ML20206J534

Marked-up Proposed Tech Specs Resolving NRC Concerns Re Core Stability & Supporting Single Loop Operations

ML20206J534
Person / Time
Site:	Grand Gulf
Issue date:	06/20/1986
From:	MISSISSIPPI POWER & LIGHT CO.
To:
Shared Package
ML20206J522	List: AECM-86-0159, Application for Amend to License NPF-29,changing Tech Specs to Resolve NRC Concerns Re Core Stability & to Support Single Loop Operations ML20206J534
References
TAC-61083, NUDOCS 8606270189
Download: ML20206J534 (64)

Text

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r 3/4.4 REACTOR C0OLANT SYSTEM @

3/4.4.1 RECIRCULATION SYSTEM RECIRCULATION LOOPS TusGRT

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LIMITING CONDITION FOR OPERATION f if s

.4.1.1 Two reactor coolant system recirculation loops shall be in operati .

APP CABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With o e reactor coolant system recirculation loop not in eration, immedia ly initiate an orderly reduction of THERMAL POW to less than or equal 80% of the 100% Rod Line as specified in F' ure B 3/4 2.3-1, and be in a least HOT SHUTDOWN within the next 12 h rs.

b. With no reactor oolant system recirculation 100- in operation, immediately initi e an orderly reduction of T MAL POWER to less than or equal to 80 of the 100% Rod Line as pecified in Figure B 3/4 2.3-1, and init te measures to plac he unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> d in HOT SHUTDOW within the nes'. 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS n

4.4.1.1.1 Both reactor coolant ystem rec culation loops shall be verified to be in operation at least o e per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.4.1.1.2 Each reactor c lant system recircula on loop flow control valve I shall be demonstrated OP ABLE at least once per 1 months by:

a. Verifying at the control valve fails "as 1 " on loss of hydraulic pressur at the hydraulic unit, and

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

. Less than or equal to 11% of stroke per second o ning, and

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

/ "See Special Test Exception 3.10.4.

0606270109 060620 PDR ADOCK 00000416 P PDR GRAND GULF-UNIT 1 3/4 4-1 MMENbMENT' No.

1

______-_________________-____I

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

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

b. A single recirculation loop operating with:

1) a volumetric loop flow rate less than 44,600 gym, 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, 3.2.0, and 3.3.6. #V APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*

ACTION:

I

a. During single loop operation, with the volumetric loop flow rate greater than the above limit, immediately initiate corrective action g 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 <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in HOT SHUTDOWN

{ '

)

o within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />. g I

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 <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

< f. With a change in reactor operating conditions, from two

! recirculation loops operating to single' loop operation, or

restoration of two loop operation, the limits and setpoints of S specifications 2.1.2, 2.2.1, 3.2.1, -9 tre, and 3.3.6 shall be implemented withi& J f hours or declare the associated equipment f

, inoperable (or the limits to be "not satisfied'?) and .take the l ACTIONS required by the referenced specifications.

• See Special Test Exception 3.10.4.

l

\

J14 MISC 86020502 - 9

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

Insert 3/4.4-1 page 2 of 2 SURVEILLANCE REQUIREMENTS 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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: ,

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:

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

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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />. l

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

4.4.1.1.5 During single loop operation, and with both THERMAL POWER less than g

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

a) less than 100*F, between 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 k

d c) less than 50*F, between the coolant in the operating loop and o i the coolant in the loop not in operation. g l 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, erhe- and 3.3.6 shall be verified to be within the appropriate limits within

,W hours of an operational change to either one or two loops l operating.

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l Attachment 2 Revised Integrated Package of ME0D and Core Stability / SLO Technical Specifications v

June 18, 1986 J16 MISC 86061601 - 1 j

l l

l INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY................................................... 3/4 0-1 3/S.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN.............................................. 3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES......................................... 3/4 1-2 1

3/4.1.3 CONTROL RODS Control Rod Operability...................................... 3/4 1-3 Control Rod Maximum Scram Insertion Times.................... 3/4 1-6 Control Rod Scram Accumulators............................... 3/4 1-8 Control Rod Drive Coupling................................... 3/4 1-10 Control Rod Position Indication.............................. 3/4 1-12 Control Rod Drive Housing Support............................ 3/4 1-14 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Control Rod Withdrawal....................................... 3/4 1-15 Rod Pattern Control System................................... 3/4 1-16 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM................................ 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE................... 3/4 2-1 ?

DELETED 3/4.2.2 EET"0!NTS............................................... 3/4 2-3 $

i 3/4.2.3 MINIMUM CRITICAL POWER RATI0................................. 3/4 2-4 -In 3/4.2.4 LINEAR HEAT GENERATION RATE.................................. 3/4 2-7 GRAND GULF-UNIT 1 iv hmE@tM.C .~

c -- --

INDEX i l

'\ LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION................... 3/4 3-1 8/4.3.2 ISOLATION ACTUATION INSTRUMENTATION......................... 3/4 3-9 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..... 3/4 3-27 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION ATWS Recirculation Pump Trip System Instrumentation......... 3/4 3-37 End-of-Cycle Recirculation Pump Trip System Instrumentation............................................. 3/4 3-41 3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM ACTUATION INSTRUMENTATION............................................. 3/4 3-47 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION........................... 3/4 3-52 i 3/4.3.7 MONITORING INSTRUMENTATION

( Radiation Monitoring Instrumentation........................ 3/4 3-58 Seismic Monitoring Instrumentation.......................... 3/4 3-63 Meteorological Monitoring Instrumentation................... 3/4 3-66 Remote Shutdown System Instr" mentation and Controls......... 3/4 3-69 Accident Monitoring Instrumentation......................... 3/4 3-73 Source Range Monitors.......................................

3/4 3-77 Traversing In-Core Probe System............................. 3/4 3-78 Chlorine Detection System................................... 3/4 3-79 Fire Detection Instrumentation.............................. 3/4 3-80 Loose-Part Detection System................................. 3/4 3-90 Radioactive Liquid Effluent Monitoring Instrumentation...... 3/4 3-91 Radioactive Gaseous Effluent Monitoring Instrumentation..... 3/4 3-96 g 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION.,..................... 3/43-105Q 3/4.3.9 TURBINE OVERSPEED PROTECTION SYSTEM.......................... 3/4 3-110 f 3/q.31o NEUTRON FL W M ON/ TOR f AIG /N57ROMENTA TicM . . . . . . . .

3/4 3-111 h 4

GRAND GULF-UNIT 1 v A st % utur No-

l INDEX  !

l BASES j

SECTION PAGE 3/4.0 APPLICABILITY............................................... B 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN..................................... B 3/4 1-1 I 3/4.1.2 REACTIVITY AN0MALIES................................ B 3/4 1-1 3/4.1.3 CONTROL R0DS........................................ B 3/4 1-2 l

3/4.1.4 CONTROL ROD PROGRAM CONTR0LS........................ l B 3/4 1-3 '

3/4.1.5 STANDBY LIQUID CONTROL SYSTEM....................... B 3/4 1-4 3/4.2 POWER DISTRIBUTION LIMITS i.

