Text

Proposed Tech Specs Increasing Neutron Flux Noise Surveillance & Permitting Operation W/One Recirculation Loop Out of Svc
	ML20155B085
Person / Time
Site:	Grand Gulf
Issue date:	03/31/1986
From:	; MISSISSIPPI POWER & LIGHT CO.
To:	;
Shared Package
ML20155B084	List: AECM-86-0092, Application for Amend to License NPF-29,changing Tech Specs to Increase Neutron Flux Noise Surveillance Per Rev 1 to GE Svc Info Ltr 380 & to Permit Operation W/One Recirculation Loop Out of Svc.Fee Paid; ML20155B085; ML20155B091;
References
	TAC-61083, NUDOCS 8604100201
	Download: ML20155B085 (31)
	v • d • e

. . , ,

d 2.0 SAFETY _ LIMITS AND LIMITING SAFETY SYSTEM SETTINGS

(

2.1 SAFETY LIMITS THERMAL POWER, low Pressure or Lew Flow 2.1.1 THERHAL POWER shall not exceed 25% of RATED THERMAL POWER with the reactor vessel steam dome pressure less thar. 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 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN 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.

THFRMAL POWER, High Pressure and High Flow N #NY ~

2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow. c APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: g g.g With MCPR less than-i-M!and the reactor vessel steam dome pressure greater than 785 psig and core flow greater than 10% of rated flow, be in at least HOT SHUT 00WN 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.

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: OPERATIO M L CONDITIONS 1, 2, 3 and 4. .

ACTION: ,

With the reactor coolant system pressure, as measured in the reactor vessel steam dome, above 1325 psig, be in at least HOT SHUTDOWN with reactor coolant >

system pressure less than or eg'ial 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.

Durig igte loop op=0% w A +ke revder- veul. 54*a.% dome pressure greder h 78F PST4 W com -Clow g@ h tot of 44 ~

-A %e. - Mc.?R, sha.A\ act be im A 1, o7.

l 8604100201 860331 l PDR ADOCK 05000416 i P PDR GRAND GULF-UNIT 1 2-1 l AMENbMENT No. '

TABL'E 2.2.1-1

, REACTOR PROTECTION SVSTEN INSTRUMENTATION SETPOINTS m

ALLOWA8LE E

O TRIP SETPOINT VALUCS

, FUNCTIONAL (*IT c1

% 1. Intermediate Range Monitor, Neutron Flux-Hign 5 120/125- divisions $ 122/125 divisions of full scale T of full sca)e g 2. Average Power Range Monitor:

Q e. Neutron Flux-Nigh, Setdown f n% of RATED 5 20% of RATED

.' THERMAL POWER THERMAL POWER g

+ b, Flow Bf ased Simulated Theresal Pouser "8tah j

1) tnow M a d 1 0.66 W+48%, with * " C m a, with a en i-- e. a maximum of M zf l.Z./-/ 2) High Flow Clacped ~ < 11T D% 5T itMC < 113.0% of RATED

4 ITEM 2.IP ~

' THERMAL POWER 4"".Ji a*"D

c. Neutron Flux-High ~< 118% of RATED < 120% of RATED

^

THERMAL POWER THERMAL POWER

d. Inoperative NA NA

> 3. Reactor Yessel Steam Dome Pressure - High 1 1064.7 psig i 1079.7 psig n*

4. Reactor Vessel Water tevel - 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 5 54.1 inches above instrument zero* instrument zero"

6. Main Steam Line Isniation Valve - Closure 1 6% closed 5 7% closed

- 7. Main Steam Line Radiation - High i 3.0 x full power 1 3.6 y full power background background

8. Drywell Fressure - High < 1.23 psig < 1.43 psig sun ;st

9. Scram Discharge Volume Weter level - High !"" f 6" n:!: N ef fd! : &

10. Turbine Stop Valve - Closure > 40 psig** > 37 psig

11. Turbine Control Valve Fast Closure, A Trip 011 Pressure - Low ';44.3 psiO** ),42 psig 2

h 12. Reactor Mode Switch Shutdown Fasition MA NA D 13. Manual Scram NA NA "See Bases Figure B 3/4 3-1.

3 ** Initial setpoint. Final setpoint to be determined during startup test program. Any s equiied a.mge to this setpoint shall be submitted to the Commission within 90 days of test completion.

k P

,w/re. w.+ x r.,o g n o a nsy,m a

o. r sW a

L. Fw sain S w' L

i

INSERT for Table 2.2.1-1, item 2.b 2.b: 1) During two recirculation loop operation:

Flow Biased 0.66 W+43%, with 0.66 W+51%, with a) a maximum of a maximum of ,

b) High Flow Clamped 111.0% of RATED 113.0% of RATED THERMAL POWER THERMAL POWER i

t

2) During single recirculation loop operation:

Flow Biased 0.66 W+40%, 0.66 W+43%,

a) b) High Flow Clamped Not required operable Not required operable l W

4 J14 MISC 86020502 - 11

2.1 SAFETY LIMITS B*SES M

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 integr.ity Safety Limit is set such that no fuel damage is calculated to occur 4- l if the limit is not violated. Because fuel damage is not directly observable.

a step-back approach is used to establish a Safety Limit-sweh4 hat the MCPR -+e-at 1:n th= 1.05. MCPR greater than OfF 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 <

I to its relative freedom from perforations or cracking. Although some corrosion or use related cracking may occur during the life of the cladding, fission l

product migration from this source is incrementally cumulative and continuously I measurable. Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly above design condi-tions 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 threshold beyond which still greater thermal stresses may cause gross rather than incremental cladding deterioration. Therefore, the fuel cladding Safety Limit is defined with a margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions represent a significant departure from the condition intended by design for planned operation.

