Text

Proposed Tech Specs 3.2.4,3.3.1.1,3.3.1.3,3.4.1,5.6.5, B 3.2.4,B 3.3.1.1,B 3.3.1.3 & B 3.4.1,enabling Full Implementation of Enhanced Option I-A Long Term Solution to Address Neutronic/Thermal Hydraulic Instability Issue
	ML20236T166
Person / Time
Site:	Grand Gulf
Issue date:	07/20/1998
From:	; ENTERGY OPERATIONS, INC.
To:	;
Shared Package
ML20236T158	List: ML20236T155; ML20236T166;
References
	NUDOCS 9807270470
	Download: ML20236T166 (53)
	v • d • e

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ATTACHMENT 3 to GNRO-98/00053 Marked-up TS Pages I

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9007270470 990720 PDR ADOCK 05000416':

P PDR

l Definitions ,

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i 1.1 Definitions !

DOSE EQUIVALENT I-131 be those listed in Table III of TID-14844, (continued) AEC, 1962, " Calculation of Distance Factors for !

Power and Test Reactor Sites."

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EMERGENCY CORE C001.ING The ECCS RESPONSE TIME shall be that time interval l SYSTEM (ECCS) RESPONSE from when the' monitored parameter exceeds its ECCS i TIME initiation setpoint at the channel sensor until the ECCS equipment is capable of performing its l l safety function (i.e., the valves travel to their !

required positions, pump discharge pressures reach !

their required values, etc.). Times shall include !

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' diesel generator starting and sequence loading delays, where applicable. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured. i END OF CYCLE The EOC-RPT SYSTEM RESPONSE TIME shall be that l RECIRCULATION PUMP TRIP time interval from initial movement of the 1 l (EOC-RPT) SYSTEM RESPONSE associated turbine stop valve or the turbine TIME control valve to complete suppression of the electric arc between the fully open contacts of the recirculation pump circuit breaker. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured, except for the breaker arc suppression time, which is not measured but is validated to conform to the manufacturer's design value.

N ISOLATION SYSTEM The ISOLATION SYSTEM RESPONSE TIME shall be that i

RESPONSE TIME time interval from when the monitored parameter L

exceeds its isolation initiation setpoint at the channel sensor until the isolation valves travel '

gf_gg) y to their required positions. The response time may be measured by means of any series of sequential, overlapping, or total steps so that the entire response time is measured.

1 l La The maximum allowable primary containment leakage rate, L , shall be 0.437% of primary containment air weight per day I containment pressurate (Pths

)- calculated peak a

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! l l j GRAND GULF 1.0-3 Amendment No. 120 l

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INSERT TO TS PAGE 1.0-3 i Fraction of The FCBB shall-be the ratio of the power Core Boiling generated in the lower 4 feet of the active reactor Boundary (FCBB) core to the power required to produce bulk saturated boiling of the coolant entering the fuel channels.

The core boiling boundary is the axial elevation of i core average bulk saturation above the bottom of the active reactor core. j O

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FCTB ed 6C [cM 3.2.4 l

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_ .k_ Ld.D ACTIONS (continued) 3.2 POWER DISTRIBUTION LIMITS 3.2.4 Fraction of Core Boiling Boundary (FCBB)

I LCO 3.2.4 The FCBB shall be 5 1.0.

I APPLICABILITY: THERMAL POWER and core flow in the Restricted Region as specified in the COLR,

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l MODE 1 when RPS Function 2.d, APRM Flow Biased Simulated Thermal Power - High, Allowable Value is " Setup" as specified in the COLR.

ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME i

A. FCBB not within limit A.1 Restore FCBB to within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months for reasons other than ~

limit.

an unexpected loss of )

feedwater heating or unexpected reduction I in core flow, i

j B. Required Action and B.1 Initiate action to exit Immediately associated Completion the Restricted Time of Condition A Region.

not met. AND OR Immediately B.2 Initiate action to return following exit i

NOTE--------- APRM Flow Biased Required Action B.1 of Restricted Simulated Thermal Region and Required Action Power - High B.2 shall be completed Allowable Value to if this Condition is "non-Setup" value.

entered due to an unexpected loss of feedwater heating or unexpected reduction in core flow.

FCBB not within limit due to an unexpected loss of feedwater l heating or unexpected l reduction in core

. flow.

l-GRAND GULF 3.2-3b Amendment No. __

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(* W IM FCBB foge 2 d.2 1 324 SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY l SR 3.2.4. -------------------NOTE--------------------

Not required to be performed until 15 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months I minutes after entry into the Restricted AND

! Region if entry was the ret'lt of an Once within 15 unexpected transient. minutes l .......__...........__..__.........__...... following l Verify FCBB 5 1.0. unexPeeted transient l

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GRAND GULF 3.2-3c Amendment No. _

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RPS Instrumentation 3.3.1.1 f

SURVEILLANCE REQUIREMENTS- (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.5 Verify the source range monitor (SRM) and Prior to intermediate range monitor (IRM) channels withdrawing overlap. SRMs from the fully inserted position SR 3.3.1.1.6 ------------------NOTE-------------------

Only required to be met during entry into i MODE 2 from MODE 1. !

Verify the IRM and APRM channels overlap. 7 days SR 3.3.1.1.7 Calibrate the local power range monitors. -1000 MWD /T i

i average core exposure SR 3.3.1.1.8 Perform CHANNEL FUNCTIONAL TEST. 92 days SR 3.3.1.1.9 Calibrate the trip units. 92 days SR 3.3.1.1.10 ------------------NOTES------------------

1, Neutron detectors are excluded.

2. For Function 2.a, not required to be performed when entering MODE 2 from MODE 1 until 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE

>.........2................... ............

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Perform CHANNEL CALIBRATION. 184 days f -- -

l J. For Function 2.d, AFRM recirculat16D 110w u transmitters are excluded. (continued)

4. For Function 2.d, the digital components of GRAND GULF the flow control trip reference cards are e cluded.

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RPS Instrumentation 3.3.1.1 SURVEILLMCE REQUIREMENTS (continued)

SURVEILLANCE FREQUENCY SR 3.3.1.1.11 Perfom CHAfstEL FUNCTIONAL TEST. 18 months SR 3.3.1.1.12 - - - - -

NOTES - - - - - ---

1. Neutron detectors are excluded.

2. For IMs, not required to be perfomed when entering MODE 2 from MODE 1 untti 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after entering MODE 2.

Perfom CHANNEL CALIBRATION. 18 months ,

SR 3.3.1.1.13 Perfom LOGIC SYSTEM FUNCTIONAL TEST. 18 months SR 3.3.1.1.14 Verify Turbine Stop Valve Closure, Trip 18 months 011 Pressure-Low and Turbine control Valve Fast Closure Trip 011 Pressure-Low Functions are not bypassed when THERMAL POWER is 2 405 RTP.

SR 3.3.1.1.15 ------------------NOTES------------------

1. Neutron detectors are excluded.

2. For Functions 3, 4, and 5 in Table 3.3.1.1-1, the channel sensors may be excluded.

l 3. For Function 6, "n" oguals 4 channels for the purpose of detemining the STAGGEkED TEST BASIS Frequency.

Verify the RPS RESPONSE TIME is within 18 months on a m limits. STAGGERED TEST BASIS ll'N$ER7~

L D GRAND GULF

' J 3.3-5 Amendment No MG,133

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SEh ' 2 INSERT TO PAGE 3.3-5 j# #

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SR 3.3.1.1.16 Verify the simulated thermal power time 18 months constant.

SR 3.3.1.1.17 Perform APRM recirculation flow 18 months transmitter calibration.

SR' 3.3.1'.1.18 Adjust the flow control trip reference Once within 7 ,

card to conform to. reactor flow, days after reaching

! equilibrium conditions following refueling l

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RPS Instrumentation 3.3.1.1 i

Table 3.3.1.1-1 (page 1 of 3)

Reactor Protection system Instrumentation COMITIONS APPLICABLE REQUIRED REFERENCED MWES OR OTHER CNANNELS FRON PER TRIP REGUIRED SURVEILLANCE ALLOWABLE SPECIFIED SYSTEM ACTION D.1 REQUIREMENTS VALUE FUNCTION CONDITIONS

1. Intermodlate Range Monitors

a. Neutron Flux.- Nigh 2 3 N st 3.3.1.1.1 5 122/125 st 3.3.1.1.3 divisions st 3.3.1.1.5 of futt at 3.3.1.1.6 scale SR 3.3.1.1.12 ,

sa 3.3.1.1.13 j 5(83 3 I st 3.3.1.1.1 s 122/125 st 3.3.1.1.4 divisions sa 3.3.1.1.12 of full st 3.3.1.1.13 scale

b. Insp 2 3 N st 3.3.1.1.3 NA st 3.3.1.1.13 5(a) 3 g g, 3,3,g,g,4 ,g

- sa 3.3.1.1.13 ;

2. Aversee Power Range Monitors l

a. Neutron Flux - High, 2 3 N st 3.3.1.1.1 s 201 RTP Setdown SR 3.3.1.1.3 st 3.3.1.1.6 st 3.3.1.1.7 sa 3.3.1.1.10 SR 3.3.1.1.13

b. Fixed Neutron 1 3 C st 3.3.1.1.1 5 1201 RTP Flux - Nigh SR 3.3.1.1.2 SR 3.3.1.1.7 st 3.3.1.1.8 st 3.3.1.1.10 st 3.3.1.1.13 st 3.3.1.1.15

c. Inop 1,2 3 N sa 3.3.1.1.7 NA st 3.3.1.1.8 I

st 3.3.1.1.13 l (continued) l l

(a) With any centrol rod withdrawn from a core cell containing one or more fuel assemblies, w __ __ - - - _ - _ _ _ _ _

_ ~ - _ _ - - -- ___

(b) Attowebte Values specified in the COLR. Allowebte Value modification required by the COLR due to reductions in feedwater temperature may be delayed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . - - -

3 C SR 3.3.1.1.1 (b) 7

. Low Blased Simulated 1 Thermal Power - Nigh SR 3.3.1.1.2 [

SR 3.3.1.1.7 %

SR 3.3.1.1.8 SR 3.3.1.1.10 SR 3.3.1.1.13 SR 3.3.1.1.15 }

st 3.3.1.1.16 SR 3.3.1.1.17 j SR 3.3.1.1.18 7 l GRAND GULF 3.3-6 Amendment No. 120

P!DS 3.3.1.3 gQ &

e d3 3.3 INSTRUMENTATION 3.3.1.3 Period Based Detection System (PBDS)

LCO 3.3.1.3 One channel of PBDS instrumentation shall be OPERABLE.

