . . t w. n a.u. .w a.., d

. v y.

d_

its awvL w_ ..-.a._ wus

..A vu AL_

wivw wwi u wi 'sw-wIw - 9 2wW rwv i i s o w s w w .

RIVER llDID - UNIT 1 3/4 4-1 i

Insert "E" 3.4.1.1 'Ihe reactor coolant syste recirculation loops shall be in operation and in Region I 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, 3.3.6, or

b. A single loop operating with:

1. Voltmetric recirculation loop flow rate less than or equal to 33,000 gpn, and

2. 'Ibe recirculation loop flow control systs in the loop Manual (Position Contzel) Hode, and

3. 'mEr+E P0ha less than or equal to 70% of FATID 'mERAL IChu

4. Limits and setpoints for single recirculation loop operation per Specifications 2.1.2, 2.2.1, 3.2.1, 3.2.2, and 3.3.6, and APPLICABILITY: OPERATICNAL CCNDITICNS 1* and 2*

JcrICN

a. During single loop operation, with voltretric recirculation loop flow rate greater than 33,000 gpn, innediately initiate corrective action to reduce flow to less than or equal to 33,000 gpn within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months ,

b. During single loop operation, with the recirculation flow control system not in the Icop Manual node, innediately initiate corrective action to place the recirculation flew control syst s in the Icep Manual mcde within I hour.

c. During single loop operation, with 'mEfeE Pohn greater than 70% of RATED 'mERAL PChD, innediately initiate corrective action to reduce THER%L Pohn to less than or equal to 70% of

)

PATID NE Pohn within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months . ,

i

See Special Exception 3.10.4 1

I l

d. Within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months u;rn entry into single loop operation, verify that the operating limits in Specification 3.2.1 have been appropriately adjusted for single loop operation.

e. Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months upon entry in'o c single loop operation, verify that the setpoints in Specifications 2.2.1, 3.2.2 and 3.2.6 are within apprcpriate limits,

f. During single loop operatirn with either MiERMAL PNER 6 30% of RATED DIEIM.L POWER or recirculation loop flow in the operating Icop is 6 50% of rated recirculation loop flow and ta perature differences exceeding the limits in Surmillance Requirment 4.4.1.1.4, suspend THERMAL POWER or recirculation loop flow increases. *

g. With one or two reactor coolant syste recirculation loops in operation and total core flow greater than 39% and less than 45%

of rated core flow and DIERMAL PNER greater than the limit specified in Region II of Figure 3.4.1.1-1:

1. Determine the APRM and LP E** noise levels (Surveillance

4. 4.1.1. 2) :

a) At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and b) Within 30 minutes after ccupletion of a VIERMAL PCEER increase of at least 5% of PATED THERMAL POWER.

2. With the APRM or LPFM** neutron flux noise levels greater than three tires their established baseline noise levels, imediately initiate corrective action to restore the noise levels within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by increasing core flow to greater than or equal to 45% of rated core flow or by reducing UmAL Pohn to less than or equal to the limit specified in Region II of Figure l l 3.4.1.1-1. i

h. With one or two reactor coolant syste recirculation loops in operation and total core ficw less than 39% of rated core flow and MIERMAL P0h3 greater than the limit specified in Region III of Figure 3.4.1.1-1, imediately within 15 minutes initiate corrective action to increase core flow to greater than or equal to 39% of rated core flow or reduce DIERMAL Pohn to less than the limit specified in Region III of Figure 3.4.1.1-1 witidn 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

With one recirculaticn loop not in cperation and isolated, the differential taperature requirments of Suzwillance Requirment ,

4.4.1.1.4b and c are not applicable, and the provisiens of I specification 3.0.4 are not applicable with respect to surveillance Requir ment 4.4.1.1.4b and c.

- Detector levels A and C of one LPPM string in the center of the core should bn renitored.

l l

1 REACTOR COOLANT SYSTEM SURVEILLANCE REOUIREMENTS

. . . . . . . .. , . . ....-. _ _..._.._2._.._.....,_

. . . . .. . . . ... . ...,m.. ;

.,_a.,.3..,.