2.1.1 THERMAL POWER, Low Pressure or Low Flow The use of the GEXL correlation is not valid for all critical power calcula-tions at pressures below 785 psig or core flows less than 10% of rated flow.

Therefore, the fuel cladding integrity Safety Limit is established by other means. This is done by establishing a limiting condition on core THERHAL POWER with the following basis. Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly inde::endent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER.

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

1 l

l GRAND GULF-UNIT 1 B 2-1 jk u g g /W, l

[ . _ _ .

_j

SAFETY LIMITS BASES 2.1.2 THERMAL POWER, High Pressure and High Flow The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor I operation, the thermal and hydraulic conditions resulting in a departure from l nucleate boiling have been used to mark the beginning of the region where fuel damage.could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. However, the uncertainties in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the fuel cladding i integrity Safety Limit is defined as the CPR in the limiting fuel assembly for which more than 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties.

The Safety Limit Analysis Basis, GETAB,MCPR , which is aisstatistical determined modelusing thatthe Generalall combines Electric of the Thermal '

uncertainties in operating parameters and the procedures used to calculate critical power. The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) Boiling Length (L),

GEXL, correlation. The GEXL correlation is valid over the range of conditions used in the tests of the data used to develop the correlation.

l The required input to the statistical model are the uncertainties listed

in Bases Table B2.1.2-1 and the nominal values of the core parameters listed in

Bases Table B2.1.2-2.

' Thebgsesfortheuncertaintiesinthecoreparametersaregivenin NED0-20340 andthebgsisfortheuncertaintyintheGEXLcorrelationis aiven in NEDO-10958-A .4 The power distribution is based on a typical i 764 assembly core in which the rod pattern was arbitrarily chosen to produce a skewed power distribution having the greatest number of assemblies at the highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analys h.

a. " General Electric BWR Thermal Analysis Bases (GETAB) Data, Correlation and Design Application," NEDO-10958-A.

b. General Electric " Process Computer Performance Evaluation Accuracy" l NEDO-20340 and Amendment 1, NED0-20340-1 dated June 1974 and December i

1974, respectively.

The baeas 4.. the ch up in uncee-Mdies he + a. W., ige im eyrwbn ny fic in %e GsA6 #A$e Loce oph 4== lysis daked A&mry ,#AG.

GRAND GULF- NIT 1 B 2-2 AhtENot4EN T Mo.

.-n. . , , - - - - - ., ., , wm..,-r, , . , , , , , _ ,.m.m.,mne_,..,wn,w.nn-.n.,,,,.ne,,n nw4 .,g

Bases Table B2.1.2-1 UNCERTAINTIES USED IN THE DETERMINATION OF THE FUEL CLADDING SAFETY LIMIT

Standard Deviation Quantity (% of Point)

Feedwater Flow 1.76 Feedwater Temperature 0.76 Reactor Pressure 0.5 Core inlet Temperature 0.2 Core Total Flow 2.5(f.)

Channel Flow Area 3.0 Friction Factor Multiplier 10.0 Channel Friction Factor Multiplier 5.0 TIP Readings 6.3 (b)

R Factor 1.5 Critical Power 3.6

" The uncertainty analysis used to establish the cora wide Safety Limit MCPR is based on the assumption of quadrant power symmetry for the reactor core. I

{

A.) Thi > va3Je ioc. raw $ b &.O b- s'ia.Je reelreoI b Ioop ope,-.N..%

b)Thh v4ae ~nem m b 4,8f;,. g.ge e, c ,,.o oi.+..og 1. ,p op.e. h g

GRAND GULF-UNIT 1 B 2-3 AMEINbKsgrNo.

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 (APLHGRs) for each type ' -

of fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in F,igure 3.2.1-1f clueinS+wo ly op. 4o. ce 'ea F'isore 3.2. i-2 duria3 6ieste to.p op m%.

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

ACTION: i- 4 With an APLHGR exceeding the' limits of Tigure 2.2.11, initiate corrective action within 15 minutes and restore APLHGR 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 .

SURVEILLANCE REQUIREMENTS

{cfpr.mble 4.2.1 All APLHGRs shall be verified to be equal to or less than the" limits :

deterair.ed frea rigoca 2.2.1 1.

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 1

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 R00 PATTERN for APLHGR.

d. The provisions of Specification 4.0.4 are not applicable.

l l

GRAND GULF-UNIT 1 3/4 2-1 AmoMeurNo.

- .- g - , e , ,,y- .n - -

e I,

L 0 l I I I I I I i' 8 CURVE FUEL TYPE o A 8CR210 E B 8CR l60 C 8CRO7i T 13 -

E 12.6 12.6 12.6 Q O A A

~ t 12 V i2.4 :i2.4 /

. 3: E 12.1 j gd5 12.0 ig, _

8

11. 8 dE n its n.s V.7 I4 c z It3j f 9 Its It2 11.1 "0

g i II - .