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE.......... B 3/4 2-1 l DELETED 3/4.2.2 of A^^M 0 T."CINTO. .................................... B 3/4 2-2 ?E E

3/4.2.3 MINIMUM CRITICAL POWER RATI0........................

a B 3/4 2-4 4 1 3/4.2.4 LINEAR HEAT GENERATION RATE......................... B 3/4 2-7 3/4.3 INSTRUMENTATION l 3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION........... B 3/4 3-1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION................. B 3/4 3-1 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION..................................... B 3/4 3-2 3/4.3.4 RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION..................................... B 3/4 3-2 ,

3/4.3.5 REACTOR CORE ISOLATION COOLING SYSTEM 1

ACTUATION INSTRUMENTATION........................... B 3/4 3-3 3/4.3.6 CONTROL ROD BLOCK INSTRUMENTATION................... B 3/4 3-3 1

GRAND GULF-UNIT 1 xit htrnEmt.MT -

l l

l INDEX

[ BASES PAGE SECTION INSTRUMENTATION (Continued) 3/4.3.7 MONITORING INSTRUMENTATION Radiation Monitoring Instrumentation............... B 3/4 3-4 Seismic Monitoring Instrumentation................. B 3/4 3-4 Meteorological Monitoring Instrumentation.......... B 3/4 3-4 Remote Shutdown System Instrumentation and B 3/4 3-4 Controls.........................................

Accident Monitoring Instrumentation................ B 3/4 3-4 Source Range Monitors.............................. B 3/4 3-5 Traversing In-Core Probe System.................... B 3/4 3-5 Chlorine Detection System.......................... B 3/4 3-5 Fire Detection Instrumentation..................... B 3/4 3-5 Loose-Part Detection System........................ B 3/4 3-6 Radioactive Liquid Effluent Monitoring Instrumentation..................................

B 3/4 3-6 Radioactive Gaseous Effluent Monitoring Instrumentation..................................

B 3/4 3-6  ;;l 4

B 3/4 3-6 4 3/4.3.8 PLANT SYSTEMS ACTUATION INSTRUMENTATION............

B 3/4 3-7 i 3/4.3.9 TURBINE OVERSPEED PROTECTION.......................

3/4.5 .10 NeuTitM Rux nwrous 1asrunwarzw. . . . . . . . .

S w s-7 1l 8 3/4.4 REACTOR COOLANT SYSTEM 4 B 3/4 4-1 l 3/4.4.1 RECIRCULATION SYSTEM...............................

B 3/4 4-2 3/4.4.2 SAFETY / RELIEF VALVES...............................

3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE i

Leakage Detection Systems.......................... B 3/4 4-2 Operational Leakage................................ B 3/4 4-2 B 3/4 4-3 3/4.4.4 CHEMISTRY..........................................

B 3/4 4-3 3/4.4.5 SPECIFIC ACTIVITY..................................

PRESSURE / TEMPERATURE LIMITS........................

B 3/4 4-4 3/4.4.6 B 3/4 4-5 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES...................

B 3/4 4-5 l 3/4.4.8 STRUCTURAL INTEGRITY...............................

B 3/4 4-5 3/4.4.9 RESIDUAL HEAT REM 0 VAL.............................. l GRAND GULF-UNIT 1 - xiii l, i

I Ammar A.  :

r f - - - - _ _ ._ . . _ _ _ . _ __ ___

l 2.0 SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS i

2.1 SAFETY LIMITS THERMAL POWER, Low Pressure or Low Flow 2.1.1 THERMAL POWER shall not exceed 25% of RATED THERMAL POWER with the reacto vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow.

APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With THERMAL POWER exceeding 25% of RATED THERMAL POWER and the reactor vessel steam dome pressure less than 785 psig or core flow less than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

THERMAL POWER, High Pressure and High Flow *J I**P *P' 1~

f4 2.1.2 The MINIMUM CRITICAL POWER RATIO (MCPR) shall not be less than 1.06 with s the reactor vessel steata dome pressure greater than 785 psig and core flow greater j than 10% of rated flow.-; h APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION: g g.

With MCPR less than-i-46!and the reactor vessel steam dome pressure greater M than 785 psig and core flow graater than 10% of rated flow, be in at least HOT SHUTDOWN within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

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

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

ACTION:

With the reactor coolant system 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 equal to 1325 psig within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> and comply with the requirements of Specification 6.7.1.

ve w.l . 54ea.m. J.., pew.sure Durig siuge loop edi s w . A n e w i .c-gre-br h 78s' Psra e~~1 cm 46 g * % io%.4cA4 ha h Mc.?R, 56 0 acrt be im +6 1 ov. I GRAND GULF-UNIT 1 2-1 AMENbMENT No.

i .

TABLE 2.2.1-1

, REACTOR PROTEC110N SYSTEM INSTRUMENTATION SETPOINTS i i 4

ALLOWR8LE

FUNCTIONAL tl NIT TRIP SETPolNT VALUES l c4 5 1. Intermediate Range Monitor, Neutron Flum-High 5 120/125 divisions < 122/125 divisions y of full scale of full scale g 2. Average Power Range Monitor

i Q a. Meutron Flum-High, Setdown i 15% of RATE 0 1 20K of RATED g THERMAL POWER THEllMAL , POWER i b. Flow Blased Simulated Thernal Power-Hloh

! o r, Otre ~< 0.66 W+49%, with e

- 9.00 G G, with 5dm.hal Id5 FAT for a == = ' - ;, a maximum of on N 31.1%

i Th a t g ,g_;

f) High Flow C1_ ~ d i 111.uz oi Z 0 < 113.0E of RATED 9. inwr4 b,

THERMAL P0eER r rm _ _ r_ra j.gg

c. Neutron Flum-High i 118% of RATED < 120% of RATED doap *4 THERMAL POWER THERMAL F0WER WS eequest

d. Inoperative NA NA

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

4. Reactor Vessel Water Level - Low, Level 3 1 11.4 inches above 1 10.8 Inches above j Instrument zero" instrument zero"

! 5. Reactor Vessel Water Level-High', Level 8 1 53.5 Inches above 1 54.1 inches above j instrumeni. zero" instrument zero*

j 6. Main Steam Line Isolation Valve - Closure < 6% closed i 75 closed j 7. Main Steam Line Radiation - High 1 3.0 x full power < 3.6 x full power

, background background I

8. Drywell Pressure - High i 1.23 psig i 1.43 psig submiManon l S. _ Scram Discharge Volume Water Level - High  ; .- .. ... ---..  ; .. .. ... . ..

g i 10. " Turbine Stop Valve - Closure 1 40 psig** 1 37 psig

11. Turbine Control Valve Fast Closure, i3

! e Trip 011 Pressure - Low 1 44.3 psIO** 1 42 psig f( 12. Reactor Mode Switch Shutdown Position NA NA l3 i to

13. Manual Scram NA NA

! f.a. "See Bases Figure 8 3/4 3-1.