- . . e ..

. . u.. . . . . . . . . . . . . . . ..., , ..

.. . u . sm.

s___ .. 3.... .... ... 2.

. . . .u.s.ss. .. . _ . . . _ . , ..... . . . .2. n..a

. . c e ... n._n ie

.- =" '---- - - - - - r-- -- - - - ' ' - -#*

.... ... i. . . . , s s. .. ..w.. s. ._ u...__ s.

u.

. . 2. , .. .. 3. 4. . .. .

& u. . s e. ,o .., s. . *_ w. .. .. . .u. .. .

gr . . . . . . .

. u.. . .

u.g. 2 ._ _ . . s ,.

.i.. .. i y .g

. .,. u. . .... u.. .

.. ____ .... ,....,...s..._...__... __.

4. . ...

s. . u. . a. ,, ,....

a1

_ f. . .... * ..k... . ..

..1 ..

,,W ,

..__L.

pt

_____J

. r . . . k..._ _ , ...2

.... .6..... .. .. . ......, . . . ,....

,,e u,

7., ......2

._ 1 11_1_.9 f . . k i. a. ...k.

L ..i,l. .

. . . . . B B.ad....J

.. . ,.A

. .e.n . .nu . . . .. A. 1. m ..2..

. . e__..A._.__

._g..,

wMhie

,__ .. . , ....., . . r,.e

. ,,nu . r, _

s :

.2

___> .. ..n

___.s.

g,r... ...n a.._ _..a..__.._ a. ._ . ._ . u. .. s. ._ u.

.w i. a.si..,

n_

u._.._.

- __. _>__ ... __>__... uv _. a__

..y.

,,,y a. 3_ ..y.... . . . , x. . a.ss

_ __ . u _._>.

..a4.

I . s .g s.. . _...._.a. .5....i

.i...--

i

. . . .isw

,...u...,

1 . 4 w.. 34.4 . 6. . .. ___.4.....,..

7 .. ... ., . . . . .

u___ ..

7. .,

. . . . . . , , , _u.... __..e. _ . . . . t. . s.

)

w44ge. l Replace with Insert F i

l i

f 1

I I

1

, _ . . _ , . . __ ,, _, ___ .r,ma nu

... ...... n ...,2 . .. .u. .., ,ci

_._,__v.,___ .... ..__..o. ____

_ . y... . . . . . . ._ , . . . 2_.. .___

.m..__.._._.__

__s,,__ s. _n _ i. ' _ __ __ . A, ____ . t ._ . s 2 t.._ ___.. 4.4eeed.

__2

. . . . e. . a. . . . . i.n nu, m . . . , . ...... . u.

. ... . o j

I i

I 1

1

RIVER BEND - UNIT 1 3/4 4-2 J

l

Insert "F" 4.4.1.1.1 Each reactor coolant syst e recirculation loop flow control valve shall be deconstrated OPERABLE at least once per 18 nonths by Verifying that the control valve fails "as is" on loss of a.

hydraulic pressure at the hydraulic control unit, and

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

1. Iess than or equal to lit of stroke per second opening, ard

2. Iess than or equal to 11% of stroke per second closing 4.4.1.1.2 Establish a baseline APPM and LP M* neutron flux noise valve within the regions for which nonitoring is required (Specification 3.4.1.1 AcrION c) within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months of entering the region for which nonitoring is required unless baselining has previously been performed in the region since the last refueling outage.

4.4.1.1.3 Initially, within 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months upon entry into single loop operation and once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months thereafter, verify that

a. 'IMERRL POWER is less than or equal to 70% of PATED 'IHERVAL POWER, and

b. The recirculation flow control system is in the Icop Manual (Position Centrol) mode, and

c. 'Ihe voltretric recirculation flow rate is less than or equal to 33,000 gpn.