. w 10 4

. 1 58 io.4 10.2

, g

~ x3 so j hg 9.7 9.6

\,

J l -

y J 9 - 9.0 j 9.0 I e i l l I I I I

! O 5,000 10,000 15,000 20,000 25,000 30,000 35,000 40,000

A i 1 A/ERAGE PLANAR EXPOSURE (mwd /t)

E h

7 FIGURE 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) i g

VERSUS AVERAGE PLANAR EXPOSURE -=

INITIAL CORE FUEL TYPES 8CR210, 8CR160 AND 8CR071 h fa Two400eaumenov -"

i

l e 12 , a ,i e i i 1' CURVE FUEL TYPE

,1 A SCR2to 8 '

e SCR160 3 II l h

-4 y n p-ne ne na C SCRo71 ,

i i: H d m.s as ** no

,,3 .3 5=

<g z

,= 'o '

_ ni ,, . **

a. s.e 88 as

4 3:

4 3-

< 4 a a i.

gy w 3 - as e.*

S.7 2 o

, :s

' E3 .2 3, k

! N UE g m, -

li s .

7.7 4

I W 7.7 i E i a 1 .

a a a i a a a

{ 25,000 M,ooo n, coo 40,000 o 5,000 10,000 15.000 20,000

{

AVERAGE PLANAR EXPOSURE ( mwd /t) i ~6

! FIGURE 3.2.1-2 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHGR) VERSUS i AVERAGE PLANAR F.XPOSURE FOR SINGLE LOOP OPERATION

{ INITIAL CORE FUEL TYPES 8CR210E SCRl60 AND 8CRO71.

b $

bbor'S sbglo r eaVeulde loop w trem M*en :

Tr;p 6e+po'id AlW44 %e _

POWER DISTRIBUTION LIMITS S f (o.w w + Ao*A)T SS (0.& H +4's%) T 3/4.2.2 APRM SETPOINTS sg g(ogw + 545)T St aflo.a w es75 )T LIMITING CONDITION FOR OPERATION 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 (Sgg) shall be estab]ishedDur'se3 according to the No P aiorevieArion loopfollowing oyesw+ ion:relationships:

Trip Setpoint Allowable Value S < (0.66W + 48%)T S < (0. 56W + 51%)T -

Sj$(0.66W+42%)T R

Sj$(0.66W+45%)T p

where: 5 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 112.5 million Ibs/hr.

T = Lowest value of the ratio of FRACTION OF RATED THERMAL POWER (FRTP) divided by the MAXIMUM FRACTION OF LIMITING POWER DENSITY (MFLPD). 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:

With the APRM flow biased simulated thermal power-high scram trip setpoint and/

or the flow biased neutron flux-upscale control rod block trip setpoint less conservative than the value shown in the allowable value-column for S or S asabovedetermined,initiatecorrectiveactionwithin15minutesandrestNe, S and/or Son to within the required limits

within 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months or reduce THERMAL POWER to 1 Hs 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 .

SURVEILLAMCE REQUIREMENTS 4.2.2 The FRTP AND MFLPD for each class of fuel shall be determined, the value of T calculated, and the most recent actual APRM flow biased simulated thermal power-high scram and flow biased neutron flux-upscale control rod 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 completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and j

c. Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months wl'r the reactor is operating !

with MFLPD greater than or equal to FRTP. j

d. The provisions of Specification 4.0.4 are not applicable.

With MFLPD greater than the FRTP during power ascension up to.90% of RATED THERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may be adjusted such that APRM readings are greater than or equal to 100% times MFLPD 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 g ag g 4 l

.- l

- ~

l l TABLE 3.3.6-2

cs g CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS c, TRIP FUNCTION .

TRIP SETPOINT ALLOWA8LE VALUE' i k 1. R00 PATTERN CONTROL SYSTEM

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

Z POWER g b. High Power Setpoint < 70% of RATED THERMAL POWER < 70% of RATED THERMAL POWER APRMN~0 During two loop opuden i (o.66W + 42*/-)T

2 ( o.64 W + 45"/,) T

2*

j 2) Duriq imic la*P *P"*Ii+^ 5 ( 0 6(*W i 34*/*) T* 5 (O. Gb W + 37%) T*

Flow Biased Neutron Flux-l a.

l UDsCale ~

gv.vu - ,w f . sv.v. - m.-, .

l b. Inoperative HA NA

c. Downscale > 4% of RATED THERMAL POWER

> 3% of RATED THERMAL POWER

d. Neutron Flux - Upscale j Startup i 12% of RATED THERMAL POWER 1 14% of RATED THERMAL POWER j w 3. SOURCE RANGE MONITORS

! A a. Detector not full in MA , NA 5;* b. Upscale 5

< 1 x 10 cps < 1.5 x 10 5 c,,

$ c. Inoperative NA NA i d. Downscale 1 0.7 cps 1 0.5 cps

~

I

4. INTERMEDIATE RANGE MONITORS

a. Detector not full in MA NA

, b. Upscale < 108/125 of full scalc < 110/125 of full scale i c. Inoperative NA NA