1 ** Initial setpoint. Final setpoint to be determined during startup test program. Any i equlicit thavige tie this setpoint shall be submitted to the Commission within 90 days of test completlon.

j a. Transmitter / Trip Unit g 60% of full scale < 63% of full scale I b. Float Switch 5 64" 2 65" 1

i

)

1 INSERT for Table 2.2.1-1, item 2b (MEOD) 2.b 1) During two recirculation loop operation 3 0.66 W+64%, with 6 066 W+67%, with s.L4%d a) Flow Biased .n s -a -s4 a maximum of a maximum of b) High Flow Clamped $111.0% of RATED $113.0% of RATED THERMAL POWER THERMAL POWER

2) During single recirculation loop operation:

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

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

4 k

l 4

}

4 l J12 MISC 86030402 - 2 i

i

2.1 SAFETY LIMITS BASES M ,

2.0 INTRODUCTION

l 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 5 during nomal plant operations and anticipated transients. The fuel cladding 1

• i

' integr.ity Safety Limit is set such thatfuel Because no damage fuel damage is calculated is not directly to occurl g observable.

if the limit is not violated. y]

a step-back approach is used to establish a Safety Limit =9'th:t the MCPR ++-  ;

nt b; th= 1.05. MCPR greater than 4:4HF represents a conservative margin The .4 l relative to the conditions required to maintain fuel cladding integrity.

fuel cladding is one of the physical barriers which separate the radio materials from the environs. Although some corrosion to its relative freedom from perforations or cracking.

or use related cracking may occur during the life of the cladding, fission product migration from this source is .incrementally cumulative and continuousl; measurable.

Fuel cladding perforations, however, can result from thermal stresses which occur from reactor operation significantly While fission above design product condi-migration tions and the Limiting Safety System Settings.

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 l margin to the conditions which would produce onset of transition boiling, MCPR of 1.0. These conditions rcpresent a significant departure from the condition j

intended by design for planned operation.

l 2.1.1 THERMAL POWER, low Pressure or low Flow l

' 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. i Therefore, the fuel cladding integrity Safety Limit is established by other means.

This is done by establishing a limiting condition on core THERMAL POWER with the following basis. Since the pressure drop in the bypass region is

.i  ;

i 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 108 lbs/hr, bundle pressure drop is nearly independent 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 10 8 lbs/hr. Full scale ATLAS test data taken at j

pressures from 14.7 psia to 800 psia indicateWith thatthe thedesign fuel assembly critical peaking factors, power at this flow is approximately 3.35 MWt.this corresponds to a THE Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. ,

l GRAND GULF-UNIT 1 B 2-1 jfew&,

i SAFETY LIMITS i

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 retult in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from 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 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 is determined

, which is a statistical model that using combines the General all of theElectric Ther 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 vaMd over the range of conditions used in the tests of the data used to develop the correlation.

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 I.

Bases Table B2.1.2-2. i, n

Thebgsesfortheuncertaintiesinthecoreparametersaregivenin i NEDO-20340 and the bgsis for the uncertainty in the GEXL correlation is .e niven in NED0-10958-A .4 The power distribution is based on a typical 4, 764 assembly core in which the rod pattern was arbitrarily chosen to produce 'E a skewed power distribution having the greatest number of assemblies at the ~f highest power levels. The worst distribution during any fuel cycle would not be as severe as the distribution used in the analysis.

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

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

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3m.p yr Ay ses k %c GsWs "A$e Lamp oph 4=.ly+b d=Aed re nary.A., lo.p,e, , #8G .

GRAND GULF-UNIT 1 B 2-2 AMENDMEN T 0

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(p.)

3.0 g Channel Flow Area '

Friction Factor Multiplier 10.0 $

Channel Friction Factor $

Multiplier 5.0 ] ,

TIP Readings 6.3 (b) a R Factor 1.5 Critical Power 3.6

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

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'1MITING SAFETY SYSTEM SETTINGS

! BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETP0!NTS (Continued)

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

The APRM trip system is calibrated using heat balance data taken during steady state conditions. Fission chambers provide the basic input to the sys-tem and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Neutron Flux-High 118% setpoint; i.e, for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel. For the Flow Biased Simulated Thermal Power-High setpoint, a time constant of 6 1 1 seconds is introduced into the flow biased APRM in order to simulate the fuel thermal transient characteristics. A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1. Lssgf p The APRM setpoints were selected to provide adequate margin for the Safety Limits and yet allow operating margin that reduces the possibility of unneces- g sarv shutdown. ITTiFfhnr-ceteranW trip setpoint aus b specified formula in S

.. tain these margins 4 . . TP. . ._ ._ . _

3. Reactor Vessel Steam Dome Pressure-High ~$

High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products. A pres-sure increase while operating will also tend to increase the power of the reactor by compressing voids thus adding reactivity. The trip will quickly reduce the neutron flux, counteracting the pressure increase. The trip set-ting is slightly higher than the operating pressure to permit normal operation without spurious trips. The setting prov< des for a wide margin to the maximum allowable design pressure and takes into account the location of the pressure measurement compared to the highest pressure that occurs in the system during  !

a transient. This trip setpoint is effective at low power / flow conditions when l

l the turbine stop valve closure trip is bypassed. For a turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.

GRAND GULF-UNIT 1 8 2-7

/}magpr Nc ,

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

INSERT F .

Slow In these, biased equations, the variable w, is the loop recirculation flow as a percentage of the total loop recirculation flow which produces s rated core flow of 112.5 million 1bs/hr.

i a

j d

l t

i 1

l l

l

LIMITING SAFETY SYSTEM SETTING BASES t

l REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued)

9. Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram. Should this volume fill up to a point where there is insufficient voltme to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered. The reactor is therefore tripped when the water level has reached a point nigh enough to indicate that it is indeed filling up, but the volume is still great enough to accosumdate the water from the movement of the rods at pressures below 65 psig when they are tripped. The trip setpoint for each scram discharge volume is equivalent to a contained volume of 26 gallons of water.

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

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

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

) [

The turbine control valve fast closure trip anticipates the pressure, 4 neutron flux, and heat flux increase that could result from fast closure of y 9

the turbine control valves due to load rejection coincident with failure of 1

the turbine bypass vahes. The Reactor Protection System initiates a trip when j fast closure of the control valves is initiated by a low EHC fluid pressure in d the control valve and in less than 100 milliseconds after the start of control 1

valve fast closure. This loss of pressure is sensed by pressure transmitters 3 which output to trip units whose contacts form the one-out-of-two twice logic -1 j "

input to the Reactor Protection System. The trip setpoint is 43.3 psig. This 1 trip setting and a different valve characteristic from that of the turbine stop [

l valve combine to produce transients which are very similar to that for the stop \

j valve. Relevant transient analyses are discussed in Section 15.2.2 of the Final Safety Analysis Report. (.x:K93g r (vs }

! 12. Reactor Mode Switch Shutdown Position

The reactor mode switch Shutdown position provides trip signals into system trip channels which are redundant to the automatic protective instru-mentation channels and provides additional manual reactor trip capability.

13. Manual Scram f

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

/ channels and provides manual reactor trip capability.

! GRAND GULF-UNIT 1 8 2-9 i___.___.__..__,__..__--__. . _ ___ __ . . _ _ . , _ _ , _ , _ _ _ _ _ , _ _ . _ _ _ _ _ - _ _ _ . . _ _ _ , , _

I INSERT K The automatic bypass setpoint is feedwater temperature dependent due to the subcooling changes that affect the turbine first stage pressure-reactor power relationship. For RATED THERMAL POWER operation wf*.h feedwater temperature greater than or equal to 420*F an allowable  ;

setpoint of 6 26 % of control valve vide open turbine first stage i

pressure is pr ded for the bypass function. This setpoint is also l i

i applicable to operation at less than RATED THERMAL POWER with the correspondingly lower feedwater temperature. The allowable setpoir.t is l reduced to f,22.5% of control valve vide open turbine first stage ,

pressure for RATED THERMAL POWER operation with a feedwater temperature between 370*F and 420*F. Similarly, the reduced setpoint is applicable to operation at less that RATED THERMAL POWER with the correspondingly lower feedwater temperature.