ANL Each OPERABLE channel of PBDS instrumentation shall not indicate Hi-Hi DR alarm.

APPLICABILITY: THERMAL POWER and core flow in the Restricted Region specified in the COLR, THERMAL POWER and core flow in the Monitored Region specified in the COLR.

ACTIONS' CONDITION REQUIRED ACTION COMPLETION TIME A. Any OPERABLE PBDS A.1 Place the reactor mode Immediately channel indicating switch in the Hi-Hi DR Alarm, shutdown position.

B. Required PBDS channel B.1 ---------NOTE--------Only inoperable while in applicable if RPS Immediately the Restricted Region. Function 2.d, APPJ4 Flow Biased Simulated Thermal Power - High, Allowable value is

" Setup".

Initiate action to exit the Restricted Region.

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PBDS 3212 i

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CONDITION REQUIRED ACTION COMPLETION TIME i B. -(continued) B.2 Place'the reactor mode Immediately switch in the i shutdown position.

C. Required PBDS channel C.1 Initiate action to exit 15 minutes i inoperable while in the Monitored Region.

the Monitored Region. l SURVEILLANCE REQUIREMENTS l

SURVEILLANCE FREQUENCY SR 3.3.1.3.1 Verify each OPERABLE channel of PBDS 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months instrumentation,not in Hi-Hi DR Alarm. j i

SR 3.3.1.3.2 Perform CHANNEL CHECK. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months !

SR 3.3.1.3.3 Perform CHANNEL FUNCTIONAL TEST. 24 months l GRAND GULF 3.3-13b l Amendment No. _

Recirculation Loops Operating 3.4.1 3.4 REACTOR COOLANT SYSTEM (RCS) 3.4.1 Recirculation Loops Operating i

LCO 3.4.1 Two recirculation loops with matched flows shall be in A _ _.

operation.p ih ...., wre ;s ;,.. ,,;.. m ;w;.r,:n ":;ia- 0

-' riger: .'. er " vier, C ;f Fig.re 0.'. I derie; raat el red -ithd e 21 f:r :tirtea_

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(Onerecirculationloopshallbeinoperationwith%t:1 ce, e

= ::: ;n;.... n r- perr: n,.....,.2 1 ;r .

( D-1:n C Of "fi; 7';e 2 ' ' "" '- :: tre' red dthdrir_-E-h r.'equired limits modified for single loop

-- -+--'"- l operation as specified in the COLR.

NOTE-----------------------------

Required limtt modifications for single recirculation loop operation may be delayed for up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after transition from two recirculation loop operation to single recirculation loop operation.

i APPLICABILITY: MODES 1 and 2.

l ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME A. Recirculation loop jet A.1 Shutdown one 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months pump flow mismatch not recirculation loop.

within limits.

(continued) l l

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GRAND GULF 3.4-1 Amendment No. 120

Recirculation Loops Operating 3.4.1 KTIONS (continued) 87 j CONDITION REQUIRED ACTION COMPLETI[ TIME 1

B. Total ore flow as a B.1 Restore total core 2 ho s !

functio of THERMAL flow as a function of POWER wit n Region B. THERMAL POWER to within Region D.

AND APRM neutron flu ;

noise level s 10% l peak-to-peak of RTP.

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C. Total core flow as a C.1 Restore otal core 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months function of THERMAL flow a function of POWER within Region C .. E L POWER to when not withdrawing w in Region D. 1 control rods for startup.

AND ,

1 APRM neutron flux l noise level s 10%

peak-to-peakofRTP./

D. Total core flow a D.1 Place the reactor Immediately function of THE L mode switch in the POWER within gion 8 shutdown position.

or Region C en not withdrawin control rods for artup.

AND l AP neutron flux i

no e level > 10%

p ak-to-peak of RTP.

(continuM)

I-GRAND GULF 3.4-2 Amendment No. 120

Recirculation Loops Operating 3.4.1 ACTIONS (continued)

{

j CONDITION REQUIRED ACTION COMPLETION TIME E. Total ow as a E.1 Place the reactor pi function of T mode switch in the POWER within Region A. shutdown pos .

9E No recirculation loo in operation MODE 1

[ No recirculation loops Be in MODE 3. 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months in operation in MODE 2.

O M Required limit Declare associated Immediately modifications not limit (s) not met, performed.

GRAND GULF 3.4-3 Amendment No. 120

Recirculation Loops Operating 3.4.1 I

l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.4.1.1 ..-..--...-......--N0TE--------------------

Not required to be performed until 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both recirculation loops are. in operation.

Verify recirculation loop jet pump flow 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months mismatch with both recirculation loops in operation is:

a. s IP." of rated core flow when open ting at < 70% of rated core flow; and

b. s 5% of rated core flow when operating at a: 70% of rated. core flow.

SR 3.4.1.2 Veri y f 34-h urs THERMAL POWER to e flow asRegion a function DJ o#

Figure 3.4.1-1 or Regio 3.4.1 1 duri rod withdrawa s I

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l GRAND GULF 3.4 4 Amendment No. 120 i

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0 9 0 0 1 1 8 7 6 5 4 3 2 1 mW3o aekW[ oWF$ u.o $wu$'

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. Reporting Requirements 5.6 1

i 5.6 Reporting Requirements

! 5.6.2 Annual. Radiological Environmental Operatino Report (continued) results are not available for inclusion with the report, the report shall be submitted noting and explaining the reasons for the missing results. The missing data shall be submitted in a supplementary report as soon as possible, l

5.6.3 Radioactive Effluent Release Report !

The Radioactive Effluent Release Report covering the operation of j the unit during the previous calendar year shall be submitted by i May 1 of each year. The report shall include' a summary of the quantities of radioactive liquid and gaseous effluents and solid waste released from the unit. The material provided shall be ,

consistent with the objectives outlined in the ODCM and process i control program and in conformance with 10 CFR 50.36a and '

10 CF'l 50, Appendix I,Section IV.B.I.

5.6.4 Monthly Operatino Reports Routine reports of operating statistics and shutdown experience, including documentation of all challenges to the main steam safety / relief valves, shall be submitted on a monthly basis no later than the 15th of each month following the calendar month covered by the report. /!

5.6.5 Core Operatino limits Report (COLR)

a. Core operating limits shall be established prior to each

. - reload cycle, or prior to any remaining portion of a reload cycle, and shall be documented in the COLR for the following:

1) LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR),

2) LC0 3.2.2, Minimum Critical Power Ratio (MCPR), end -

3) LCO 3.2.3, Linear Heat Generation Rate (LHGR) y 1 -

(continued)_,

0,,,

p3o; tab 13a-dary C&

[ D L.co 3. 3 l . I , RPs hsdrunen oudfa wr,4 4)LCO 3. 3. l. 3, Period Cra os),

Sa se e l Oe-lec/Un Sysbom l

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GRAND GULF 5.0-18 Amendment No. 120

, Reporting Requirements 5.6 i

5.6 Reporting Requirements 5.6.5 Core Operatina Limits Report (COLR) (continued)

10. XN-NF-85-74(P)(A), "RODEX2A (BWR): Fuel Rod Thermal-Mechanical Response Evaluation Model," Exxon Nuclear Company, . Inc., Richland, WA.

I

11. XN-CC-33(P)(A), "HUXY: A Generalized Multirod Heatup Code with 10CFR50 Appendix K Heatup Option," Exxon Nuclear i

Company, Inc., Richland, WA.

g 0 12. XN-NF-825(P)(A), "BWR/6 Generic Rod Withdrawal Error Analysis, MCPR for Plant Operation Within the Extended j ( H, Operating Doma,in," Exxon Nuclear Company, Inc., Richland, WA.

D- t f

bd4 13. XN-NF-81-51(P)(A), "LOCA-Seismic Structural Response of an Exxon Nuclear Company BWR Jet Pump Fuel Assembly,"

Exxon Nuclear Company, Inc., Richland, WA.

O%* {

NSG d 14. XN-NF-84 '97(P)(A), "LOCA-Seismic Structural Response of 4

an ENC 9x9 BWR Jet Pump Fuel Assembly," Advanced Nuclear I

Rg Fuels Corporation, Richland, WA.

.'q-

@tC 15. XN-NF-86-37(P), " Generic LOCA Break Spectrum Analysis for BWR/6 Plants," Exxon Nuclear Company, Inc., Richland, WA.

4 f l q 16. XN-NF-82-07(P)(A), " Exxon Nuclear Company ECCS Cladding 4& [K Swelling and Rupture Model," Exxon Nuclear Company, Inc.,

l 4df j

Richland, WA.