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

l 5. SCRAM DISCHARGE VOLUME l a. '. Water Level-High i 32 inches 1 33.5 inches D

3 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW i

( 7.

a. Upscale REACTOR M00E SWITCH SHUTDOWN i 108% of rated flow 1 111% of rated flow NA j POSITION NA i

! *The Average Power Range Monitor rod block function is varied as a function of recirculation loop flow I

i h

(W) and the ratio of FRACTION of RATED THERMAL POWER to the MAXIMUM FRACTION of LIMITING POWER DENSITY (T factor). The trip setting of this function must be maintained in accordance with Specification 3.2.2. .

i ,

INSTRUMENTATION 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.10 The APRM and LPRM* neutron flux noise levels shall not exceed three (3) timas their established baseline value.

APPLICABILITY OPERATIONAL CONDITION 1 with operation in Region I as specified in Figure 3.4.1.1-1.

AC' ION a. With no established baseline flux noise levels, immediately initiate action to either reduce THERMAL POWER to within Region III as specified in Figure 3.4.1.1-1 or increase flow to within Region II as specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ,

b. With the flux noise levels greater than three (3) times their established baseline noise levels, initiate' corrective action within 15 minutes to reduce the noise levels to within the required limits within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ; if unsuccessful, either reduce THERMAL POWER to within Region III as specified in Figure 3.4.1.1-1 or increase flow to within Region II as specified in Figure 3.4.1.1-1 within the next 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

SURVEILLANCE REQUIREMENTS 4.3.10.1 The APRM and LPRM* neutron flux noise levels shall be determined to be less than or equal to the limit of Specification 3.3.10:

a. Within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months after entering the applicable region, and

b. At least once per 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months , and

c. Within 30 minutes after completion of a change in THERMAL POWER of at least 5% of RATED THERMAL POWER.

The provisions of specification 4.0.4 are not applicable.

4.3.10.2 Establish two loop baseline APRM and LPRM neutron flux noise levels at a point in Region II less than 60% of rated total core flow prior to operation in Region I of Figure 3.4.1.1-1 provided the baseline has not been established since the last CORE ALTERATION.

Detector A and C of one LPRM string per core ' octant plus detector A and C of one LPRM string in the central region of the core shall be monitored.

GRAND GULF - UNIT 1 3/4 3-111 AMENDMENT NO.

J14 MISC 86020502 - 1

4.3.10.3 a. Establish single loop baseline APRM and LPRM neutron flux noise levels at a point in Region II less than 60% of rated core flow prior to single loop operation in Region I of Figure l 3.4.1.1-1 provided the baseline has not been established since the last CORE ALTERATION; or

b. In lieu of establishing single loop baseline data, the baseline established in 4.3.10.2 may be utilized for single loop ;

operation in Regions I of Figure 3.4.1.1-1.

+

M GlLF - lEIT 1 3/4 3-112 k NDMENT No.

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

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

TWrERT 3/4 . 4-I LIMITING CONDITION FOR OPERATION y'

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

b. With no reactor co 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 STARTUP within 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months an in HOT SHUT 00 within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months , s

SURVEILLANCE REQUIREMENTS ,s 4.4.1.1.1 Both reactor coolant stem recirc ation loops shall be verified to be in operation at least o per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 ths by:

a. Verifying at the control valve fails "as is" n loss of hydraulic 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 t.han or equal to 11% of stroke per second closing.

"See Special Test Exception 3.10.4.

l GRAND GULF-UNIT 1 3/4 4-1 /f#ENhWNMd._

= - -

i Insert 3/4.4-1 page 1 of 2 l 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 single 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, 3.2.2, 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, immediately 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 0.00167 hours 9.920635e-6 weeks 2.283e-6 months and in HOT SHUTDOWN within the next 6 hours6.944444e-5 days 0.00167 hours 9.920635e-6 weeks 2.283e-6 months .

d. During single loop operation, with temperature differences exceeding the limits of SURVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the THERMAL POWER or recirculation 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 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

l

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 setpeints of specifications 2.1.2, 2.2.2. 3.2.1, 3.2.2, and 3.3.6, shall be implemented within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months or declare the associated equipment inoperable, (or the limits to be "not satisfied") and take the ACTIONS required by the referenced specifications.

See Special Test Exception 3.10.4.

f:

1 J14 MISC 86020502 - 9

.u , _ _ _ - _ - --

- J

Insert 3/4.4-1 page 2 of 2 SURVEILLANCE REOUIREMENTS 4.4.1.1.1 The reactor coolant recirculation system shall be verified to be in operation and not in Region IV of Figure 3.4.1.1-1 at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . i 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: j

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

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

i

1. Less than or equal to 11% of stroke per second opening, and 4 '

2. Less than or equal to 11% 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 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months .

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 0.00667 hours 3.968254e-5 weeks 9.132e-6 months .

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 following differential temperature require-ments are met within 15 minutes prior to beginning either a THFRMAL

POWER increase or a recirculation loop flow increase and within every hour during the THERMAL POWER or recirculation loop flow increase

a) less than 100*F, bwtween the reactor vesseI 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 loop and the coolant in the loop not 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, 3.2.2 and 3.3.6 shall be verified to be within the appropriate limits within i 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of an operational change to either one or two loops operating.

i J14 MISC 86020502 - 10

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

l l

l 2_

4 8

y 8

O E -

2 .