4.4.1.1.4 With one reactor coolant system recirculation loop not in operation, and either 'IHER%L POWER less than or equal to 30% of RAH:D

'IHER%L POWER or the recirculation loop flow in the operating loop is less than or equal to 50% of rated recirculation loop flow, within 15 minutes prior to an increase in NRL POWER or recirculation locp flow, verify that the following differential ternperature requirments are net:

Detector levels A and C of one LPPM string per core octant plus detectors A and C of one 1PRM string in the center of core should be nonitored.

I i

a. 6 100 F between reactor vessel steam space coolant and bottan head drain line coolant, and

b. i 50 F between the reactor coolant within the loop not in !

operation and the coolant in the reactor pressure vessel **, and

c. 6 500 F between the reactor coolant within the loop not in (

operation and the c5mrating loop.**

t l

l i

L i

1 1

1

- With one recirculation loop not in operaticri and isolated, the differential terrperature requirments of Surveillance Pequirenent ,

, 4.4.1.1.4b and c are not aIplicable and the provisicri of :

l Suneillance Requirement 4.0.4 are not applicable with respect to l 1 Surveillance Requirment 4.4.1.1.4b and c. l i !

1 a

Rcpleco with now Figure Insert H e

\l i i e t i I l l t ' i i . ./

\ t 6 i i e i i e e if

\. \ s t i t 6 i i i a 6 6/t N g i , i i i i i e i/ i .

i i e ! ,

i i 6 e i e i ! ./ 6 i i

i\i\

4 e i i i A 4 ite i !/ 6 i~

\\ 6V '

i VN t/1 i I t i i 6 i i i l

\ ix /t i g i i il i \ -

i i i t i I e i 4/ e 6 e 6 i 11 \ t i / l i 1 \ /

\ \ / .

I \i \ / -

d I \ l

/i i e i W

\ \ / i i i i- N / i l i g @

g

" i '

! \ l \ / t i g " y3 i \ i i s / I e Ii ea i e .

l\ l i \ Y I i e6 i E ".

1 ii % / ' i 6 t i f. $ g,i 7

ti Li6 i ey I i e i 3 i

i t

e i i i l

\i 6 i 6

\ii6

./

/i iX

\l 46 , 4 6 , 4 i i . .

0 m, ,

w$

i i t t t 1,. 4

/. i e%. t i i i i e i i g 6 6 6 t 6 I i\i I/ 6 i N it i I 6 6 e l i g i i i i i i t\ !A i e iX 6 i i i e i e i . u y i i ! i i e i i i1/4 eie6 i \ i .

e i . . . , 3 i s i a .

i A6 6 6 i i i i 1 i . . . .

i eii i i 3 I I / \ ! t ! i iN 6 e t i i e i . I , =!

16 i i i t i i e / \6 l 1 4 4 e i\ 4 i e i i i e i ;

i e e i e if i .\ t t i e i , I i \ . i i . .

6 i e i e i/ i 6 e ir 6 e ie i i e i A4 e i e i i e i i ' i . Vi i e i \ t i i e i . . i\i . .

i i i e ! ./. i e i 6 1 i e i e i i i , e xi . i . .

I I I IA6 i

6 \ t i e e i i a t i i\ 6 iei 6 ;

6 e e /I 4 i e i 'i t 6 6 4 i i . 4 e i\1 e i i 4 6 i /t 6 i i t ieT4 i i 8 e i e i t i e i Ae e i e d

i e i / i i i e i e \l e i e i i . . e i . IN.

. e/o i e i e i 16 ! 6 i i e i e i . .

.\i i . I

e iA i e i! i 6 \ii ' i 6 4 i a e i . e i Ne i

' I / i i t i i ll i i t i e i ; i 6 t t i \i

/i i t i 6 i i i i e i i i i i l . . IN fi e i i , 3 i i e i i . . i e i , t o e i , a e e i\ l 2 a l

, t 3 3 8 2 2.