17. XN-NF-80-19(A), Volumes 2, 2A, 2B, & 2C, " Exxon Nuclear

{ t

\y5 / Methodology for Boiling Water Reactors EXEM BWR ECCS l Evaluation Model," Exxon Nuclear Company, Inc., Richland, WA. j j s i T 18. XN-NF-79-59(P)(A), " Methodology for Calculation for

^

d i Pressure Drop in BWR Fuel Assemblies," Exxon Nuclear Company, Inc., Richland, WA.

i m- I M

q j *19. NEDE-24011-P-A, General Electric Standard Application for m i Reactor Fuel (GESTAR-II) with exception to the misplaced i

{ fuel bundle analyses as discussed in GNR0-96/00087 and f G- l the generic MCPR Safety Limit analysis as discussed in 1

@ GNRO-96/00100, letters from C. R. Hutchinson to USNRC. i

) *20. J11-02863SLMCPR, Revision 1, "GGNS Cycle 9 Safety Limit MCPR Analysis."

] ;

(continued)

Items 19 and 20 of TS 5.6.5.b are applicabile only for Cycle 10 operation. l<

' GRAND GULF 5.0-20 Amendment No. 120,I??, 136

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LHGR B 3.2.3 BASES ,

1 ACTIONS A.1 (continued)

I restore the LHGR(s) to within its required limit (s) such that the plant is operating within analyzed conditions and within the design limits of the fuel rods. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is normally sufficient to restore the LHGR(s low pro)bability of a transient or Design Basis Acciden occurring simultaneously with the LHGR out of specification.

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),d If the LHGR cannot be restored to within its required limit

! within the associated Completion Time, the plant must be brought to a MODE or other specified condition in which the LC0 does not apply. To achieve this status, THERMAL POWER !

must be reduced to < 25% RTP within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months . The allowed -

l Completion Time is reasonable, based on operating !

experience, to reduce THERMAL POWER to < 25% RTP in an i orderly manner and without challenging plant systems.

SURVEILLANCE SR 3.2.3.1 REQUIREMENTS The LHGRs are required to be initially calculated within ,

12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after THERMAL POWER is = 25% RTP and then every !

24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months thereafter. They are compared with the specified ^

limits in the COLR to ensure that the reactor is operating within the assumptions of the safety analysis. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and recognition of the slowness of changes in power distribution under normal conditions. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months allowance after !

THERMAL POWER = 25% RTP is achieved is acceptable given the large inherent margin to operating limits at lower power levels.

REFERENCES 1. UFSAR, Chapter 15.

2. UFSAR, Chapter 4.

3. NUREG-0800, " Standard Review Plan," Section 4.2, i

II.A.2(g), Revision 2, July 1981.

_TNSEPT Nt%)

L SecJion 83.2 7 l .

GRAND GULF B 3.2-11 Revision No. 2

FCBB

{

P 3.2.4

& W W fgfW D E 3, 2-/ .

B 3.2 POWER DISTRIBUTION LIMITS # 0 B 3.2.4 Fraction of Core Boiling Boundary (FCBB)

BASES

)

BACKGROUND General Design Criterion 12 requires protection of fuel thermal safety limits from conditions caused by neutronic/ thermal-hydraulic instability. Neutronic/ thermal-hydraulic instabilities result in power oscillations which could result in exceeding the MCPR Safety Limit (SL) . The MCPR SL ensures that at least 99.9% of the fuel rods avoid boiling transition during normal operation and during an anticipated operational occurrence (ADO) (refer to the Bases for SL 2.1.1.2).

The FCBB is the ratio of the power generated in the lower 4 ;

feet of the active reactor core to the power required to I produce bulk saturated boiling of the coolant entering the l fuel channels. The value of 4 feet above the bottom of the I active fuel is set as the boiling boundary limit based on analysis described in Section 9 of Reference 1. The boiling boundary limit is established to ensure that the core will remain stable during nornal reactor operations in the Restricted Region of the power and flow map defined in the COLR which may otherwise be susceptible to neutronic/ thermal-hydraulic instabilities and therefore the MCPR SL remains protected.

Planned operation in the Restricted Region is accommodated by manually establishing the " Setup" values for the APRM Flow-Biased Simulated Thermal Power - High Scram and APRM Flow-Biased Neutron Flux - Upscale Control Rod Block functions. The " Setup" Allowable Values of the APRM Flow-Biased Thermal Power - High Function (refer to LCO 3.3.1.1, Table 3.3.1.1-1, Function 2.d.) are consistent with assumed operation in the Restricted Region with FCBB $ 1.0.

Operation with the " Setup" values enables entry into the Restricted Region without a control rod block that would otherwise occur. Plant operation with the " Setup" values is limited as much as practical due to the effects on plant operation required to meet the FCBB limit.

(continued)

GRAND GULF B 3.2-11a l LDC 98037 i_

FCBB gj B 3.2.4

\

TA]SFRr 12>

Pee B 3. 24 '

BASES (continued) h[ 7 APPLICABLE The analytical methods and assumptions used in establishing SAFETY ANALYSES the boiling boundary limit are presented in Section 9 of Reference 1. Operation with the FCBB 5 1.0 (i.e., a bulk saturated boiling boundary 2 4 feet) is expected to ensure that operation within the Restricted Region will not result in neutronic/ thermal-hydraulic instability due to either steady-state operation or as the result of an ADO which initiates and terminates entirely within the Restricted Region. Analysis also confirms that AOOs initiated from outside the Restricted Region (i.e., without an initial restriction on FCBB) which terminate in the Restricted Region are not expected to result in instability. The types of transients specifically evaluated are loss of flow and coolant temperature decrease which are limiting for the onset of instability (Ref.1)

Although the onset of instability does not necessarily occur if the FCBB is greater than 1.0 in the Restricted Region, bulk saturated boiling at the 4 foot boiling boundary limit has been adopted to preclude neutronic/ thermal-hydraulic instability during operation in the Restricted Region. The effectiveness of this limit is based on the demonstration (Ref. 1) that with the limit met large margin to the onset of neutronic/ thermal-hydraulic instability exists and all major state parameters that affect stability have relatively small impacts on stability performance.

The FCBB satisfies Criterion 2 of the NRC Policy Statement. l i

LCO I Requiring FCBB 5 1.0 ensures the bulk coolant boiling boundary is 2 4 feet from the bottom of the active core.

)

Analysis (Ref. 1) has shown that for anticipated operating 1 conditions of core power, core flow, axial and radial power shapes, and inlet enthalpy, a boiling boundary of 4 feet ensures variations in these key parameters do not have a significant impact on stability performance.

Neutronic/ thermal-hydraulic instabilities result in power oscillations which could result in exceeding the MCPR Safety Limit (SL). The MCPR SL ensures that at least 99.9 % of the fuel rods avoid boiling transition during normal operation and during an ADO (refer to the Bases for SL 2.1.1.2).

(continued)

GRAND GULF B 3.2-11b LDC 98037

FCCB B 3.2.4

,aa, Sae n QSf/V BASES (continued) c l p ,fa 3. 2 "

l l

. APPLICABILITY- The FCBB. limit is used to prevent core conditions necessary for the onset of instability and thereby preclude l- neutronic/ thermal-hydraulic instability while operating in the Restricted Region defined in the COLR. 4 l' The boundary of the Restricted Region in the Applicability i

,- of this LCO is analytically established in terms "" thermal power and core flow. The Restricted Region is defined by the APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints, which are a function of reactor recirculation drive flow. The Restricted Region Entry Alarm {

-(RREA) signal is generated by the Flow Control Trip Reference (FCTR) card using the APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints. As a result, the RREA is. coincident with the Restricted Region Boundary when the setpoints are not " Setup," and provides indication l

of entry into the Restricted Region. However, APRM Flow Biased Neutron Flux - Upscale Control Rod Block signals provided by'the FCTR card, that are not coincident with the Restricted Region boundary, do not generate a valid RREA.

The Restricted Region boundary for this LCO applicability is specified in the COLR.

The PCBB limit is also used to ensure that core conditions, while operating with " Setup" values, remain consistent with analyzed transients initiated from inside and outsida the Restricted Region.

When the APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints are " Setup" the applicable setpoints used to' generate the RREA are moved to the interior boundary of the

"' Restricted Region to allow controlled operation within the Restricted Region. While the setpoints are " Setup" the Restricted Region boundary remains defined by the normal

("non-Setup") APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints.

Parameters such as reactor power and core flow available at the reactor controls may be used to provide immediate confirmation that entry into the Restricted Region could reasonably have occurred.

Operation outside the Restricted Region is not susceptible to neutronic/ thermal-hydraulic instability when applicable thermal power distribution limits such as MCPR are met.

l (continued)

GRAND GULF B 3.2-11c LDC 98037 I w - - _ - _ _ _ _ _ _ . _ _ _ _ - _ _

FCBB

< UE"W < &<

.L&&

B 3. 2 - / /

BASES (continued) fhy"6yj j gf/ eqo 7 ACTIONS A.1 If FCBB is not within the required limit, core conditions necessary for the onset of neutronic/ thermal-hydraulic instability may result. Therefore, prompt action should be taken to restore the FCBB to within the limit such that the stability of the core can be assured. Following uncontrolled entry into the Restricted Region (i. e.,

operation in the restricted region without the APRM Flow Biased Simulated Thermal Power-High Function Setup ) ,

prompt restoration of FCBB within limit can be expected if FCBB is known to not significantly exceed the limit.

Therefore, efforts to restore FCBB within limit following an uncontrolled entry into the Restricted Region are appropriate if operation prior to entry was consistent with planned entry or the potential for entry was recognized as demonstrated by FCBB being monitored and known to not significantly exceed the limit. Actions to exit the Restricted Region are appropriate when FCBB can not be expected to be restored in a prompt manner.

Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on engineering judgment as to a reasonable time to restore the FCBB to within limit. The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is acceptable based on the availability of the PBDS per Specification 3.3.1.3, " Period Based Detection System" and the low probability of a neutronic/ thermal-hydraulic instability event.