)

S

_ g I .

li!

Rb' x ,

5 2 GM 3 x '

5 d 5 - g

a N

s-B

.-________ ~

d '

g j IO h C e

m Z

~

h

$N kJ E

-9 .

t f f f 1 0

I I l i e g g 2 ? 2 2 9 l

! 8 3 2 R W3 mod WWH3H1031VM 30 IN30W3d f GRAND G11F - lNIT 1 3A 4 .lA AMDOMEur No. l 4 6

l l

l REACTOR COOLANT SYSTEM JET PUMPS 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 SHUTD0WN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS in an operm+as loor 4.4.1.2.1 Each of the above required jet pumpsFshall be demonstrated OPERABLE with THERMAL POWER in excess of 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 by determining recirculation loop flow, total core flow and diffuser-to-lower plenum differential pressure for each jet pump and verifying that no two of the following conditions occur: tr both %dic:ted recirc&-

tien 1;;p f?;n ;r; in ;;,sli; ace with Sp;;ificetica 3.4.1.2.

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

b. The indicated total core flow differs by more than 10% from the established total core flow value derived from recirculation loop flow measurements.

c. The indicated diffuser-to-lower plenum differential pressure of any individual jet pump differs from established patterns by more than 10%.

4.4.1.2.2 The provisions of Specification 4.0.4 are not applicable provided the diffuser-to-lower plenum differential pressures'of the individual jet pumps are determined to be within 50%* of the loop average within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months after entering OPERATIONAL CONDIT0N 2 and at least once per 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter.

hitialvalue. Final value to be determined during startup test program.

Any required changes to the value shall be submitted to the Commission within 90 days of test completion.

GRAND GULF-UNIT 1 3/4 4-2 NfN0Nrnr Mo.

REACTOR COOLANT SYSTEM l

RECIRCULATION LOOP FLOW LIMITING CONDITION FOR OPERATION

,wkes ko ims are .n opere.%so 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 core flow.

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

^

APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

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

i ther:,

,w -/estoretherecirculationloopflowstowithinthespecifiedlimit within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ).+c 'Ef unw uw%l,eWr:

pa,J e. mph w A n, vv...aw,9

' % wheler n.-h o ne q: .h: recirculation loop with th: 1:wcr ft:w r.:t fr. :p:r ti:n er.d take the ACTION requir: 4 Specification 3.4.1.1,4, or-

b. 15e '. 4 le

Hat % m aw%n 12 hon .

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

I I

A See Special Test Exception 3.10.4.

l

(

GRAND GULF-UNIT 1 3/4 4-3 Amsuce<rar Ma. ,,

l l

l REACTIVITY CONTROL SYSTEMS i

BASES 3/4.1.3 CONTROL RODS

The specifications of this section ensure that (1) the minimum SHUTDOWN i MARGIN is maintained, (2) the control rod insertion times are consistent with l those used in the accident, non-accident and transient analyses, and (3) the

, potential effects of the rod drop accident and rod withdrawal error event are

! limited. The ACTION statements permit variations from the basic requirements

! but at the same time impose more restrictive criteria for continued operation. !

A limitation on inoperable rods is set such that the resultant effect on total i rod worth and scram shape will be kept to a minimum. The requirements for the '

i various scram time measurements ensure that any indication of systematic prob-less with rod drives will be investigated on a timely basis.

l 4

Damage within the control rod drive mechanism could be a generic problem, therefore with a control rod immovable because of excessive friction or mechani-cal interference, operation of the reactor is limited to a time period which is reasonable to determine the cause of the inoperability and at the same time prevent operation with a large number of inoperable control rods.

Control rods that are inoperable for other reasons are permitted to be l taken out of service provided that those in the nonfully-inserted position are

consistent with the SHUTDOWN MARGIN requirements.

i

-The number of control rods permitted to be inoperable but trippable could be more than the eight allowed by the specification, but the cccurrence of eight inoperable rods could be indicative of a generic problem and the reactor must l

be shut down for investigation and resolution of the problem.

The control rod system is designed to bring the reactor subcritical at a gqq rate fast enough to prevent the MCPR from becoming less than-1-061uring the l limiting power transient analyzed in Section 15.4 of the FSAR. This analysis l shows that the negative reactivity rates resulting from the scram with the

average response of all the drives as given in the specifications, provide the >

l required protection and MCPR remains greater than 4-06, The occurrence of l p ,'s,,, scram times longer than those specified should be viewed as an indication of a

y; tale-problem with the rod drives and therefore the surveillance interval l -*-

M M Mh-.l is reduced in order to prevent operation of the reactor for long periods of time wit.h a potentially serious problem.

l The scram disch&rge volume is required to be OPERABLE so that it will be :

available when needed to accept discharge water from the control rods during a i

reactor scram and will isolate the reactor coolant system from the containment when required. ,

, Control rods with inoperable accumulators are declared inoperable and Spe-l cification 3.1.3.1 then applies. This prevents a pattern of inoperable accumu-lators that would result in less reactivity insertion on a scram than has been -

analyzed even though control rods with inoperable accumulators may still be ;

slowly scrammed via reactor pressure or inserted with normal drive water pres- l sure. 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.