1 i (GBAYW %) WIMod WWW3Ha l t

RIVER BEND - UNIT 1 3/4 4-3 1

i

vvs&Rr H l

1 l

l 70

-Kf REGION Ill

! i n

50 Q "

d'

% :-___' / i "84 40 lu ,

._.i x P f_ _'_ _ _

~

j ,

o ,6 kEGION a l

' I 30 m,' '

d e: '

m i e

y -- .__-

i A '

20 10 _._____

i .

0 65 75 85 15 25 35 45 55 CORE FLOW (T; RATED)

FIGURE 3.4.1.1-1 THERMAL POWER VERSUS CORE FLOW t

i

REACTOR CCOLANT SYSTEM JET PUMP 5 LIMITING CONDITION FOR OPERATION 3.4.1.2 All jet pumps shall be OPERABLE.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With one or more jet pumps inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . ,

SURVEILLANCE REQUIREMENTS During two Recirulation loop operation each 4.4.1.2.1 of the above required jet pumps shall be demonstrated OPERABLE prior to THE L POWER exceeding 25% of RATED THERMAL POWER, and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while greater than 25% of RATED THERMAL POWER, ,by determining recirculation.loep flow, total core flow and diffuser-to-lower plenum differen-tial pressure for each jet pump and verifying that no two of the following conditions occur when both recirculation loop indicated flows are in compliance i with Specification 3.4.1.3,

a. The indicated recirculation loop flow differs by more than 10% from the established" flow control valve position-1 cop flow characteristics. l

b. The indicated total core flow differs by more than 10% from the '

establishee*' total core flow value derived from recirculation loop I flew measurements.

c. The indicated diffuser-to-1cwer plenum differential pressure of any individual jet pump differs from establishoo# patterns ey more snan {

t 10%.

d. The provisions of Specification 4.0.4 are rot applicable proviced i that this surveillance is performed witnin 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after excr.eding 25% of RATED THERMAL POWER.

1 l

insert t )

l I

1

Te w e cet -et ce eu 'n; the eterte; t::t ;re; tr.

RIVER BEND UNIT 1 3/4 4-4 ,

1

Insert "I" 4.4.1.2.2 During single recirculation loop operatim , each of the required jet punps in the operating recirculation locp shall be deonstrated OPEPABIE at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months ktlile greater than 25%

of RATTD 7EEME PChu, by detemining recirculation loop flow in the

! cperating loop, total core flow and diffuser-to-lawer plentrn i differential pressure for each jet pwp in the operating loop and verifying that no two of t!.e following conditions occur: ,

a. The indicated recirculation loop flow in the operating loop differs by nere than 10% fra the established

single recirculation flow control valve position -

locp flow characteristics,

b. The indicated total core ficw differs by more than 10% frun the established

jet pmp flow / recirculation pmp flow ,

characteristic for the operating locp.

c. The indicated diffuser-to-lower plen a differential pressure of :

any individual jet pump differs frm established

single recirculation Icep patterns by more than 10%.

d. The provisions of specification 4.0.4 are not applicable provided that this surveillance is perfomed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after exceeding 25% of FA77D THERE PCED.

t l

l l

1 l

! l

7b be detemined during initial use of single loop operation.

Surveillance Requirments of 4.4.1.2 are not required to allow deteminaticr) of characteristic curves.

a i

i

REACTOR COOLANT SYSTEM RECIRCULATION LOOP FLOW LIMITING CONDITION FOR OPERATION 3.4.1.3 Recirculation loop flow mismatch shall be maintained within:

a. 5% of rated recirculation flow with core flow greater than or equal to 70% of rated c,cre flow,

b. 10% of rated recirculation flow with core flow less than 70% of rated core flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*g. during two recirculation loop operation.

ACTION:

With recirculation loop flows different by more than the specified limits, either:

a. Restore the recirculation loop flows to within the specified. limit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , or

b. 1:r: t: r 'r:$?5$f:r5 f (( Ib E: 1: : ' W et ' : :-itie- l and take the ACTION required by Specification 3.4.1.1T *~

SURVEILLANCE REOUIREMENTS 4.4.1.3 Recirculation loco flow mismatch shall be verified to be within the limits at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

"See Special Test Exception 3.10.4.