I i

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

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

l

- - - , , -,, e ,,, - --,... - , - - ~ - -y-- , ,, --se ---- g - - - - - - - - w-~w--m- y-we +v- e ,

us muliiphed by 'iht s mallt r c,( cMkete ike

{Wa- d epend erd MATLH e,R (ach, (Marv Ac 4) o( ~4 Que e 32.1-2, or tke pe r d t.pendud

3. 7. i - 3 f4AOL.d CR (a cie.< (MAM Ach e{ T9 :e M.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE Y S

LIMITING CONDITION FOR OPERATION I

• t 3.2.1 All AVERAGE PLANAR LINEAR HEAT GENERATION RATES (APLHGRs) for each type Gf fuel as a function of AVERAGE PLANAR EXPOSURE shall not exceed the limits shown in Fiaure 3.2.1-1/ W um .e ~ , ,

OPERATIONAL CONDITION 1, when THERMAL POWER is greater than APPLICABILITY:

er equal to 25% of RATED THERMAL POWER.

ACTION: r..& so b m hl --*

en Mar 31,lB With an APLHGR exceeding th limits of Ti; a 3.2.! 1, initiate corrective cction within 15 minutes and restore APLHGR to within the required limits c:ithin 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

SURVEILLANCE REQUIREHENTS

{c. ppt. w s t,4h1 en & 3i.Mc 4.2.1 All APLHGRs shall be verified to be equal to or less than the limits :

M e.e.i.;d f.e; Ti;..e 0.0.1 1.

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

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

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

d. The provisions of Specification 4.0.4 are not applicable.

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

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

1 .

o 14 I I I I I l l E CURVE FUEL TYPE E 8CR2iO A

o B 8CR l60 j E C 8CRO71 _

t T

c g3 _

h ys A i26 12 V _ 12.4 12.6 112.4 12.6

/

A I y

'2

12.O ' 12.i L El ,,.

8 e s's ii.s -

ii.s c

z (ii.7it4 itsj 9 its it.2 ii.I

11. 0 _

g gg _

'o 4 iO.4 5 y, 2 w 3 io -

9.7 h 9.6 97

] 9.0 9 -

99 1

l i i 1. I I I I

! e 35,000 40,000 O 5,000 10,000 15,000 20,000 25,000 30,000 A AVERAGE PLANAR EXPOSURE (mwd /t) i i

1 FIGURE 3.2.1-1 MAXIMUM AVERAGE PLANAR LINEAR HEAT GENERATION RATE (MAPLHCR)

,b*

g VERSUS AVERAGE PLANAR EXPOSURE -=

INITIAL CORE FUEL TYPES 8CR210, 8CR160 AND 8CR071 i

fa rwn avenw _

.,_ma- --a - .- - . _ . - - _a

,l t

/i4 I a l 5

= 4 3 . I,. , .

y

\ - ,

1 - - e

.f i 3  : 3 - kI'

=i.g ,

.g ... -

5,8  :

5e i e  : - I. =gi C Wo S

l i  %. 5 I

• 3a

- I. *= "-

a E y~$

i I.IlI ,E

=

.b. .

=

l a

. I.

e l y=kg

g o

i

- n

- I*

- e

a E l

I

?

. i . .

g3 e e > e

. = 3 (48/M) 31VW N011VW3N30 AV3M Nf3 Nil WVNvid 30VW3AV MnnfXVM I

--swo s u - unT 1 sa r-sa neoen m. __

1 I 1.1 s

//  ;

t-CLAMPED DURING ~

0.9 O '

E - FOR MAX FLOW = 102.5%

/r/ l . NE LOOP OPERATION

.s

< 0.8 0

$ 2 FOR MAX FLOW = 107.05 l s Q

g 0.7 ,

) -

of l E $a a 4 i g 2

0.6 0

0.5 0 to 40 60 80 10 0 120 CORE FLOW (% RATED), F FIGURE 3.2.1-2 MAPFACf 1

GRAND GULF UNIT ONE 3/4 2-te

1.1 4.0 i

0.9 "L

U t

FOR P>70% ;

3 0.8 FOR 25%1P140%;

r DURING ONE LOOP OPERATION =

E q CORE FLOW FS50%

r FOR 40%<Ps 100%; Y h

g ALL CORE FLOWS {

g a

xV / 4 q

L r w.

0.6

/

FOR 25%$ P$40%;

O.5 CORE FLOW F > 50%

0.4 0 20 40 SO 90 00 0 12 0 CORE THERMAL POWER (% RATED) P l

FIGURE 3.2.1-3 MAPFACp i

GRAND GULF UNIT ONE 3/4 2-Eb AMENDMENT No. __

y,ceyeo 6 ...

i,,. .r. u..,  % . s . 7 r

Te.,dw 4.A -

. sw. WLe g n.

I POWER DISTRIBUTION LIMITS 55 (o 4*dT 55 +45%) T T

~

I;f 3/4.2.2 APRM SETPOINTS g ag, w , 545) T Seaflo hs W' '

i LIMITING CONDITION FOR OPERATION e T

.2.2 The APRM flow biased simulated thermal power-high scram trip setpoint k and flow biased neutron flux-upscale control rod block trip setpoint (5 g) 1 c

sha 1 be estab]ighed gcording to the following relationships:

Tris Setpo Y ,

ANowableValue $d 5 < (0.66W + 48%)T 5 < (0.66W + 51%)T 5 sgg 5 (0.66W + 45%)T j s,g 5 (0.66W + 42%)T where: 5 S RB a e in Percent of RATED THERMAL POWER.

h W =L p recirculation flow as a percentage of th loop recirculation million 1bs/hr.

fic%whichproducesaratedcoreflowof11 T = Lowest alue of the ratio of FRACTION OF ATED THERMAL POWER (FRTP) 'vided by the MAXIMUM FRACTION LIMITING POWER DENSITY (MFLPD). is applied only if less t n or equal to 1.0.

APPLICABILITY: OPERATIONAL ONDITION 1, when TH L POWER is greater than or equal to 25% of RATED THERMAL OWER. 3 ACTION:

With the APRM flow biased simulate hern power-high scram trip setpoint and/

or the flow biased neutron flux-upsc e ontrol rod block trip setpoint less M conservative than the value shown in allowable value-column for 5 or S i as above determined, initiate correc ve etionwithin15minutesandrestNe, S and/or S u to within the require imit within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or reduce THERMAL POWER to 1 H s than 25% of RATED T RMAL within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

$URVEILLANCE REQUIREMENTS , s

! 4.2.2 The FRTP AND MFLPD r each class of fuel 11 be determined, the value

! of T calculated, and the st recent actual APRM f1 biased simulated thersal power-high scram and f1 biased neutron flux-upscale ontrol rod block trip j setpoints verified to e within the above limits or ad sted, as required:

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

c. In ially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the rea r is operating

th MFLPD greater than or equal to FRTP.

d. The provisions of Specification 4.0.4 are not applicable.