GRAND GULF-UNIT 1 B 3/4 1-2 Meow &,

wey..eq,,-,.----.me.p. y,,wwy.%,.pg,wyp,--a.-y,%y+.,,,,-._n,.ywegq---Wgy-q,gpqyg:,y--- y% ,p y w- g r-wp--e,s-=..w79.-wg q.49 a-.m .ug

3/4.2 POWER DISTRIBUTION LIMITS BASES The specificati.ons 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.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the pesk cladding temperature following 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 accident-is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondar-ily on the rod to rod power distribution within an assembly. The peak clad tem-perature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value i for APLHGR is shown in Figure 3.2.1-lj 4 I The daily requirement for calculating APLHGR when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWEP, is sufficient since power distribu-tion shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate APLHGR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 stablize. The requirement for calculating APLHGR after initially determining a LIMITING CONTROL R0D PATTERN exists ensures that APLHGR will be known following a change in THERMAL POWER or power shape, that could place operation exceeding a thermal limit.

The calculational procedure used to establish the APLHGR &E- 9;;r: !

3.2.1-1 is based on a loss-of-coolant accident analysis. The analysis was per- I formed using General Electric (GE) calculational models which are consistent :

with the requirements of Appendix K to 10 CFR 50. A complete discussion of each l code employed in the analysis is presented in Reference lw Differences in this l analysis compared to previous analyses can be broken down as follows. Nna (,

a. Input Changes

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 top guide were recalculated using a more accurate technique.

3. Corrected guide tube thermal resistance,.

4. Correct heat capacity of reactor internals heat nodes.

=GW he leep op%on sed Aye 3.2.V2, Eer % q e leep oper~ Yon . The <s'enqls le p eper=%.s GmW ee i e io c cseud 4 sener- bd.Cag MW.Moa a veva % % 4.u)s loop are 'n LocA amlysk. ~~

GRAND GULF-UNIT 1 B 3/4 2-1 AMNoMEXIT U l

POWER DISTRIBUTION LIMITS BASES AVERAGE PLANAR LINEAR HEAT CENERATION RATE (Continued)

b. Model Change

1. Core CCFL pressure differential - 1 psi - Incorporate the assumption 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 the SAFE-REFLOOD pressure transfer when tha pressure is increasing was changed.

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

a. Input Change

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

b. Model Change

1. Improved Radiation and Conduction Calculation - Incorporation of CHASTE 05 for heatup calculation.

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.

3/4.2.2 APRM SETPOINTS 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 flow biased simulated thermal power-high. scram setting and flew biased simulated thermal power-upscale control rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than

' At EMY 106 or that > 1% plastic strain does not occur in the degraded situation.

LM f heT scram set'Eings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and MFLPD indicates a peak power distribution to ensure than an LHGR transient wculd not be increased in degraded conditions.

The daily requirement to verify the APRM control rod block and scram setpoints 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 verify the APRM setpoints within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 to verify the APRM setpoints once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after initially determining MFLPD to be greater than FRTP ensures that the consequences of an LHGR transient would not be increased in degraded conditions.

AmsnumaT b GRAND GULF-UNIT 2 B'3/4 2-2

POWER DISTRIBUTION LIMITS Bases Table B 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 6

Vessel Steam Output ................... 17.3 x 10 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 fL2,

b. Small Breaks 0.1 ft 2, Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATIr: DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GE0 METRY (kW/ft) ' FACTOR RATIO Initial Core 8 x 8 RP 13.4 1.4 1.174

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 Te.chnical Specification LINEAR HEAT GENERATION RATE limit.

M Dur q Mngle loop ope.rabon, dtpac kute {rens nuckede Milink is a ssume.d io occur O.1 secord (elicu$63 ike LotA, r epc cllest e( hilid M C?R..

GRAND GULF-UNIT 1 8 3/4 2-3 /lmt@mmT No. -

~

y 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 sf 1.00, and an analysis of abnormal opera- l tional transients. For any abnormal operating transient analysis evaluation 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 reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient yields the largest delta MCPR.

When added to the Safety Limit MCPR cf 1.05, the required minimum operating limit MCPR of Specification 3.2.3 is obtained =d pm:;ted in Figr 3.2.34: l The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for generation of the MCPR operating limits. g g is.,,g a p % g ,

The evaluation of a given transient beains with the system initial parameters shown in FSAR Table 15.0-2/that are input to a GE-core dynamic aehavior l transient computer program. The code used to evaluate pressurization events is described in NE00-24154(3) and the program used in non pressurization events is described in NE00-10802(2) . The outputs of this program along with the initial MCPR form the input for further analyses of the thermally limiting bundle with the single channel transient thermal hydraalic TASC code described in NEDE-25149(4) The principa? result of this evaluation is the reduction in MCPR caused by the transient.

The purpose of the MCPR f and MCPR p

of Figures 3.2.3-1 and 3.2.3-2 is 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 f and MCPR p at the existing core flow and power state. The MCPRf s are established to protect the core from inadvertent core flow increases such that the 99.9% MCPR limit requirement can be assured.