- The Provisions of Specification 3.0.4 are not Appliceble.

y I

J e

RIVER BEN) - UNIT 1 3/4 4-5

REACTOR COOLANT SYSTEMS 4

1 BASES 3/4.1.3 CONTROL RODS The specifications of this section (1) ensure that the minimum SHUTDOWN MARGIN is maintained and the control rod insertion times are consistent with l those used in the safety analyses, and (2) limit the pctential effects of the rod drop accident. The ACTION statements permit variations from the basic requirements but impose more restrictive criteria for. cortinued operation. A )

limitation on inoperable rods is set such that the resultant effect on total ;

rod worth and scram shape will be kept to a minimum. The requirements for the various scram time measurements ensure that any indication of systematic pro- l blems with rod drives will be investigated on a timely basis.

' Damage within the control rod drive mechanism could be a generic problem.

iherefore, with a control rod immovable t'ecause of excessive friction or aechanical interference, operation of the reactor is limited to a time period )

that is long enough to permit determining the cause of the inoperability yet prevent operation with a large number of inoperable cant-ol rods.

Con' trol rods that are inoper'able for other reasons are permitted to be ,

taken out of service provided that those not fully inserted are consistent l with the SHUT 00WN MARGIN requirements, f

- fhe number of control rods pe'rmitted to be inoperable could be more than the alght allowed by the specification, but the occurrence of eight inoperable rods could be indicative of a generic problem and the reactor must be shut down for investigation and resolution of the problem the fuel cladding safety limit

+ The control rod system'is designed to bring the reactor suberitical at a i cate fast enough to prevent the MCPR from becoming less than . uring the l l limiting power transient analyzed in Section 15.0 of the FSAR. This analysis shews that the negative reactivity rates, resulting from the scram with the average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than 1-09. The occurrence of l scram times longer then those specified should be iewed as an indication of a systemic problem with the rod drives and, theref e, the surveillance interval is reduced in order to prevent operation of the eactor for long periods of time with a potentially serious problem. the fuel cladding safety limit The scram discharge volume is required to be OPERABLE so that it will be available when needed to accept discharge water from the control rods during a reactor scram and will isolate the reactor coolant system from the containment l when required.

Control rods with inoperable accumulators are declared inoperable and

Specification 3.1.3.1 then applies. This prevents a pattern of inoperable I accumulators that would result in less reactivity insertion on a scram than has been analyzed even though control rods with inoperable accumulators may still be inserted with normal drive water pressure. Operability of the' accumulator ensures that there is a means available to insert the control rods even under the most unfavorable depressurization of the reactor.

RIVER BEND - UNIT I B 3/4 1-2 AmendmentNo.k l

POWER DISTRIBUTION LIMITS .

BASES

-l AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

3. Corrected guide tube thermal resistance.

4. Correct heat capacity of reactor internals heat nodes.

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption l that flow from the bypass to lower plenum must overcome a 1 psi pressure. drop in core.

2. Incoporate NRC pressure transfer assumption - The assumption used in l

the SAFE-REFLOOD pressure transfer when the pressure is increasing was changed.

A few of the changes affect the accident calculation irrespective of CCFL.

These changes are listed below, 4

a. Input Change i

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

b. Model Change

1. Improved Radiation and Conduction Calcul$ tion - Incorporation of CHASTE-05 for heatup calculation.

a j A. list of the significant plant input parameters to the loss-of-coolant ,

accident aralysis is presented in Bases Table B 3.2.1-1. l Insert J-3/4.2.2 APRM)SETPOINTS j The fuel cladding integrity Safety Limits of Specification 2.1 were based on a power distribution which would yield the design LHGR at RATED THERMAL POWER.