• W h MFLPD greater than the FRTP during power ascension up to 905 of RAT ERMAL POWER, rather than adjusting the APRM setpoints, the APRM gain may e l adjusted such that APRM readings are greater than or equal to 1005 times MFL

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 M EWDMfur' Me.

he t.gct. wit A Ntw F ic u gt 3.2.3- t l

I I

l \ /

\ j 1..

l

\,\ /

g,4

/

s

\.

x ,

\

/ NN i

~

/ %g '

d MCPR '

t Limit

[

i u / \

[ \

l gn.

0 10 40 40 to 10 0 120 l

c . rio., s or no,.a cor. rio.

MCPR, l FIGURE 3.2.3-1 (Tit MD mtN"i Nt. .

Rtpeg et. wiy a ut w F i cu ct. 3. 7. 3 - 2.

s

1. \

l.6 N r

/ 1

(

\ \ /

N Nx /

x x

[ \ / 4 3

A

~ / NN i 1.2 g

/ NN S i

/ \

II

/ \

/

O to 40 SO 80 100

\ 20 Thermot Power. % of Roted Thermot Power  ;

MCPR l FIGURE 3.2.3-2 GRAND GULF-UNIT 1 3/4 2-6 g go,

SLAJ .s s..a . n gg) l 1.7  !

l l

16

\

15

\\ FOR MAX FLOW = 107.05

( ,

(

'~'

FOR MAX FLOW = 102.5%

/ N\

% l.3

\ \

1.2

\\ g, i

l s.1 1

RATED NCPR OPERATING LIMITal.18 l l ' ' l 10 0 20 40 00 00 00 0 12 0 CORE FLOW (% RATED), F i

FIGURE 5.2.5-I MCPR g l

l l

GRAf0 GULF UNIT ONE 5/42-5

- - - , - - - - - y , , , -,----y....,-,,,.-,--,__y, _

..,,,%,,., w-__

gaatfd .~ s 2 36 s

5 I l 5 3 TERMAL POWER 25% s.Pi40%

2.2 FLOW > 50%

2.O eTHERNA _ POWER 25% $ P$40%

CORE FLOW 5 50%

b i.8 I.6

THERNAL POWER 4 No < P S 70%

ALL CORE FLOWS a

E w O

2 ,,4 i HERMAL POWER P>70%

l 4L CORE FLOWS

! K xs r.2 10 O 20 40 60 80 00 0 120 CORE THERMAL POWER (% RATED) P FIGURE 3.2.5-2 NCPR p 4

[. - - - -

GRAND SULF - UNIT I 3/42-6

o i TABLE 3.3.1-1 (Continued)

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

(b) The " shorting links" shall be removed from the RPS circuitry prior to and during the- time any control rod is withdrawn

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

(c) An APRM channel is inoperable if there are 1er.s than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.

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

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

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

(g) With any control rod withdrawn. Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2. uffy, bs/ow t4E 9,Appaapaink (h) This function shall be automatically bypassed when turbine first stage [gg+ d V

pressure is_ less than:IO%** of -the .wal_ue of turbine fi genure~

P in psia, rJr valves wide open_LYWD) ent to THERMAL POWER _

_less_than4%-of-RATEDTHERMAL POWER. _

ci)$2e.Y4"4 -tLe wedut of, 4ur bine fest stSt pr essur e ej 45 uAde ge.n G wo) s M Mew du cptb g JiN q .

gM (ud Mr bys_s alu re c{ og ech.r 1ko s er egurd de Wf $

4

%522.s*/? ck de *lue c4 4ur6 Crs4 s4ay pressure e Vt0 C st % { \ow w La.n opero d w k rob M b W 8 hp%4 ur( Ig4 m aan 3 To*F h el 4 2 0 F. k "Not reuuired for control rods removed Der SDeCifiCation 3.9.10.1 or 3.9.10.2. j

""Initisi .et; 9 + Final setpoint to be determined durino p ar+ O te.i. pr6 gram. 9 Any required change to this w#* "" ,e uumitted to the Commission +-

J within 90 dag Of t..i. completion.

H AucuacMef vo.lues c( iue bine. { irs 4 319 pressure e.guiu ste_ni h -

wa:nw wtr less , Lo.n 4o% ( g ATc p Tu t stra a t roon.

. $gfpoINt GRAND GULF-UNIT 1 3/4 3-5

[\mm't)MwT No-

INSERT G Pressure setpoint of; i

l I

L I

c. Trecomitter/ Trip tlnit 5 M st IR) 1.2.5 U) 1.2.5 U)

~

b. Float Switch NA M R TA8tt 4.3.1.1-1 (Continued)

REAC70ll PROTECfl0N SYSTEM INSTilUNElffAfl0N SURytltLAIICE REQUIREMENTS OPERATIONAL E CHANNEL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR tellCM

' y; e FtIIICTfBML INelf _ CHECK _ TEST CAllBRATION SURVEltLANCE REQUIRE 0 k 9. Scram Discharge Volume 1 tater gi3 SuMal on

--M- -#gg). 1, 2, 3

-4 t

Level - High 0 h 2% N w II)

10. Turbine Stop Valve - Closure 5 M R 1 1 11. Turbine Control Valve Fast Closure Valve Trip System 011 I9)

Pressure - Low 5 M R 1

12. Reactor IInde Switch 1,2,3,4,5 Shutdown Position MA R NA i

13. Manual Scras IIA M MA 1,2,3,4,5 j

i T*

(a) IIeetron detectors may be excluded from CHAlWlEL Call 8RATI0lt.

T (b) The IIIM and SilM channels shall be determined to overlap for at least 1/2 decade during each startup af ter entering OPERAi!0NAL CONDITION 2 and the IllM and APRM channels shall be deter-88 eined to overlap for at least 1/2 decade during each controlled shutdown, if not performed within the previews 7 days.

(c) (DELETED)

(d) This calibration shell consist of the adjustment of the APflM channel to confors to the power values

! calculated by a heat balance during OPERAfl0NAL CON 0lil0N 1 when THERMAL POWER > 25% of RATE 0

' THEllMAL POWER. Adlust the APitM_ channel if the_ absolute difference is greater LEen 2% of RATED 34,agg ,

_ THERMAL POWER.I Any nrnri cn.....ei v.;.. Mjr9 - ' =

• in compliance with Specification 3.2.2 .r. a . s t.

I shall not be included in determinine the absolute dif ference.

y (e) thlt Calibration shall Consist of the adjustment of the APRM flow biased channel to conform to a calibrated flow signal.

g

' (f) The LPRMs shall be calibrated at least once per 1000 MWD /T using the ilP system.

(g) Calibrate trip unit at least once per 31 days.

m (h) Verify measured drive flow to be less than or equal to established drive flow at the existing flow con

• l96 13 . trol valve position.

fu (1) This calibration shall consist of verifying the 6 i 1 second simulated thermal power time constant.

(j) llot appilcable when the reactor pressure vessel head is unbolted or removed per Specification 3.10.1.

h l

(k) Not appilcable when ORYWELL INTEGRITY is not required.

(1) Appilcable with any control rod withdrawn. Not applicable to control rods removed per Speiltfra-

!p i

tion 3.9.10.1 or 3.9.10.2.

1, i

l

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

1. Turbine Stop Valve - Closure 2(b)

2. Turbine Control Valve - Fast Closure 2 fD) pnm Bet w) n.w APPO.cmW

zes w za.

se wa n w (a) A trip system may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> for required surveillance provided that the other trip system is OPERABLE.