B.1 and B.2 Changes in reactor core state conditions resulting from an unexpected loss of feedwater heating or reduction in core flow--any unexpected reduction in feedwater temperature, recirculation pump trip, recirculation pump down shift to slow speed, or significant flow control valve closure (small changes in flow control valve position are not considered significant)--require immediate initiation of action to exit the Restricted Region and return the APRM Flow Biased Simulated Thermal Power - High Function (refer to LCO 3.3.1.1, Table 3.3.1.1-1, Function 2.d.) to the "non-Setup" l value. Condition B is modified by a Note that specifies l that Required Actions B.1 and B.2 must be completed if this j Condition is entered due to an unexpected loss of feedwater 1 (continued) t GRAND GULF B 3.2-11d LDC 98037 !

j l

j

FCBB

5 3.2.4 f)V$k l:

7 ) Jf d b 0 ' '" j f

psp. r J ~1 l ACTIONS B.1 and B.2 (continued) heating or reduction in core flow. The completion of Required Actions B.1'and B.2 is required even though FCBB may be calculated and determined to be within limit. Core conditions. continue to change after an unexpected loss of

^

feedwater heating or reduction in core flow due to transient induced changes with the potential that the FCBB may change and the limit not be met. The potential for changing core conditions, with FCBB not met,.is not consistent with operation in the Restricted Region or with the APRM Flow Biased Simulated Thermal Power - High Funct in " Setup".

Therefore, actions to exit the Restricted Region and return the APRM Flow Biased Simulated Thermal Power - High function to the "non-Setup" value are required to be completed in the event Condition B is entered due to an unexpected loss of feedwater heating or an unexpected reduction in core flow.

If operator actions to restore the FCBB to within limit are not successful within the specified Completion Time of Condition A, reactor operating conditions may be changing and may continue to change such that core conditions necessary for the onset of neutronic/ thermal-hydraulic instability may be met. Therefore, in the event.the Required A'c tion and associated Completion Time of Condition A is not met, immediate action to exit the Restricted Region and return the APRM Flow Biased Simulated Thermal Power -

High Function to the "non-Setup" value is required.

Exit of the Restricted Region can be accomplished by control rod insertion and/or recirculation flow increases. Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. The time required to exit the Restricted Region will depend on existing plant conditions. Provided efforts are begun without delay and continued unti1~the Restricted Region is exited, operation is acceptable.

l (continued)

GRAND GULF B 3.2-11e LDC 9803'1

i4 o

FCBB Y rl ypp- Gy 18 3,2 -/ /

BASES (continued) / hI' M

SURVEILLANCE SR 3.2.4.1 l- REQUIREMENTS l Verifying FCBB $ 1.0 is required to ensure the reactor is 1

operating within the assumptions.of the safety analysis.

The boilintj boundary limit is established to ensure that the

^ core will remain stable during normal reactor operations in

! the Reptri.cted Region of the power and flow map defined in the CCLR vhich may otherwise be susceptible to neutrenic/ thermal-hydraulic instabilities.

FCBB is required to be verified every 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months while operating in the Restricted Region defined in the COLR. The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is based on both engineering judgment and i recognition of the slow rate of change in power distribution during normal. operation. l The second Frequency requires FCBB to be within the limit within 15 minutes following an unexpected transient. The verification of the FCBB is required as a result of the possibility that the unexpected transient results in the limit not being met. The 15 minute frequency is based on both engineering judgment and the availability of the PBDS to provide the operator with information regarding the l

potential imminent onset of neutronic/ thermal-hydraulic instability. The 15 minute Frequency for this SR is not to ;

be used to delay entry into Condition B following an unexpected reduction in feedwater heating, recirculation pump trip, recirculation pump down shift to slow speed, or j significant flow control valve closure (small changes in i flow control valve position are not considered significant).

This Surveillance is modified by a Note which allows 15 minutes to verify FCDB following entry into the Restricted Region if the entry was the result of an unexpected transient (i.e., an unintentional or unplanned change in j core thermal power or core flow) . The 15 minute allowance is based on both engineering judgment and the availability of'the PBDS to provide the operator with information regarding the potential imminent onset of neutronic/ thermal-hydraulic instability. The 15 minute allowance of the Note is not to be used to delay entry into Condition B if the entry into the Restricted Region was the result of an unexpected reduction in feedwater heating, recirculation I

(continued) f GRAND GULF B 3.2-11f LDC 98037

i FCBB B 3.2.4

&Jett/ $ f9e

/) 3, .2 - il BASES [@/ 7 y 7

_ M l SURVEILLANCE SR 3.2.4.1 (continued)

REQUIREMENTS pump trip, recirculation pump down shift to slow speed, or significant flow control valve closure (small changes. in flow control valve position are not considered significant).

REFERENCES 1. NEDO-32339-A, " Reactor Stability Long Term Solution:

Enhanced Option I-A,".

1 I

l GRAND GULF B 3.2-11g j LDC 98037 j

RPS In.ctrir at*. tion i.1.1 l- BASES APPLICABLE 2.b. Averace Power Rance Monitor Fixed Neutron Flux-Hiah l

SAFETY ANALYSES, (continued)

LCO, and

! APPLICABILITY The Average Power Range. Monitor Fixed Neutron Flux-High Function is required to be OPERA 8LE in MODE 1 where the potential consequences of the analyzed transients could result in the SLs (e.g., MCPR and RCS pressure) be.'n1 exceeded. Although the Average Power Range Monitor Fixed Neutron Flux-High Function is assumed in the CRDA anaijcis that is applicable in MODE 2, the Average Power Range Monitor Neutron Flux-High, Setdown Function conservatively bounds the assumed trip and, together with the assumed IRM trips, provides adequate protection. Therefore, the Average Power Monitor Fixed Neutron Flux-High Function is not required in MODE 2.

2.c. Averace Power Rance Monitor-Inon This signal provides assurance that a minimum number of

. APRMs are OPERABLE. Anytime an APRM mode switch is moved to

! any position other than Operate, an APRM module is unplugged, the electronic operating voltage is low, or the APRM has too few LPRM inputs (< 14), an inoperative trip signal will be received by the RPS, unless the APRM is bypassed. Since only one APRM in each trip system may be bypassed, only one APRM in each trip system may be

! inoperable without resulting in an RPS trip signal. This Function was not specifically creditM in the accident analysis, but it is retained for fte RPS as required by the NRC approved licensing basis.

Six channels of Average Power Range Monitor-Inop with three channels in each trip system are required to be OPERABLE to ensure that no single failure will preclude a scram from l this Function on a valid signal.

There is no Allowable Value for this Function.

l This Function is required to be OPERABLE in the MODES where the APRM Functions are required.

SEW /Ve d \ _

Se.c W B 3.3.II 2 cl ,

r (continued)

GRAND GULF B 3.3-9 Revision No. 0 l

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

l RPS Instrumentetien B 3.3.1.1 sruser rs/n )

I

[3 3. 3 - 'T BASES [ opp / #f h J

APPLICABLE 2.d. Average Power Range Monitor Flow Biased Simulated SAFETY ANALYSES, Thermal Power - High LCO, and APPLICABILITY The Average Power Range Monitor Flow Biased Simulated (continued) Thernal Power - High Function monitors neutron flux to approximate the THERMAL POWER being transferred to the reactor coolant. The APRM neutron flux is electronically filtered with a time constant representative of the fuel heat transfer dynamics to generate a signal proportional to the THERMAL POWER in the reactor. The trip level is varied as a function of recirculation drive flow and is clamped at an upper limit that is always lower than the Average Power Range Monitor Fixed Neutron Flux - High Function Allowable Value. The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function provides a general definition of the licensed core power / core flow operating domain. .

The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is not associated with a limiting safety system setting. Operating limits established for the licensed operating domain are used to develop the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function Allowable Values to provide preemptive reactor scram and prevent gross violation of the licensed operating domain. Operation outside the license operating domain may result in anticipated operational occurrences and postulated accidents being initiated from conditions beyond those assumed in the safety

. analysis. Operation within the licensed operating domain also ensures compliance with General Design Criterion 12.

General Design Criterion 12 requires protection of fuel l

)

thermal safety limits from conditions caused by .

neutronic/ thermal-hydraulic instability. Neutronic/ thermal-hydraulic instabilities result in power oscillations which uld result in exceeding the MCPR SL.

The area of the core power and flow operating domain susceptible to neutronic/ thermal-hydraulic instability can !

be affected by reactor parameters such as reactor inlet feedwater temperature (Ref. 12). Two complete and independent sets of Average Power Range Monitor Flow-Biased j Simulated Thernal Power - High Function Allowable Values are specified in the COLR. Set 1 (Normal Trip Reference Set)

.provides protection against neutronic/ thermal-hydraulic instability.during expected reactor operating conditions.

Set 2 (Alternate Trip. Reference Set) provides protection GRAND GULF. B 3.3-9a LDC 90037 l

I-

'o l

RPS Instrumentation

_B 3.3.1.1_

h fSP)?h h 0&d v 2j.a e'

\

against neutronic/ thermal-hydraulic instability during reactor operating conditions requiring added stability protection and is conservative with respect to Set 1.

Feedwater temperature values requiring transition between flow control trip reference card sets are specified in the COLR. In the event of a feedwater temperature reduction, Allowable value ' modification (from the 2:ormal~ Trip Reference

-Set-to the Alternate Trip Reference Set) is required to

,p reserve the margin associated with the. potential for the, onset of neutronic/ thermal-hydraulic instability which existed prior to the feedwater temperature reduction. The Allowable Value modification required by the COLR may.be

delayed up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months to allow time to adjust and check the adjustment of each flow control trip reference card. At the end of the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months period, the Allowable Value modifications must be complete for all of the required channels or the applicable Condition (s) must be entered and the Required Actions taken. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time period is acceptable based on the low probability of a neutronic/ thermal-hydraulic instability event and the continued protection provided by n .the flow control trip reference card. In addition, when the feedwater temperature reduction results in operation in-either the Restricted Region or the Monitored Region, the requirements lfor the Period Based Detection System (LCO 3.3.1.3, Period Based Detection System (PBDS)) provide added protection against neutronic/ thermal-hydraulic instability during the 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months time period.