The f.ow biased simulated thermal power-high scram trip setpoint and the flow biased neutron flux-upscale control rod block trip setpoints of the APRM instru-ments must be adjusted .. ....... ..... .... ~ n .....................,( l or that > 1% plastic strain does not occur in the degraded situation. The scram

~

settings and rod block settings are adjusted in accordance with the formula in this specification, when the combination of THERMAL POWER and CMFLPD indicates a peak power distribution, to ensure that an LHGR transient would not bc

)j increased in degraded conditions.

for both two recirculation loop operation and single recirculation loop i operation to ensure that MCPR does not become less than the fuel cladding safety

! Ilmit l

B 3/4 2-2 Amendment No, h(

RIVER BEND - UNIT 1 ,

1 I

l

4

- Insert "J" For plant operation with a single recirculation loop, the MAPLHGR limits !

of figures 3.2.1-1 through 3.2.1-6 are multiplied by 0.64. The constant factor 0.84 is derived from LOCA analyses initiated from single recirculation loop operation to account for earlier boiling transition at che limiting fuel mode compared to the standard LOCA evaluations.

4 f

l i

e r

i i

l i .

1 -

i I

I I

[

I a t i ;

t

., . -~

POWERDISTRIBUTIONLIMIH RECEIVED Bases Table 8 3.2.1-1 SQC SIGNIFICANT INPUT PAFjMETERS TO THE LOSS-0F COOLANT ACCIDE]J,T,, ANALYSIS Plant Parameters; Core THERMAL POWER .................... 3015 Mwt* which corresponds to 105% of rated steam flow Vessel Steam Output ................... 13.08 x 106 lb e r which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure............. 1060 psia Design Basis Recirculation Line Break Area tor:

a. Large B%aks 2.2 ft2.

b. Small Breaks 0.09 ft2.

Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL ASSEMBLY GENERATION RATE PEAKING POWER FUEL TYPE , GEOMETRY (kw/ft) FACTOR RATIO 13.4 1.4 1.17 #/ l Initial Core 8x8

'A more detailed listing of input of each model and its source is presented in Section II of NEDE 20566II) and subsection 6.3.1 of the FSAR.

"This power level meets the Appendix K requirement of 102%. The core heatup calculation assums an assembly power consistent with operation of the highest powered rod at 102% of its Technical Specification LINEAR HEAT GENERATION RATE limit.

For single recirculation loop operation, loss of nucleate boiling is assumed at 0.01 after LOCA regardless of initial MCPR.

RIVER BEND - UNIT 1 8 3/4 2-3

~

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 Mi !NN CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR of 1.07 and an analysis of abnormal operational transients. For any abnormal operating transient analysis, 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 settings given in Specification 2.2.

To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operationa.1 transient, the most limiting tran-sients have been analyzed to determine which result in the largest reduction la CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow increase in pressure and power, positive reactivity insartion, and coolant temperature decrease. The limiting transient yields the largu t delta MCPR.

When added to the Safety limit MCPR of 1.07, the required minimum operating imit MCPR of Specification 3.2.3 is obtained and is presented in Figure 3.2.3-1.

insert "K"'a,he power-flow map of Figure B 3/4 2.3-1 shows typical regions of plant operation.

The evaluation of a given transient begins with the system initial param-eters identified in Reference 2 that are input to a GE core dynamic behavior transient corrputer program. The codes used to evaluate transients are described in Reference 2. The principal result of this evaluation is the reduction in MCPR caused by transient.

The purpose of the MCPRf and MCPR p of Figures 3.2.3-1 and 3.2.3-2 !s to define operating limits at other than rated core flow and power conditions.