(b) This function shall be automatically bypassed when turbine first stage 7

) _,

pressure is less than!30%* ef the vslue of turbin; fir:t :t;g: pr:::;rc,  %

1 in p;i;, et v;1v;;E nATPM

.- AL__ A M&'

widT11P ;p MLA n (VE'0)

A t

t

:: fi w, ;;uiv; lent to T"E""" L MAi ff* n

"^h'E", G sEJJ b' Ep i s Y vis vs nyggby a s s6 ns wik u w rT b n . I

• nitial ;;tpcint, fin:1 ::tpcint i: bedeter'n:ddurng:t:rtupt::tpr;rb i

k

^ny

. r quired chang t; thi; ;;tpcint ; hall 5: ;;br,itted t; th: C;x,i;;i;n wiihin "O d y: f 1::t :: p!:ti n- ]

g See. DGER_T~

3/43-4'S

)

i I

1

- GRAND GULF-UNIT 1 3/4 3-43 g'

INSERT 3/4 3-43 a)4 26.9 I of the value of turbine first stage pressure at valves wide open(WO) steam flow when operating wigh rated feedwater temperature of greater than or equal to 420 F; or b)h22.5% of the value of turbine first stage pressure at Wo steamflowghenoperagingwithratedfeedwatertemperature

• between 370 F and 420 F. Jetpo

• n+

s of turbine first stage pressure These represent allowabl equivalent ta THERMAL POWER less than 401 of RATED THER.M PokTR.

i 1

l I

l l

4 TABLE 3.3.6-2 CONTROL ROD BLOCK INSTRUMENTATION SETPOINTS

h 5 TRIP SETPOINT ALLOWABLE VALUE TRIP FUNCTION 1

I g

1. R0D PATTERN CONTROL SYSTEM e

20 + 15, -0% of RATED THERMAL 20 + 15, -0% of RATED THERMAL POWER E a. Low Power Setpoint POWER Z < 70% of RATED THERMAL POWER

~ b. High Power Setpoint i 70% of RATED THERMAL POWER ,

2. APRM Trot.nct wit 4 14stnr (or Tani.e 3.3.(e-2

a. Flow Biased Neutron Flux- \

~

g g ; g ,,

Upscale  : (0.00 t! 42%)T* < (0.55 " ^ 15%)T* g,, 3,, is NA .

i b. Inoperative NA

, ,j],hg

• , 4

c. Downscale 1 4% of RATED THERMAL POWER 1 3% of RATED THERMAL POWER l g g;,

! d. Neutron Flux - Upscale Startup i 12% of RATED THERMAL POWER 114% of RATED THERMAL POWER ,

w 3. SOURCE RANGE MONITORS s NA

'

• a. Detector not full in MA 5 5

< 1.5 x 10 cps

b. Upscale < 1 x 10 cps l V NA

$ c. Inoperative RA

d. Downscale 1 0.7 cps 1 0.5 cps

4. INTERMEDIATE RANGE MONITORS l NA l

a. Detector not full in MA

< 110/125 of full scale l b. Upscale < 108/125 of full scale NA

c. Inoperative NA

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

5. SCRAM DISCHARGE VOLUME lp a. Water Level-High i 32 inches i 33.5 inches E 6. REACTOR COOLANT SYSTEM RECIRCULATION FLOW z i 111% of rated flow i

, -.y a. Upscale i 108% of rated flow

7. REACTOR MODE SWITCH SHUTDOWN NA p 1 P0SITION NA _.

2.

!? *The Average Power Range Re.-ii.v. .cf black function is varied _me = f=:ti . vi rectrculation loop flow  ;

(W) 'and the ratio of FRACTION af DATEL'T;;EEi. ruwtn i.o tt: = W MUM FRACTION of LIMITING POWER DENSITY l Tne THp setting of this function must be maintained in accoruam.c with-Spec 111 cation 3.2.2.  %

i __U_ M irr). -~ g, i

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

1) During two recirculation loop

' operation a) Flow Biased 3 0.66 W+58%, with 50.66 W+61% with i

a maximum of a maximum of b) High Flow Clamped i108.0% of RATED 5110.0% of RATED THERMAL POWER THERMAL POWER l 2) During single recirculation loop operation:

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

b) High Flow Clamped Not required Not required

OPERABl.E OPERABLE l

i l

J12 MISC 86030402 - 1

I i

INSTRUMENTATION TABLE 4.3.6-1 (Continued)

CONTROL ROD BLOCK INSTRUMENT ATION SURVEILLANCE REQUIREMENTS 900TES:

a. Neutron detectors may be excluded from CHANNEL CALIBRATION.

b. Within 7 days prior to startup.

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

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

e. [ Deleted]

This calibration shall consist of the adjustment of the APRM channel g

f. i, to conform to the power values calculated by a heat balance during 4 OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 5

25% of RATED THERMAL POWER.

Adjust the APRM channel if the absolute _

difference is eater than 2% of RATED THERMAL POWER. t "m ; CiiiiFnel g ica .2.2 shall not *.

ining the 4

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

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

Not applicable to control rods removed i

• With any control rod withdrawn.

per Specification 3.9.10.1 or 3.9.10.2.

1 i

3/4 3-57 GRAND GULF-UNIT 1 Annt;nwwt N e. -

INSTRUMENTATION 1

1 3/4.3.10 NEUTRON FLUX MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 1

i 3.3.10 The APRM and LPRM* neutron flux noise levels shall not exceed 1 three (3) times their established baseline value.

APPLICABILITY OPERATIONAL CONDITION 1 with operation in Region I as specified 4

in Figure 3.4.1.1-1.

With no established baseline flux noise levels, immediately V

ACTION a.

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 l

Region II as specified in Figure 3.4.1.1-1 within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

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

} Figure 3.4.1.1-1 within the next 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. A b

SURVEILLANCE REQUIREMENTS 3

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

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

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

l

, l i

i

• 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 i monitored.

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

J14 MISC 86020502 - 1 l

-,-r- --- ,,-v_m.~.,,.,-------..-,--,,s.,- . - - - - - - . - , , , ~ . . , _ . , , - . , _ ...

,-~-,,a.- . -, - ,_ w.-.,..-,,,-.---c-,,-- .me e----,e-. ,., -,,

A 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 4 3.4.1.1-1 provided the baseline has not been established since 1 the last CORE ALTERATION; or -5

b. In lieu of establishing single loop baseline data, the baseline )4 established in 4.3.10.2 may be utilized for single loop operation in Regions I of Figure 3.4.1.1-1.

i e

i M GlLF - lMT 1 3/!4 3-112 k e 0 MENT No. ___

i

i i

l 3/4.4 REACTOR COOLANT SYSTEM g 3/4.4.1 RECIRCULATION SYSTEM

-o m4 RECIRCULATION LOOPS l IAISEfr LIMITING CONDITION FOR OPERATION f

. 4.1.1 Two reactor coolant system recirculation loops shall be in operati .

AP CABILITY: OPERATIONAL CONDITIONS 1* and 2*.