The area of the core power and flow operating domain susceptible to neutronic/ thermal-hydraulic instability is i affected by the value of Fraction of Core Boiling Boundary

-(LCO 3.2.4, FCBB)- '(Ref. 12). " Setup" and normal ("non-Setup") Average Power Range Monitor Flow Biased Simulated Thermal.. Power -~High Function' Allowable Values are specified in the COLR, (continued) 4 1

i L

GRAND GULF B 3.3-9a LDC 98037

- i

! l l.

' Sf f YD l e /3 3 . 3 - 1 l f oj e 3 7 4 ,

J l

The normal ("non-Setup") value provides protection against neutronic/ thermal- I hydraulic instability by preventing operation in the susceptible area of the operating domain when operating outside the Restricted Region specified in the COLR with the FCBB limit not required to be met. When the " Setup" value is selected, meeting the FCBB limit provides protection against instability.

1

" Setup" and "non-Setup" values are selected by operator manipulation of a '

Setup button on each flow control trip reference card. Selection of the ;

" Setup" value is intended only for planned operation in the Restricted Region '

as specified in the COLR. Operation in the Restricted Region with the Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function

" Setup" requires the FCBB limit to be met and is not generally consistent with l normal power operation. l The Average Power Range Monitor Flow Biased Simulated Thermal Power - High !

Function uses a trip level generated by the flow control trip reference card based on recirculation loop drive flow. Proper trip level generation as a function of drive flow requires drive flow alignment. This is accomplished by selection of appropriate dip switch positions on the flow control trip reference cards (Refer to SR 3.3.1.1.18). Changes in the core flow to drive flow functional relationship may vary over the core flow operating range. j These changes can result from both gradual changes in recirculation system and I core components over the reactor life time as well as specific maintenance performed on these components (e.g., jet pump cleaning).

The APRM System is divided into two groups of channels with four APRM inputs l to each trip syntem. The system is designed to allow one channel in each trip '

system to be bypassed. Any one APRM channel in a trip system can cause the associated trip system to trip. Six channels of Average Power Range Monitor Flow Biased Simulated Thermal Power - High with three channels in each trip system arranged in a l

f

c.___ . _ _ . _ _ . _ . _ _ _______.___.___________________ _ _ __ _ ____ _ ____ _ _ _ ___ _ _ ._ _ _ _ ___ _ _

- t RPS In trumentation

, % B 3.3.1.1 y

$%W 7D f O SC hE) e Y YY one-out-of-three logic are required ~to,be OPERABLE to ensure i that'.no single instrument failure will preclude a scram from this Function on a valid signal. In addition, to i- provide adequate coverage of the' entire core, at least 14 LPRM inputs are required for each APRM channel, with at {

1 east two LPRM inputs from each of the four axial' levels at wh'ich the.LPRMs are located. Each APRM channel receives one

, total drive flow signal representative of total core flow.

l The recirculation loop drive flow signals are generated by eight flow units. One flow unit from each recirculation loop is provided to each APRM channel. Total drive flow is determined by each APRM by summing up the flow signals provided.to the APRM from the two recirculation loops.

The THERMAL POWER time constant is based on the fuel heat transfer dynamics and provides a signal proportional to.the THERMAL POWER, The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function is required to be OPERABLE in MODE 1 when there is the possibility of neutronic/ thermal-hydraulic' instability. .The potential to exceed the SL applicable to high pressure and core flow conditions (MCPR I SL), which provides fuel. cladding integrity protection, exists if neutronic/ thermal-hydraulic instability can occur.

During MODES 2 and 5, other IRM and APRM Functions provide protection for fuel cladding integrity.

(continued)- l l

L" l

1 l

l I.

I I

1.

' GRAND GULF B 3.3-9c LDC 98037

RPS Instrumentation B 3.3.1.1 BASES l

SURVEILLANCE SR 3.3.1.1.9 REQUIREMENTS (continued) The calibration of trip units provides a check of the actual trip setpoints. The channel must be declared inoperable if the trip setting is discovered to be less conservative than the Allowable Value specified in Table 3.3.1.1-1. If the l

trip setting is discovered to be less conservative than accounted for in the appropriate setpoint methodology, but is not beyond the Allowable Value, the channel performance is still within the requirements of the plant safety

! analysis. Under these conditions, the setpoint must be l

readjusted to be equal to or more conservative than .

accounted for in the appropriate setpoint methodology. '

The Frequency of 92 days for SR 3.3.1.1.9 is based on the reliability analysis of Reference 9.

-- ~ - -

SR 3.3.1.1.10. R 3.3.1.1.12 d d SF 3' 3* l' )

- - - _a-

/- -

/

A CHANNEL CALIBRATION is a complete check of the instrument loop and the sensor. This test verifies the channel responds to the measured parameter within the necessary range and accuracy. CHANNEL CALIBRATION leaves the channel

- adjusted to account for instrument drifts between successive calibrations consistent with the plant specific setpoint j

j methodology.

L , .

Note 1 states ~ that neutron detectors are excluded from

! CHANNEL CALIBRATION because of the difficulty of simulating a meaningful signal. Changes in neutron detector sensitivity are compensated for by performing the 7 day calorimetric calibration (SR 3.3.1.1.2) and the 1000 MWD /T LPRN calibration against the TIPS (SR 3.3.1.1.7). A second Note is provided that requires the APRM and IRM SRs to be

performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months of entering MODE 2 from MODE 1.

l Testing of the MODE 2 APRM and IRM Functions cannot be l performed in MODE 1 without utilizing jumpers, lifted leads l

or movable links. This Note allows entry into MODE 2 from MODE 1 if the associated Frequency is not met per SR 3.0.2.

f Twelve hours is based on operating experience and in l

consideration of providing a reasonable time in which to. '

complete the SR.]Th: f5;;: ;y e7 = 3.3.1R6R F 3A1.1.!! is b: sed 0;;;; the-assumption-of-the-magnitude-of

)3 !

i,.;;t 6 ift in the setpeint-analysis r .--

(continued)

GRAND GULF B 3.3-27 Revision No. 0

~

v

-e h${W h f094 Qfoje 3. 3 -2Y

/ '6 j ,

t'

Note. 3 to SR 3.3.1.1.10 states that the APRM recirculation flow transmitters are excluded from CHANNEL CALIBRATION of Function 2.d, Average Power Range Monitor Flow Biased

,. Simulated Thermal Power-High. Calibration of the'flov

}. transmitters is performed 'on an 18-month ' frequency (SR 3.3.1.1.17). Note 4 to SR 3.3.1.1.10 states that the digital components of the flow control-trip reference card are excluded from CHANNEL CALIBRATION cf Function 2.d, Average Power Range Monitor Flow Biased Simulated Thermal Power-High. The analog output potentiometers of the flow l r'ontrol trip reference card are not excluded. The flow control trip reference card has an automatic self-test feature which periodically tests the hardware which performs the digital algorithm. Exclusion of the digital components of the flow control trip reference card from CHANNEL

' CALIBRATION of Function 2.d is based on the conditions required to perform the test and the likelihood of a change in the status of these components not being detected.

I The Frequency of SR 3.3.1.1.10, SR 3.3.1.1.12 and SR 3.3.1.1.17 is based upon the assumption of the magnitude of equipment drift in the setpoint analysis.

l l

p

.~. -~. .

l l

. RPS Instrumentation B 3.3.1.1 BASES l SURVEILLANCE SR -3.3.1.1.15 (continued) l REQUIREMENTS RPS RESPONSE TIME tests are conducted on an 18 month STAGGERED TEST BASIS. Note 3 requires STAGGERED TEST BASIS I l' Frequency to be determined based on 4 channels per trip system, in lieu of the 8 channels specified in. Table 3.3.1.1-1 for the MSIV Closure Function. This Frequency-ts

! based on the logic interrelationships of the various channels required to produce an RPS scram signal.

Therefore, staggered testing results in response time verification of these devices every 18 months. This requency is consistent with the typical industry refueling lecycle and is based upon plant operating experience, which j 5p/[I g3 6 p3,d g shows that random failures of instrumentation components i8 A causing serious time degradation, but not channel failure, g,, are infrequent.

REFERENCES 1. UFSAR, Figure 7.2-1.

2. UFSAR, Section 5.2.2.

3. UFSAR, Section 6.3.3.

4. UFSAR, Chapter 15.

5. UFSAR, Section 15.4.1.

6. NED0-23842, " Continuous Control Rod Withdrawal in the Startup Range," April 18, 1978.

7. UFSAR, Section 15.4.9.

l l

1 (continued) l l

1 GRAND GULF B 3.3-29a LDC 97078 )

a RPS In;trumentation D 3.3.1.1

! BASES

[d8fW D DE f3 ,7, 3 -j2 9 gy N /* / 8ij SR 3.3.1.1.16 The Average Power Range Monitor Flow Biased Simulated Thermal Power - High Function uses an electronic filter circuit to generate a signal proportional to the core THERMAL POWER from the APRM neutron flux signal. This filter circuit is representative of the fuel heat transfer

' dynamics that produce the relationship between the neutron flux and the core THERMAL POWER. The filter time constant must be verified to ensure that the channel is accurately reflecting the desired parameter.

The Frequency of 18 months is based on engineering judgment and reliability of the components.