At less than 100% of rated flow and power the required MCPR is the larger value of the MCPR and MCPR at the existing core flow and power state. The MCPRf s f p dre established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The MCPR f s were calculated such that, for the maximum core flow rate and ,

I the corresponding THERMAL POWER along the 105%-of-rated steam flow control line, the limiting bundle's relative power was adjusted until the MCPR was slightly above the Safety Limit. Using this relative bundle power, the MCPRs were calcu-lated at different points along the 105%-of-rated steam flow control line correspo: sing to dif ferent core flows. The calculated MCPR at a given point of core flow is defined as MCPRf.

l RIVER BEND - UNIT 1 B 3/4 2-4 Amendment No. k l

l

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

O * *

Insert "K" Analysis of transients occurring during single recirculation loop operation indicates that the maximum operating limit MCPR will be bounded by the limits in Specification 3.2.3.

l l

I 1

I l

I a

l

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

i 3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM Replace with Insert "L" 4

0; r:tir with :n rnet:r n:i=t rnir=hti= in; in;;r:th i: pre-

'ibited =ti' = =:inti:r :' th: ;;r':=== 0' th: 5005 dur' ; := !=;\

r

ti= hr b:= ;;r': nd, ad n:h :;;r:ti= h= inn det:=f =d t: 5: -

z z;t 910.

An inoperable jet pump is not, in itself, a sufficient reason to declare a recirculation loop inoperable, but it does, in case of a design basis accident, increase the blowdown area and reduce the capability of reflooding the core; thus, tne requirement for shutc'own of the facility with a jet pump in. operable.

,M,,

Jet pump failure can be detected by monitoring jet oump performance on a I"S*It prescribed schedule for significant degradation.1 Kvcirculation loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria. <-

The limits will ensure an adequate core flow coastdown.from either recircu-1ation loop following a LOCA. Insert "N" f or two reci rculat ion =r--

loop opeyation.

In order to prevent undue stress on the vessel nozzles and bottom head r,egion, the recirculation loop temperatures shall be within 50*F of each other **

prior to startup of an idle loop. The loop temperature must also be within

50*F 6? the. reactor pressure vessel coolant temperature to prevent thermal shock to the bett = c'recirculation the '!::: ! pump f: :t :and h = recirculation r t =;;r:tur: th= nozzles.{:ir.n

_ n:1=t bth; th: :ni=t '- th:

up;;r-regen c' th: care, =d= :tr:n = th: vuni =uld nit if the ter.p;r ture "i ";ra n a r; gr nts th e 100'.". hsert"0" The objective of GE BWR plant and "al design is to provide stable opera-tion . tith margin over the normal eqerating domain. However, at the high power /

low flow corner of the operating demain, a small probability of limit cycle neutron flux oscillations exists depending on combinations of operating condi-

.tions (e.g., rod pattern, power shap.1). To provide assurance that neutron flux limit cycle oscillations are detectM. and suppressed, APRM and LPRM neutron flux noise levels should be monitorgd while operating in this region.

/ .

Stability tests at operating BW8s were reviewed to determine a generic region of the power / flow map in which surveillance of neutron flux noise levels should be performed. A conservative decay ratio of 0.6 was chosen, as the basis for determining the generic region for surveillance, to account for the plant-to-plant variability of decay ratio with core and fuel designs. This generic region has been determined to correspond to a core flow of less than or equal to 45% of rated core flow and a thermal power greater than that specified in Figure 3.4.1.1-1 (Reference 1).

Plant-specific calculations can be performed to determine an applicable

,:gion for monitoring neutron flux noise levels. In this case tha degree of conservatism can be reduced since plant-to plant variability would be eliminated.

In this case, adequate margin will be assured by monitoring'the region which has a decay ra,tio greater than or equal to 0.8.

RIVER BEND - UNIT 1 B 3/4 4-1

REACTOR COOLANT SYSTEM l

I 3/4.4 REACTOR COOLANT SYSTEM 3

BASES ECIRCULATIONSYSTEM(Continued)

Neutron flux no'ise limits are also established to ensure early detection of limit cycle neutron flux oscillations.. BWR cores typically operate with neutron flux noise caused by random boiling and flow noise. Typical neutron" flux noise levels of 1 to 12% of rated power (peak-to peak) have been reported for the range of low to high recirculation loop flow during both single and ;

dual recirculation loop operation. Neutron flux noise levels which signifi- !