ACTION:

a. With e reactor coolant system recirculation loop not in eration, immedia ly initiate an orderly reouction of THERMAL POW to less than ,

or equal 80% of the 100% Rod Line as specified in F* ure B 3/4 2.3-1, and be in a least HOT SHUTDOWN within the next 12 h rs.

b. With no reactor oolant system recirculation loo in operation, immediately initi e an orderly reduction of T MAL POWER to less than or equal to 8 of the 100% Rod Line as ecified in Figure B 3/4 2.3-1, and init te measures to plac he unit in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> d in HOT SHUTD0 within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

SURVEILLANCE REQUIREMENTS m

4.4.1.1.1 Both reactor coolant ystem rec culation loops shall be verified to be in operation at least o e per 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> 4.4.1.1.2 Each reactor c lant system recircula on loop flow control valve shall be demonstrated OP ABLE at least once per 1 months by:

a. Verifying at the control valve fails "as i " on loss of hydraulic pressur at the hydraulic unit, and

b. Ver ying that the average rate of control valve vement is:

. Less than or equal to 11% of stroke per second o ning, and i

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

"See Special Test Exception 3.10.4.

GRAND GULF-UNIT 1 3/4 4-1 MNENDMEar No.

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

a. Two recirculation loops operating with limits and setpoints per Specifications 2.1.2, 2.2.1, 3.2.1, 0.2.2 and 3.3.6, or f

) b. A single recirculation loop operating with:

1) a volvaetric loop flow rate less than 44,600 syn, and

~

l 2) the loop recirculation flow control in the manual mode, ad

! 3) limits and setpoints per Specifications 2.1.2, 2.2.1, APPLICABILITY:

3.2.1, 0.0.0, and 3.3.6.

f l l OPERATIONAL CONDITIONS 1* (nd 2*

l ACTION:

i I

s. During single loop operation, with the volunetric loop flow rate

greater than the above limit, immediately initiate corrective action g to reduce flow to within the above limit within 30 minutes.

i b. During single loop operation, with the loop flow control not in the -

annual mode, place it in the manual mode within 15 minutes.

c. With no reactor coolant system recirculation loops in operation, j insediately initiate an orderly reduction of THERMAL POWER to within 3

Region III as specified in pigure 3.4.1.1-1, and initiate measures to place the unit in at least STARTUp within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in NOT SHUTDOWN {o within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.

g

d. During single loop operation, with temperature differences exceeding k the limits of SURVEILLANCE REQUIREMENT 4.4.1.1.5, suspend the j THERMAL POWER or recirculation loop flow increase.

4

e. With operation in Region IV as specified in Figure 3.4.1.1-1,

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

f. With a change in reactor operating conditions, from two recirculation loops operating to single loop operation, or i restoration of two loop operation, the limits and setpoints of l

g specifications 2.1.2, 2.2.1, 3.2.1, +reet, and 3.3.6. shall be f implemented withitT)f hours or declare the associated equipment inoperable, (or the limits to be "not satisfied") and.take the l ACTIONS required by the referenced specifications.

i

• See Special Test Raception 3.10.4.

I J14 MISC 86020502 - 9 -

J l

l l

l l l

l Insert 3/4.4-1 page 2 of 2 gURVEILLANCE REQUIREMENTS j 4.4.1.1.1 The reactor coolan: 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 <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

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

a. Verifying that the control valve fails "as is" on loss of ,

I hydraulic pressure at the hydraulic unit, and I

]

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

1

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

2. Less than or equal to 11% of stroke per second closing. j 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 <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.

I

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

4.4.1.1.5 During single loop operation, and with both THERMAL POWER less than g I 36% of RATED THERMAL POWER and the operating recirculation pump not j j on high speed, verify the following differential temperature require- i ments are met within 15 minutes prior to beginning either a THERMAL '

p0WER increase or a recirculation loop flow increase and within every l l hour during the TRERMAL POWER or recirculation loop flow increase:

{

a) less than 100'F, between 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 i operation and the coolant in the reactor vessel, and c) less than 50*F between the coolant in the operating loop and a

1 the coolant in the loop mot in operation. a 4

The differential temperature requirements 4.4.1.1.5.b and e do not

apply when the loop not in operation is isolated from the reactor pressure vessel.

I 4.4.1.1.6 The limits and setpoints of specifications 2.2.1, 3.2.1, 9rtre and

] 3.3.6 shall be verified to be within the appropriate limits within X bours of an operational change to either one or two loops l

) operating.

t J14 MISC 86020502 - 10 .  !

l

,susmm shi/es 8

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8 a

I . g e -

S s I 5

su Rh g

• s' IEl

_ g i s E e

S g

. g*

_ _ - - - - _ _ _ _ . _ _ - - b H ===-

h-g  ;. '

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E B R S E S S e

• S R 2 M3 mod wnW3H1031VW A0 AN33W3d l

1. 40 81F - WIT 1 3A 4-3A m m,

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 purg: inoperable, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.

SURVEILLANCE REQUIREMENTS ,

in an opm+g loor I 4.4.1.2.1 EachoftheaboverequiredjetpumpskhallbedemonstratedOPERABLE 3+

with THERMAL POWER in excess of 25% of RATED THERMAL POWER and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> by determining recirculation loop flow, total core flow and I diffuser-to-lower plenum differential pressure for each jet pump and verifying .j that no two of the following conditions occur: A:r 50th %di::ted r :ir &-

[g ti;r. 1;;; i k e ;r; i, ;;r. plier,ca ith Specificatier. 0.4.1.0.

w

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 <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> l after entering OPERATIONAL CONDIT0N 2 and at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> thereafter.

• Initial value. 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 NENdMENT Mo.

1 1

REACTOR COOLANT SYSTEM .

I RECIRCULATION LOOP FLOW LIMITING CONDITION FOR OPERATION

,wken N. l p. ar. h ,p.-b, 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. i

b. 10% of rated recirculation flow with core flow less than 70% of rated d; core flow. /,,

f, APPLICABILITY: OPERATIONAL CONDITIONS 1* and 2*.

~4 ACTION:

• i With recirculation loop flows different by more than the specified limits, 4

.:.u...

+ L/estoretherecirculationloopflowstowithinthespecifiedlimit within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).* M un,ocu+,41,eA r:

4-a,J e. mph wAMe vr. ea n+S h h:1:r ene q: .5: recirculation loop uitt th: i:wcr fica n:t ir. :p;r:ti n end take the ACTICN requir by-Specification 3.4.1.3A ,or-

b. Be 'ia adI-A McT sumu wh 82 h o w,. .

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

R See Special Test Exception 3.10.4.

• l GRAND GULF-UNIT 1 3/4 4-3 Amam7 N, __,,

l 4 .

l

. REACTIVITY CONTROL SYSTEMS I 1

! BASES  !

~

3/4.1.3 CONTROL RODS l The specifications of this section ensure that (1) the minimum SHUTDOWN  !

l MARGIN is maintained, (2) the control rod insertion times are consistent with 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.

l 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-l 1ess with rod drives will be investigated on a timely basis.

l Damage within the control rod drive mechanism could be a eneric problem, i

therefore with a control rod immovable because of excessive fr ction 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  !

l prevtat operation with a large number of inoperable control rods.