SR 3.3.1.1.18 The Average Poyer Range Monitor Flow Biased Simulated Thermal Power - High Function uses a trip level generated by the flow control trip reference card based on the recirculation loop drive flow. The drive flow is adjusted by a digital algorithm according to selected drive flow alignment dip switch settings. This SR sets the flow j control trip reference card to ensure the drive flow I alignment used results in the appropriate trip level being f generated from the digital components of the card.

l The Frequency of once following a refueling outage is based on the expectation that any change in the core flow to drive !

flow functional relationship during power operation would be gradual and that maintenance on recirculation system and j core components which may impact the relationship is ;

expected to be performed during refueling outages. The completion time of 7 days after reaching equilibrium conditions is based on plant conditions required to perform the test and engineering judgment of the time required to collect and analyze the necessary flow data and the time required to adjust and check the adjustment of each flow control trip reference card. The completion time of 7 days after reaching equilibrium conditions is acceptable based on the low probability of a neutronic/ thermal-hydraulic instability event.

RPS Instrumentation B 3.3.1.1 l

BASES

. REFERENCES 8 .' Letter, P. Check (NRC) to G. Lainas (NRC), "BWR i

(continued) Scram Discharge System Safety Evaluation,"

December 1, 1980, as attached to NRC Generic Letter dated December 9, 1980.

9. NED0-30851-P-A, " Technical Specification Improvement Analyses for BWR Reactor Protection System,"

March 1988.

10. NED0-32291-A, " System Analyses for Elimination of Selected Response Time Testing Requirements," October l 1995.

11. GhRI-97/00181, Amendment 133 to the Operating License.

I

12. NEDO-32339-A, "Long Term Stability Solution: Enhanced Option I-A."

~ __

I l

l GRAND GULF B 3.3-30 LDC 97078

l SRM Instrumentation B 3.3.1.2 i i i

BASES SURVEILLANCE SR 3.3.1.2.6 (continued) l REQUIREMENTS otherwise verified to be OPERABLE (i.e., satisfactorily performing the CHANNEL CHECK) and the time required to perform the. Surveillance.

' REFERENCES None.

b SErr NPLO \

I Se c-Mon 3 3 3 .l. 3 k l l

[

GRAND GULF B 3.3-39 Revision No. 0 E--_---___-.--_

l l

PBDS B 3.3.1.3 I W Mv /- /r> [$E B 3.3 INSTRUMENTATION p

B 3.3.1.3 Period Based Detection System (PBDS) / #3 C

Ib i l'

, BASES I

BACKGROUND General Design Criterion 12 requires protection of fuel I

thermal. safety limits from conditions caused by neutronic/ thermal-hydraulic instability. Neutronic/ thermal-hydraulic instabilities can result in power oscillations which could result in exceeding the MCPR Safety Limit (SL).

The MCPR SL ensures that at least 99.9% of the fuel rods avoid boiling transition during normal operation and during an anticipated operational occurrence (AOO) (refer to the Bases for SL 2.1.1.2).

The PBDS provides the operator with an indication that conditions consistent with a significant degradation in the stability performance of the reactor core have occurred and L

the potential for imminent onset of neutronic/ thermal-hydraulic instability may exist. Indication of such degradation is cause for the operator to initiate an immediate reactor scram if the reactor is being operated in either the Restricted Region or Monitored Region. The j Restricted Region and Monitored Region are defined in the 1 COLR. I i

The PBDS instrumentation of the Neutron Monitoring System consists of two channels. Each of the PBDS channels includes input from local power range monitors (LPRMs) within the reactor core. These inputs are continually monitored by the PBDS for variations in the neutron flux consistent with the onset of neutronic/ thermal-hydraulic instability. Each channel includes separate local indication and control room Hi-Hi DR Alarm. While this LCO specifies OPERABILITY requirements only for one monitoring and indication channel of the PBDS, if both are OPERABLE, a Hi-Hi DR Alarm from either channel results in the need for the operator to take actions.

The primary PBDS component is a card in the Neutron I Monitoring System with analog inputs and digital processing.

The PBDS card has an automatic self-test feature to

! periodically test the hardware. circuit.

The self-test I functions are executed during their allocated portion of the executive loop sequence. Any self-test failure indicating loss of critical function results in a control room alarm.

(continued)

I GRAND GULF B 3.3-39a LDC 9803*/

i - >

gg PBDS g @t-i O M 3.3.1.3 J

0.Sd BASES

$ WB 72> 3 3-[3 T ;

A[4je A ef i )

BACKGROUND The inoperable condition is also displayed by an indicating (continueoJ. light on the card front panel. A manually initiated internal test sequence can be actuated via a recessed push button. This internal test consists of simulating alarm and inoperable conditions to verify card OPERABILITY.

Descriptions of the PBDS are provided in References 1 and 2. j i

Actuation of the PBDS Hi-Hi DR Alarm is not postulated to !

occur due to neutronic/ thermal-hydraulic instability outside {

the Restricted Region and the Monitored Region. Periodic '

perturbations can be introduced into the thermal-hydraulic l behavior of the reactor core from external sources such as I recirculation system components and the pressure and ;

feedwater control systems. These perturbations can ;

potentially drive the neutron flux to oscillate within a frequency range expected for neutronic/ thermal-hydraulic instability. The presence of such oscillations would be recognized by the period based algorithm of the PBDS and potentially result in a Hi-Hi DR Alarm. Actuation of the PBDS Hi-Hi DR Alarm outside the Restricted Region and the Monitored Region would indicate the presence of a source external to the reactor core and are not indications of neutronic/ thermal-hydraulic instability.

APPLICABLE ' Analysis, as described in Section 4 of Reference 1, confirms SAFETY ANALYSES that AOOs initiated from outside the Restricted Region without stability control and from within the Restricted Region with stability control are not expected to result in neutronic/ thermal-hydraulic instability. The stability control applied in the Restricted Region (refer to LCO 3.2.4, " Fraction of Core Boiling Boundary (FCBB) ") is established to prevent neutronic/ thermal-hydraulic instability during operation in the Restricted Region.

Operation in the_ Monitored Region is only susceptible to instability under hypothetical operating conditions beyond those analyzed in Reference 1. The types of transients specifically evaluated are loss of flow and coolant temperature decrease which are limiting for the onset of instability.

l

' The initial conditions assumed in the analysis are reasonably conservative and the immediate post-event reactor conditions are significantly stable. However, these assumed i

initial conditions do not bound each individual parameter which impacts stability performance (Ref. 1). The PBDS )

1 GRAND GULF B 3.3-39b LDC 98037 s

PBDS g OM - - -

B 3.3.1.3

' zuser re P m

'^ p 3. 3 - M p ee 3 y9 N -

APPLICABLE instrumentation provides the operator with an indication SAFETY ANALYSES that conditions consistent with a significant degradation in 1 (continued) the stability performance of the reactor core has occurred I and the potential for imminent onset of neutronic/ thermal- !

hydraulie instability may exist. Such conditions are only postulated to result from events initiated from initial conditions beyond the conditions assumed in the safety analysis (refer to Section 4, Ref. 1)

The PBDS has no safety function and is not assumed to function during any FSAR design basis accident or transient j analysis. However, the PBDS provides the only indication of I the imminent onset of neutronic/ thermal-hydraulic instability during operation in regions of the operating domain potentially susceptible to instability. Therefore, the PBDS is included in the Technical Specifications.

LCO One PBDS channel is required to be OPERABLE to monitor reactor ncutron flux for indications of imminent onset of neutronic/ thermal-hydraulic instability. OPERABILITY requires the ability for the operator to be immediately alerted to a Hi-Hi DR Alarm. This is accomplished by the instrument channel control room alarm. The LCO also requires reactor operation be such that the Hi-Hi DR Alarm is not actuated by an OPERABLE PBDS instrumentation channel.

APPLICABILITY At least one of two PBDS instrumentation channels is

, required to be OPERABLE during operation in either the Restricted Region or the Monitored Region specified in the COLR. Similarly, operation with the PBDS Hi-Hi DR Alarm of any CDERABLE PSDS instrumentation channel is not allowed in the Restricted Region or the Monitored Region. Operation in these regions is susceptible to instability (refer to the Bases for LCO 3.2.4 and Section 4 of Ref. 1). OPERABILITY of at least one PBDS instrumentation channel and operation with no indication of a PBDS Hi-Hi DR Alarm from any OPERABLE PBDS instrumentation channel is therefore required during operation in these regions.

l L The boundary of the Restricted Region in the Applicability of this LCO is analytically established in terms of thermal power and core flow. The Restricted Region is defined by the APRM Flow Biased Neutron Plux - Upscale Control GRAND GULF B 3.3-39c LDC 98037 l

PODS g g [OM B 3.3.1.3 MMF 7D [Te*

8^888 B 3. 3 - 31 A e # vr 'l

~

l APPLICABILITY Rod Block setpoints, which are a function of reactor (continued) recirculation drive flow. The Restricted Region Entry Alarm (EREA) signal is generated by the Flow Control Trip Reference (FCTR) card using the APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints. As a result,

'the RREA is coincident with the Restricted Region boundary when the setpoints are not " Setup", and provides indication of entry into the Restricted Region. However, APRM Flow Biased Neutron Flux - Upscale Control Rod Block signals provided by the FCTR card, that are not coincident with the Restricted Region boundary, do not generate a valid RREA.

The Restricted Region boundary for this LCO Applicability is specified in the COLR.