The reference core flow increase event used to establish the MCPR f is a hypothesized slow flow runout to maximum, that does not result in a scram from neutron flux overshoot exceeding the APRM neutron flux-high level (Table 2.2.1-1 item 2). With this basis the MCPRf curve is generated from a series of steady state core thermal hydraulic calculations performed at several core power and flow conditions along the steepest flow control line. This corresponds to the 105% steamflow flow control line (Figure B 3/4 2.3-1). In the actual calcula-tions a conservative highly steep generic representation of the 105% steamflow flow centrol line has been used. Assumptions used in the original calculations of this generic flow control line were consistent with a slow flow increase transient duration of several minutes: (a) the plant heat balance was assumed GRAND GULF-UNIT 1 B 3/4 2-4 AMENDMENr" M.

-_ _ _ _ _ - - L __ _ _ _ _ _ _ _ _ _ _

POWER DISTRIBUTION LIMITS l l

BASES MINIMUM CRITICAL POWER RATIO (Continued) to be in equilibrium, and (b) core xenon 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. l The first state analyzed corresponded to the maximum core power at maxi-mum core flow (102.5% of rated) after the flow runout. Several evaluations were performed at this state iterating on the normalized 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 performance were determined at other power / flow conditions on the generic flow control line, assuming the same normalized core power distribution. The result is a ,

definition of the MCPR7 performance requirement such that a flow increase event to maximum (102.5%) will not violate the safety limit. (The assumption of con-stant power distribution curing the runout power increase has been shown to be conservative. Increased negative reactivity feedback in the high power limiting bundle due to doppler and voids would reduce the limiting bund 1e relative power in an actual runout.)

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

Core power dependent setpoints are incorporated (incremental contial rod with-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 '

core thermal margins are large. However, the increased rod withdrawal requires 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 presented in GESSAR II, Appendix '5B.

! Since the severity of other (core-wide) transients at off-rated conditisns is limited by the requirement to setdown the APRM flow biased simulated thermal power-high scram trip setpoint, Specification (3.2.2), the rod withdrawal i error is the limiting transient and establishes MCPR p requirements.- h,rgser" l At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, the reactor will be operating at minimum recirculation pump speed and the modera-tor void content will be very small. For all designated control rod patterns i which may be employed at this point, operating plant experience indicates that '

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 i at 25% of RATED THERMAL POWER level with minimum recirculation pump speed, i The MCPR margin will thus be demonstrated such that future MCPR evaluation be-low this power level will be shown to be unnecessarv. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to 25% of RATED

'1 THERMAL POWER is sufficient since power distribution shif ts 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after the completion of a THERMAL POWER in- .

crease of at least 15% of RATED THERMAL POWER ensures thermal limits are met )

l GRAND GULF-UNIT 1 8 3/4 2-6 / MEN /)MMTYo. !

I o .

l INSERT B2-6 The abnormal operating transients analyzed for single loop operation are discussed in reference 5. The current MCPR limits were found to be bounding.

NochangetotheoperritingMCPRlimitisre$uiredforsingleloopoperation.

' r l

.i E

i i

a 1

p 0

t I l J14 MISC 86020502 - 12

)

l

POWER DISTRIBUTION LIMITS BASES l

MINIMUM CRITICAL POWER RATIO (Continued) after power distribution shifts while still allotting time for the power dis-tribution to stabilize. The requirement for calculating MCPR after 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.

3/4.2.4 LINEAR HEAT GENERATION RATE This specification assures that the Linear Heat Generation Rate (LHGR) in any rod is less than the design linear heat generation even if fuel pellet densification is postulated. 1 The daily requirement for calculating LHGR when THERMAL POWER is greater than or equal tc 25% of RATED THERMAL POWER is sufficient since power distri-bution shifts are very slow when there have not been significant power or control rod changes. The requirement to calculate LHGR within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months 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 LHGR after initially determining a LIMITING CONTROL ROD PATTERN exists ensures that LHGR will be known following a change in THERMAL POWER or power shape that could place operation exceeding a themal limit.

References:

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

2. R. B. Linford, Analytical Methods of Plant Transient Evaluations for i the GE BWR, February 1973 (NEDO-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. CG45 h jor "Pe.rk,r m ne.e. lawprevtmeed 'P< g re g b gle, t.ee}

operaks Arm.I 1 %, wo sua.m MIArpr+- febe%1%.

6. GeeroA Elec.lrit Cs.mpc.ng AnctlyMcd N1octe.1 ge L ess . .{ .

c la.<d A%lpis in Acc.orciano A loc 5 RSo, Appudh )(-

Ame=4,4en4 2, one hiirculo4 ion L.e.p oJ -.h hryA , ,

i Hedo 2ost,t.- 2,1t vb'% i , LW . 89 T8.

GRAND GULF-UNIT 1 B 3/4 2-7 //AENb AENT 80,

INSTRUMENTATION b BASES 3/4.3.9 TURBINE OVERSPEED PROTECTION This specific & tion is provided to ensure that the turbine overspeed protection instrumentation and the turbine speed control valves are OPERABLE and will protect the turbine from excessive overspeed. Protection from turbine excessive overspeed is required since excessive overspeed of the turbine could generate potentially damaging missiles which could impact and damage safety-related components, equipment or structures.