cantly bound these values are considered in the thermal / mechanical design of GE BWR 'uel and are found to be of negligible consequence (Reference 2). In addition, stability tests at operating SWRs have demonstrated that when stabil-ity related neutron flux limit cycle oscillations occur they result in peak-to-peak neutron flux limit cycles of 5 to 10 times the typical values. Therefore, actions taken to reduce neutron flux noise levels exceeding three times the typical value are sufficient to ensure early datection of ifmit cycle neutron flux oscillations. '

j Typically, neutron flux noise levels show a gradual increcse in absolute magnitude as core flow is increased (constant control rod pa# tern) with two reactor recirculation loops in operation. Therefore, the baseline neutron flux of noise core flows. level obt'ained at a specific core flow can be applied oigr a ' range l To maintain a* reasonable variat' ion between the low ilow and high flow ends of the flow range, the range over which a specific baself.ne is applied should not exceed 20% of rated core flow with two recirculation loops l in operation. Data from tests and operating plants indicate that e range of l 20% of rated core' flow will result in approximately a 50% increase in neutron l flux noise level during operation with two recirculation loops. Baseline da,ta should be taken near the maximum rod line at which the majority of operation will occur. However, baseline data taken at lower rod lines (i.e., lower !

power) will result in a conservative value since the neutron flux noise level I is proportional to the power level at a given core flow.

References (1) "BWR Core Thermal-Hydraulic Stability," Service Information Letter 380, Revision 1, February 1984.

(2) G. A. Watford, "Compliance of the General Electric Boiling Water Reactor Fuel Designs to Stability Licensing Criteria," December 1982 (NEDE 22277-P).

Insert "P" 3/4.4.2 SAFETY / RELIEF VALVES The safety valve function of the safety / relief valves (SRV) is to prevent the reactor coolant system from being pressurized above the Safety Limit of 1375 psig, in accordance with the ASME Code. A total of 9 OPERABLE safety-relief valves is ,equired to limit reactor pressure to within ASME III allowabl values for the worst case upset transient. Any combination of 4 SRVs operating in the relief , mode and 5 SRVs operating in the safety mode is acceptable.

RIVER BEND - UNIT 1 8 3/4 4-2 i 1

1

c..* ,,, .

Insert "L" The impact of single recirculation loop operation upon plant safety has been assessed and single recirculation loop operation is permitted if the MCPR fuel cladding safety limit is increased as noted by Specification 2.1.2; APRM scram and control rod block setpoints are adjusted as noted in Tables 2.2.1-1 and 3.3.6-2, respectively; MAPLHGR limits are decreased by the factor given in Specification 3.2.1 (Reference 3). MCPR operating limits are adjusted per specification 3/4.2.3, for both single and two recirculation loop operation.

Additionally, surveillances on the volumetric flow rate of the operating recirculation loop is imposed to exclude the possibility of excessive core internal vibration. The surveillance on differential temperatures below 30% THERMAL POWER or 50% rated recirculation loop flow is to mitigate the undue thermal stress on the vessel nozzles, recircule. tion pump and vessel bottom head during extended operation in the single recirculation loop code.

Insert "M" During single loop operation the jet pump operability surveillances are only performed for the jet pumps in the operating recirculation loop, as the loads on the inactive et pumps are expected to be very low due to the low flow in the reverse direction chrough the jet pumps.

Insert "N" In the case where the mismatch limits cannot be maintained during two recirculation loop operation, continued operation is permitted in a single recirculation loop operation mode.

F o . . sk .

Insert "O" Sudden equalization of a temperature difference 100 F between the reactor vessel bottom head coolant and the coolant in the upper region of the reactor vessel by increasing core flow rate would cause undue stress in the reactor vessel bottom head.

Insert "P" (3) Mp Opemion Analysis for River Bend Station. Unit 1,"

i I

l

O Enclosure 3 Single-Loop Operation Analysis for River Bend Station, Unit 1 May 1987 (NEDO-31441) 19

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

ML20148R648

Text

2.0 INTRODUCTION

Navigation menu

Search