When the APRM Flow Biased Neutron Flux - Upscale Control Rod i Block setpoints are " Setup" the applicable setpoints used to generate the RREA are moved to the interior boundary of the Restricted Region to allow controlled operation within the Restricted Region. While the setpoints are "Setupa the Restricted Region boundary remains defined by the .norrel APRM Flow Biased Neutron Flux - Upscale Control Rod Block ;

setpoints. )

Parameters such as reactor power and core flow available at l

the reactor controls, may be used to provide immediate confirmation that entry into the Restricted Region could reasonably have occurred.

l The Monitored Region in the Applicability of this LCO is analytically established in terms of thermal power and core flow. However, unlike the Restricted Region boundary the Monitored Region boundary is not specifically monitored by plant instrumentation to provide automatic indication of region entry.- Therefore, the Monitored Region boundary is i

defined in terms of thermal power and core flow. The

{

Monitored Region boundary for this LCO Applicability is l specified in the COLR. '

l l Operation outside the Restricted Region *and the Monitored

)'

1 Region is not susceptible to neutronic/ thermal-hydraulic instability even under extreme postulated conditions.

GRAND GULF B 3.3-39d '

LDC 98037 l

d) B 3.3 3

! BASES (continued) Zg A) W ~f D f.SY 3, J g fW I Y $

ACTIONS A.1 If at any time while in the Restricted Region or Monitored Region, an OPERABLE PBDS instrumentation channel indicates a valid Hi-Hi DR Alarm, the operator is required to initiate an immediate reactor scram. Verification that the Hi-Hi DR Alarm is valid may be performed without delay against another output from a PBDS card observable from the reactor ;

i controls in the control room prior to the manual reactor scram. This provides assurance that core conditions leading {

te neutronic/ thermal-hydraulic instability will be '

mitigated. This Required Action and associated Completion Time does not allow for evaluation of circumstances leading to-the Hi-Hi DR Alarm prior to manual initiation of reactor scram.

B.1 and B.2 ,

Operation with the APRM Flow Biased Simulated Thermal Power

- High Function (refer to LCO 3.3.1.1, Table 3.3.1.1-1, Function 2.d) " Setup" requires the stability control applied in the Restricted Region (refer to LCO 3.2.4) to be met.

Requirements for operation with the stability control met are established to prevent reactor thermal-hydraulic instability during operation in the Restricted Region. With the required PBDS channel' inoperable, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal-hydraulic instability as a result of unexpected transients is lost. Therefore, action must be immediately initiated to exit the Restricted Region. While the APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints are " Setup," operation in the Restricted Region may be confirmed by use of plant parameters such as I reactor power and core flow available at the reactor controls.

Exit of the Restricted Region can be accomplished by control i I

rod insertion and/or recirculation flow increases. Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in unstable reactor conditions and are not allowed to be used to comply with this Required Action.

The time required to exit the Restricted Region will depend ;

on existing plant conditions. Provided efforts are begun I without delay and continued until the Restricted Region is l (continued)

GRAND GULF B 3.3-39e LDC 98037 i

PBDS B 3.3.1.3 4

h. I h.5W h f%

BASES

[ g , O13" V Q f*)"&11

~

ACTIONS B.1 and B.2 (continued)

]

exited,-operation is acceptable based on the low probability of a transient which degrades stability performance occurring simultaneously with the required PBDS channel in' operable.

Required Action B.1 is modified by a Note that specifies that initiation of action to exit the Restricted Region only applies if the APRM Flow Biased Simulated Thermal Power -

High Function is " Setup". Operation in the Restricted j Region without the APRM Flow Biased Simulated Thermal Power

- High Function " Setup" indicates uncontrolled entry into the Restricted Region. Uncontrolled entry is consistent

-with the occurrence of unexpected transients, which, in combination with the absence of stability controls being met may result in significant degradation of stability performance. .

When the APRM Flow Biased Neutron Flux - Upscale Control Rod Block setpoints are not " Setup" uncontrolled entry into the Restricted Region is identified by receipt of a valid RREA.

Immediate confirmation that RREA is valid and indicates an actual entry into the Restricted Region may be performed without delay. Immediate confirmation constitutes observation that plant parameters immediately available at the reactor controls (e.g., reactor power and core flow) are reasonably consistent with entry into the Restricted Region.

l

' This immediate confirmation may also constitute recognition that plant parameters are rapidly changing during a transient (e.g., a recirculation pump trip) which could l reasonably result in entry into the Restricted Region.

For uncontrolled entry into the Restricted Region with the required PBDS instrumentation channel inoperable, the l ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal-hydraulic instability is lost and continued operation is not justified. Therefore, Required Action B.2 requires immediate reactor scram.

c.1 In the Monitored Region the PBDS Hi-Hi DR Alarm provides indication of degraded stability performance. Operation in the Monitored Region is susceptible to neutronic/ thermal (continued)

GRAND GULF B 3.3-39f LDC 98037

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d 8( j B 3.3_

BASES r o sen r ro pok 8 ,*). 3 - 7 f f') ! 7f 9 ACTIONS C.1 (continued) hydraulic instability under postulated conditions exceeding those previously assumed in the safety analysis. With the required PBDS channel inoperable, the ability to monitor conditions indicating the potential for imminent onset of neutronic/ thermal-hydraulic instability is lost. Therefore, action must be initiated to exit the Monitored Region. ;

Actions to restart an idle recirculation loop, withdraw control rods or reduce recirculation flow may result in approaching unstable reactor conditions and are not allowed to be used to comply with this Required Action. Exit of the Monitored Region is accomplished by control rod insertion and/or recirculation flow increases. However, actions which reduce recirculation flow are allowed provided the Fraction of Core Boiling Boundary (FCBB) is recently (within 15 minutes) verified to be 5 1.0. Recent verification of FCBB being met, provides assurance.that with the PBDS inoperable, planned decreases in recirculation drive flow should not result in significant degradation of core stability performance.

The specified Completion Time of 15 minutes ensures timely operator action to exit the region consistent with the low probability that reactor conditions exceed the initial conditions assumed in the safety analysis. The time l required to exit the Monitored Region will depend on I existing plant conditions. Provided efforts are begun within 15 minutes and continued until the Monitored Region is exited, operation is acceptable based on the low probability of a transient which degraJes stability performance occurring simultaneously with the required PBDS channel inoperable.

SURVEILLANCE SR 3.3.1.3.1 REQUIREMENTS During operation in the Restricted Region or the Monitored Region the PBDS Hi-Hi DR Alarm is relied upon to indicate ,

conditions consistent with the imminent onset of !

neutronic/ thermal-hydraulic instability. Verification every I 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months provides assurance of the proper indication of the alarm during operation in the Restricted Region or the Monitored Region. The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency supplements less l formali but more frequent, checks of alarm status during operation.

t 1

(continued)

GRAND GULF B 3.3-39g LDC 98037

l 1e PBDS d 8D-lO N B 3.3.1.3 BASES IN-.ccsr g 3_ g9 x /4e h f*H E $ 1 SURVEILLANCE BR 3.3.1.3.2 REQUIREMENTS (continued) Performance of the CHANNEL CHECK every 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months ensures that a gross failure of instrumentation has not occurred. This CHANNEL CHECK is normally a comparison of the PBDS indication to the state of the annunciator, as well as comparison to the same parameter on the other channel if it is available. It is based on the assumption that the instrument channel indication agrees with the immediate indication available to the operator, and that instrument channels monitoring the same parameter should read

, similarly. Deviations between the instrument channels ceiuld be an indication of instrument component failure. A CHANNEL CHECK will detect gross channel failure; thus, it is key to verifying the instrumentation continues to operate properly between each CHANNEL FUNCTIONAL TEST. Agreement criteria are determined by the plant staff based on a combination of the channel instrument uncertainties, including indication and readability.

f The 12 hour1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months Frequency is based on operatLng experience that I

' demor.strates channel failure is rare. The CHANNEL CHECK supplements less formal, but more frequent, checks of channels during normal operational use of the displays l associated with the channels required by the LCO.

SR 3.3.1.3.3 l

A CHA!;NEL FUNCTIONAL TEST is performed for the PBDS to ensure that the entire system will perform the intended function. The CHANNEL FUNCTIONAL TEST for the PBDS includes manual initiation of an internal test sequence and j.

I

' verification of appropriate alarm and inop conditions being reported.

Performance of a CHANNEL FUNCTIONAL TEST at a Frequency of 24 months verifies the performance of the PBDS and associated circuitry. The Frequency considers the plant !

conditions required to perform the test, the ease of performing the test, and the likelihood of a change in the system or component status. The alarm circuit is designed to operate for over 24 months with sufficient accuracy on signal amplitude and signal timing considering environment, initial calibration and accuracy drift (Ref. 2).

(continued)

GRAND GULF B 3.3-39h LDC 98037 j

PBDS B 3.3.2.3 BASES (continued) 8 7M5f F 7b B 3. 3 -M [o'M >

-" Q fose 4 g 9 REFERENCES 1. NEDO-32339-A, " Reactor Stability Long Term Soluticn:

Enhanced Option I-A.

2. NEDC-32339P-A, Supplement 2, " Reactor Stability Long ,

Term Solution: Enhanced Option I-A Solution Design."

l l

L G GEF B 3.3-391 LDC 98037

t i

i

~ Recirculation Loops Operating B 3.4.1 ,

and APRM Flow Biased Simulated Thermal Power-High, j ALLOWABLE Value (LCO 3.2.4, " Fraction of Core Boiling !

BASES Boundary" (FCBB), LCO 3.3.1.1, "RPS Instrumentation", '

t and LCO 3.3.1.3. " Per i nd Raned natmetinn Evstem" (PRnA) p APPLICABLE margins during abnormal operational transients (Ref. 2),

SAFETY ANALYSES which are analyzed in Chapter 15 of the UFSAR.

(continued) !

i." A plant specific LOCA analysis has been performed assuming l only one operating recirculation loop. This analysis has demonstrated that, in the event of a LOCA caused by a pipe break in the operating recirculation loop, the Emergency l'

, Core Cooling System response will provide adequate core cooling, provided the APLHGR requirements are modified accordingly (Ref. 3).

The transient analyses of Chapter 15 of the UFSAR have also been perforined for single recirculation loop operation (Ref. 3) and demonstrate sufficient flow coastdown characteristics to maintain fuel thermal margins during the abnormal operational transients analyzed provided the MCPR requirements are modified. The APLHGR and MCPR limits for single loop operation are specified in the COLR.