3/t 3.lO N EuTROM Fi.o x MoNITOCTN6 IN+TRUMEN T4 MON IN SEAT "b*

C.

k i GRAND GULF-UNIT 1 8 3/4 3-7 AMENoMEnr No, l

INSERT "D" RASES 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION 1

This specification is to assure that neutron flux limit cycle oscillations ]

are detected and suppressed.

)

In order to identify a region of the operating map where surveillance should be performed, stability tests at operating plants were reviewed. To account for variability a conservative decay ratio of 0.6 was chosen as the basis for defining the region of potential instability. The resulting region corresponds to core flow less than 45% of rated and THERMAL POWER greater than the 80% rod line. The 80% rod line is illustrated in Figure 3.4.1.1-1.

j Neutron flux noise limits are also established to ensure the early detection of limit cycle oscillations. Typical APRM neutron flux noise levels at up to 12% of rated power have been observed. These levels are easily bounded by values considered in the thermal / mechanical fuel design. Stability tests have shown that limit cycle oscillations result in peak-to-peak

, magnitudes of 5 to 10 times the typical velues. Therefore, actions taken to 4 suppress flux oscillations exceeding three (3) times the typical value are sufficient to ensure early detection of limit cycle oscillations. The specification includes the surveillance requirement to establish the requisite baseline noise data and prohibits operation in the region of potential instability if the appropriate baseline data is unavailable.

i j J14 MISC 86020502 - 7 .

3/4.4 REACTOR COOLANT SYSTEM BASES 3/4.4.1 RECIRCULATION SYSTEM ~

2.ussu- d '

Bus 5A.4.1 Operation with one reactor core coolant recirculation loop inoperable Es-pr:hibited unti! :n : :h:ti:n of the perfor :n;; cf th: ECCS during On hp .

Oper: tier h:: beer perf;ned, evc h:ted :nd detcraincd t; bc ::: pt:bk. _

nu%%o log open. Mon,-

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, the requirement for shutd a n of the facility with a jet pump inoperable. Jet pump failure can be detected by monitoring jet pump per- i formance on a prescribed schedule for significant degradation.4 )(ecirculatica I loop flow mismatch limits are in compliance with ECCS LOCA analysis design criteria. The limits will ensure an adequate core flow coastdown from either recirculation loop following a LOCA m

= 'IusuT 'C"(nsa r: stat **h l In order to prevent undue stress on the vessel nozzles and bottom head region, 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 of the reactor pressure vessel coolant temperature to prevent thermal shock to the recirculation pump and recirculation nozzles. Since the coolant in the bottom of the vessel is at a lower temperature than the coolant in the upper regions of the core, undue stress on the vessel would result if the temperature difference was greater than 100 F.

gg y The recirculation flow control valves provide regulation of individual recirculation loop drive flows; which, in turn, will vary the flow rate of coolant through the reactor core over a range consistent with the rod pattern and recirculation pump speed. The recirculation flow control system consists of the electronic and hydraulic components necessary for the positioning of the two hydraulically actuated flow control valves. Solid state control logic will generate a flow control valve " motion inhibit" signal in response to any one of several hydraulic power unit or analog control circuit failure signals.

The " motion inhibit" signal causes hydraulic power unit shutdown and hydraulic isolation such that the flow control valve fails "as is." This design feature insures that the flow control valves do not respond to potentially erroneous control signals.

Electronic limiters exist in the position control loop of each flow control valve to limit the flow control valve stroking rate to 10i1% per second in the opening and closing directions on a control signal failure. The analysis of the recirculation flow control failures on increasing and decreasing flow are presented in Sections 15.3 and 15.4 of the FSAR respectively.

The required surveillance interval is adequat'e to ensure that the flow control valves remain OPERABLE and not so frequent as to cause excessive wear on the system components.

In cases where he am AA lim;h ceoot be ment ='.a.d. conmued ope d.% __,,_,,.

b per4%d w3h one loop;n op. cod.on . _

GRAND GULF-UNIT 1 B 3/4 4-1 AMsuonent No.

l INSERT "A" BASES 3/4.4.1

...has been evaluated and found to remain within design limits and safety margins provided certain limits and setpoints are modified. The "GGNS Single Loop Operation Analysis" identified the fuel cladding integrity Safety Limit, MAPLHGR limit and APRM setpoint modifications necessary to maintain the same margin of safety for single loop operation as is available during two loop operation. Additionally, loop flow limitations are established to assure vessel internal vibration remains within limits. A flow control mode restriction is also incorporated to reduce valve wear due to automatic flow control attempts and to ensure valve swings into the cavitation region do not occur.

INSERT "B" During single loop operation, the condition may exist in which the coolant in the bottom head of the vessel is not circulating. These differential temperature criteria are also to be met prior to power or flow increases from this condition.

INSERT "C" In accordance with BWR thermal hydraulic stability recommendations, operation above the 80% rod line with flow less than 39% of rated core flow is restricted.

J14 MISC 86020502 - 6

__

ML20155B085

Text

2.0 INTRODUCTION

References:

Navigation menu

Search