Recirculation loops operating satisfies Criterion 2 of the NRC Policy Statement.

LCO Two recirculation loops are normally required to be in i operation with their flows matched within the limits '

specified in SR 3.4.1.1 to ensure that during a LOCA caused by a break of the piping of one recirculation loon the ass ions of the LOCA analysis are satisfiedd4n

._ , th: ate! ca-- ;;= =pr= == n : ferrtia= af ] i i T"5.""",L ir. Fi;:n 3.'

n-.4..,

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!-1. "=:t5""f.L

'" i= ":;i= " er ":;ier,

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F 5 Stibi'it"

" finaratinn 4m' Dan 4am e 4, anlu .11 -- - J L-4mn ~

f 6545-5 rd S'+WiC '5Ib5Etit555pd Iternat1N1y, i with only one recirculation loop in operanon, modifications '

to the required APLHGR limits (LCO 3.2.1, " AVERAGE PLANAR !

LINEAR HEAT GENERATION RATE (APLHGR)"), .aMMCPR limits J l (LCO 3.2.2, " MINIMUM CRITICAL POWER RATIO (MCPR)y must be l applied to allow con _tinued oneration consistent with the assumptionsof(Refe'ren&i{ !

The LCO is modified by a Note which allows up to 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months :

before having to put in effect the required modifications to required limits after a change in the reactor operating !

conditions from two recirculation loops operating to single !

recirculation loop operation. If the required limits are (continued) l GRAND GULF B 3.4-3 Revision No. 0

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ a'

Recirculation Lorps Operating B 3.4.1 BASES i

I LCO not in compliance with the applicable requirements at the (continued) end of this period, the associated equipment must be declared inoperable or the limits "not satisfied," and the ACTIONS required by nonconformance with the applicable ;

specifications implemented. This time is provided due to I the need to stabilize operation with one recirculation loop, including the procedural steps necessary to limit flow and flow control mode in the operating loop, limit total THERMAL POWER, Wit:r f:r exc:::ive A""",

er.d loca] ::= r r:R F C l C-r4t6r (LPPM) neetren flux neise4eveksf and the complexity l and detail required to fully implement and confirm the required limit modifications.

APPLICABILITY In MODES 1 and 2, requirements for operation of the Reactor Coolant Recirculation System are necessary since there is !

considerable energy in the reactor core and the limiting i design basis transients and accidents are assumed to occur.

In MODES 3,'4, and 5, the consequences of an accident are reduced and the coastdown characteristics of the recirculation loops are not important.

ACTIONS A.]

With both recirculation loops operating but the flows not I matched, the recirculation loops must be restored to '

operation with matched flows within 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months . Should a LOCA occur with one recirculation loop not in operation, the core flow coastdown and resultant core response may not be bounded by the LOCA analyses. Therefore, only a limited :

time is allowed to restore the inoperable loop to operating !

status.

Alternatively, if the single loop requirements of the LCO are applied to operating limits and RPS setpoints, operation with only one recirculation loop would satisfy the requirements of the LCO and the initial conditions of the accident sequence.

The 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months Completion Time is based on the low probability of an accident occurring during this time period, on a reasonable time to complete the Required Action, and on l

(continued) i GRAND GULF B 3.4-4 Revision No. O i

1 Recirculation Loops Operating l B 3.4.1 :

BASES ACTIONS A.,d (continued) frequent core monitoring by operators allowing abrupt changes in core flow conditions to be quickly detected.

This Required Action does not require tripping the l recirculation pump in the lowest flow loop when the mismatch !

between total jet pump flows of the two loops is greater j than the required limits. However, in cases where large !

flow mismatches occur, low flow or reverse flow can occur in '

the low flow loop jet pumps, causing vibration of the jet pumps. If zero or reverse flow is detected, the condition should be alleviated by changing flow control valve position 4 to re-establish forward flow or by tripping the pump. !

U . C . : . e..d D. ic-j Due t thermal hydraulic stability concerns, operation ~

the pla is divided into four regions based on THE POWER and re flows. Region A is a power / flow ra o with power > 10 flow. Region odisline and core flow < 40% of the ated core a power / flow ratio with the ower > 80%

I and s 100% rod es and flow < 40% of the ted flow.

l Region C is a powe low ratio with the p r > 80% rod line and core flow > 40% a < 45% of the r d core flow, respectively. In Regio B, increas potential for instability exists, and o ration restricted. Action is required to exit the region it 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by reducing

-THERMAL POWER through control d insertion or by increasing ;

recirculation loop flow (i ne r more pumps are on fast speed) by opening the fl .contro valve. Operation in Region C is also more sceptible instability than ncrr.41 operating parameter . .However, oper ion in this region is allowed during s tups if required fo fuel conditioning.

Under other ci umstances, action is req red to exit the region withi 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months by reducing THERMAL ER through control r insertion or by increasing recir lation loop '

flow if ne or more pumps are on fast speed) flow c(orol valve. If evidence of instability opening curs the (i.e. APRM oscillations > 10% or periodic LPRM up ale or down ale, in addition to the guidance provided in Re rence 5) during operation in either Region 8 or C, n

.immediate reactor scram is required (Ref. 4). The all (continued)

GRAND GULF B 3.4-5 Revision No. 2 l _ _ _ . . _ . ._ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ _ _ _______ J

Recirculation Le ps Operating B 3.4.1 l BASES ACTIONS

" l. 0.1. rd 0.1 _(00 tir.- 4)

Comp on Times are reasonable, based on opera experiene to restore plant parameters to al in an orderly' mann nd without challengin nt systems.

L.1 With no recircui n loops in ration in MODE 1, or while otherwise o ing in Region A, t nit is very suscept to instability, and an i te scram is re .

This action limits the time dur Awhich thermal J.,.. e.li; i z+951ity is er iivy .s1, propani r ty.

Y \

With no recirculation loops in operation 4; ..^;E 4, the unit isrequiredjobebroughttoaMODEinwhichtheLCDdoes not apalv.atter.

et M ii.i;;e;G iM;eteb te idne78--

("k.;-J.. C- te t 41thi.. th: 1i;;t; t: :n;.re th:m ' _ - J L h,0. ..d is it.hility Orrm; :n -ffx=;d. J'h: Ort 4r mr = t r;d t.: r.: r'----

- --: ? ? n 7.nc;. A In this condition, the recirculation ' loops are not required to be operating because of the reduced severity of DBAs and minimal dependence on the recirculation loop coastdown.

characteristics. The allowed Completion Time of 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months is reasonable, based on operating experience, to reach MODE 3 from full power conditions in an orderly manner and without challenging plant systems. .

., gg

-% at 1+1utb b e- y y, Ch brw 2

ic N406 3 arS. -

If the required limit modifications for single recirculation loop operation are not performed within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after transition from two recirculation loop operation to single recirculation loop operation, the required limits which have not been modified must be immediately declared not met. The Required Actions for the associated limits must then be taken.

(continued)

GRAND GULF B 3.4-6 Revision No. O

,. Recirculation Loops Operating

, B 3.4.1 BASES (continued)

SURVEILLANCE SR 3.4.1.1 REQUIREMENTS This SR ensures the recirculation loop flows are within the' allowable limits for mismatch. At low core flow (i.e.,

< 70% of rated core flow), the MCPR requirements provide larger margins to the fuel cladding integrity Safety Limit such that the potential adverse effect of early boiling transition during a LOCA is reduced. A larger flow mismatch can therefore be allowed when core flow is < 70% of rated core flow. The recirculation loop jet pump flow, as used in this Surveillance,. is the summation of the flows from all of

'the jet pumps assuciated with a single recirculation loop.

The mismatch is measured in terms of percent of rated core flow. This Surveillance can be met by verifying that the [

recirculation loop drive flow mismatch, when two loops are in operation, is < 55 of rated recirculation drive flow with core flow a: 70% of rated core flow and < 10% of rated recirculation drive flow with core flow < 70% of rated core flow. -

This SR is not required when both loops are not in operation i since the mismatch limits are meaningless during single loop or natural circulation operation. The Surveillance must be performed within 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after both loops are in operation.

The 24 hour2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months Frequency is consistent with the Frequency for jet pump OPERABILITY verification and has been shown by operating experience to be adequate to detect off normal jet pump loop flows in a timely manner.

-% ' . 4.1.4 C This SR ens the reactor !!TRMAL POWER and core f1 re within appropria rameter limits to prevent u rolled power oscillations. limits are Region Figure 3.4.1-1, or if withdrawing trol ro r startup, either Region D or Region C. At low r ulation flows and high reactor power, the reactor its ased susceptibility to thermal hydraulic a 111ty. Inter _ ctions have beer.

developed based a guidance provided in 5.to resp rences 4 and o operation in these conditions. SR ide es when the conditions requiring these inter ions are necessary. The Frequency is based on operatin (continued)

GRAND GULF B 3.4-7 Revision No. 1

Recirculation Loops Operating g

.'* B 3.4.1 BASES SURVEIL ANCE (conti und)

REQUIREMENTS experience and the operat n drenkknowledge of reactor stalusrdncTUHTng significant changes in7HE POWER and re flow.

REFERENCES 1. UFSAR, Section 6.3.3.7.

2. UFSAR, Section 5.4.1.1.

s ,

3. UFSAR, Chapter 15, Appendix 15C.

4.t =ac na11stin as-07. cif--1 7: ;; ;, EeWei 0;cill:tions in til17., ":t:r 5::t:rt," n--=;.= inna e,

Q. 0,E L:tt:r, "!rteri: S::

i--ti;;; . 5 - t r IaS

7.Atiens iur 5taoisity] '

1 -- _

DO-32339-A, " Reactor Stability Long Term olution: Enhanced Option I-A."

~

)

n l

l GRAND GULF B 3.4-8 Revision No. 1

.

ML20236T166

Text

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