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

] consistent with the SHUTDOWN MARGIN requirements.

l The number of control rods permitted to be inoperable but trippable could

! be more than the eight allowed by the specification, but the occurrence of eight a

) inoperable rods could be indicative of a generic problem and the reactor must  ?

j be shut down for investigation and resolution of the problem. g gy.,g,g y

! The control rod system is designed to bring the reactor subcritical at a i

rate fast enough to prevent the MCPR from becoming less than-i-06 nuring the f i

limiting power transient analyzed in Section 15.4 of the FSAR. This analysis j/

! shows that the negative reactivity rates resulting from the scram with the i i average response of all the drives as given in the specifications, provide the required protection and MCPR remains greater than +:46% The occurrence of p[;,

j' 9 %" scram times longer than those specified should be viewed as an indication of a - b g d

:S problem with the rod drives and therefore the surveillance interval N Wy.ll

'is reduced in order to prevent operation of the reactor for long periods of M -

[

] time with a potentially serious problem.  !

} The scram discharge volume is required to be OPERABLE so that it will be l available when needed to accept discharge water from the control rods during a q reactor scram and will isolate the reactor coolant system from the containment j when required.

Control rods with inoperable accumulators are declared inoperable and Spe-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  ;

j analyzed even though control rods with inoperable accumulators may still be i j slowly scrammed via reactor pressure or iaserted with normal drive water pres-i 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.

l  ;

GRAND GULF-UNIT 1 B 3/4 1-2 Mpr"M.

i~

3/4.2 POWER DISTRIBUT]ON LIMITS

BASES

h specificati.ons of this section assure that the peak cladding temper-ature followi the postulated design basis loss-of-coolant accident will not exceed the 22 'F limit specified in 10 CFR 50.46. -

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE This specification assures that the peak cladding temperature following  !

the postulated design basis loss-of-coolant accident will not exceed the 1,imit 1 specified in 10 CFR 50.46.

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

• accident is primarily a function of the average heat generation rate of all , ?bl t the rods of a fuel assembly at any axial location and is dependent only secondar- pQ ily on the rod to rod power distribution within an assembly. h peak clad tem-

~~

g , (

perature is calculated assuming a LNGR for the highest powered rod which is

.5 a equal to or less than the design LHGR corrected for densification. This LNGR Q' gd 3 times 1.02 is used in the heatup code along with the exposure dependent steady P1 i

g 3 state gap conductance and rod-to-rod local peaking factor. h Technical Speci-fication AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) is this LHGR of l.!)

gj%

_n the hinhest oowered rod divided by its local peaking factor.l The h iting value' W5

) & l 4 - t,7 U C ie e L a in Tig a e 3.2.1-1i -c

~

jd

! The daily requirement for calculating APLNGR when THERMAL POWER is greater M *'

than or equal to 25% of RATED THERMAL POWER is sufficient since power distribu-

tion shifts are very slow when there have not been significant power or control

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

complation of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER X l ensures thermal limits are met after power distribution shifts while still t allotting time for the power distribution to stablize. W requirement for  ;

0

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

! exists ensures that APLHGR will be known following a change in THERMAL POWER (

or power shape, that could place operation exceeding a thermal limit. ,

h calculational procedure used to establish the APLNGR t b 9 ; = '

l O.O.1 1 is based on a loss-of-coolant accident analysis. The analysis was per- f formed using General Electric (GE) calculational models which are consistent i with the requirements of Appendix K to 10 CFR 50. A complete discussion of each

• 1 code employed in the anslysis is presented in Reference k Differences in this Jl

! analysis compared to previous analyses can be broken down as fol'ows. Nnd 6

a. Input Channes l 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 l

%8p Ouide were recalculated using a more accurate technique.

3. Corrected guide tube thermal resistance.

l

4. Correct heat capacity of reactor internals heat nodes.

l ___

! 4sr hw pe*P aped , - _^ F7 1 # 8*2 4m"

• t ', m -M . De Qa

, , m.ee is, a.asm+

7" 4.. awr.e=-en e g & '*'-.. _

B 3/4 2-1 MMrweagNT M_

l _

GRAND GULF-UNIT 1

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

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

MAPFAC 's are generated using the same data base as the MCPR p to protect the corefFomplanttransientsotherthancoreflowincreases.

J12 MISC 86030401 - 2

POWER DISTRIBUTION LIMITS BASES _

AVERAGE PLANAR LINEAR HEAT GENERATION RATE (Continued)

b. Model Chance l

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 us'ed in the SAFE-REFLOOD pressure transfer when the pressure is increasing I was changed.

A few of the changes ~ affect the accident calculation irrespective of I 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 D5 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 DELETED N 3/4.2.2 APRM SETPOINTS e fuel cladding integrity Safety Limits of Specification 2.1 wer ased on a po distribution which would yleid the design LHGR at RATED MAL POWER. The ow biased simulated thermal power-high scram setti and flow biased simulate hermal power-upscale control rod block fun ons of the APRM i

n s.c.9 nstruments must djusted to ensure that the MCPR does t become less than Ago-* -t-M or that > IX plas strain does not occur in t graded situation. 5-O l The scram settings and to lock settings are adju d in accordance with the *'

formula in this specification en the combina n of THERMAL POWER and MFLPD @

• M"'3 indicates a peak power distribut to ens than an LHGR transient would not be increased in degraded conditions.

The daily requirement to we y the A ontrol rod block and scram g setpoint when THERMAL POWER greater than o ual to 25% of RATED THERMAL 3 POWER is sufficient since r distribution shift re very slow when there a have not been signift power or control rod changes. The tsquirement to ,

verify the APRM se nts within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the comp 1 n of a THERMAL POWER increase at least 15% of RATED THERMAL POWER ensures real limits Q are met aft power distribution shifts while still allotting ti or the s power d ibution to stabilize. The requirement to verify the AP tpoints j once r 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after initially determining MFLPD to be greater than ures that the consequences of an LHGR transient would not be increased degraded conditions.

GRAND GULF-UNIT 1 8 3/4 2-2 AgrMpagur Mr.

POWER DISTRIBUTION LIMITS Bases Table 8 3.2.1-1 -

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

a. Large Breaks 3.1 ft 2,

b. Small Breaks 0.1 f t 2, Fuel Parameters:

PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM i LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING POWER FUEL TYPE GEOMETRY (kW/ft) FACTOR RATIO sa k%l Initial Core 8 x 8 RP 13.4 1.4 K* on vp 3\,t%L bMcPR("

A more detailed listing of input of each model and its source is presented in Section II of Reference 1 and subsection 5.3.3 of the FSAR.

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

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POWER DISTRIBUTION LIMITS BASES l

_ =

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safety Limit MCPR ef 1.00, and an analysis of abnormal opera u sa,,Ml ,l tional transients. For any abnormal operating transient analysis evaluation py gme '

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. ,

4 n

To assure that the fuel cladding integrity Safety Limit is not exceeded 4 during any anticipated abnormal operational transient, the most limiting tran- j sients have been analyzed to determine which result in the largest reduction x in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant } ~l ,

temperature decrease. The limiting transient yields the largest delta J9GPC C.PR. M' l 5 When added to the Safety Limit MCPR ' 2.^', the required ti..tzr operating '

oJ limit MCPR of Specification 3.2.3 is obtained :nd F;;;nt-d in Tier: 3.2.0-1. en w ,w/

3Er The power-flow map of Figure B 3/4 2.3-1 defines the analytical basis for  % 3i,m 3g generation of the MCPR operating limits. m g. ,,g .g g % 4

.I 2 5 The evaluation of a given transient beains with the system initial

_ parameters shown in FSAR Table 15.0-27that are input to a GE-core dynamic athavior en sam

~

j. transient computer program > The code used to evaluate pressurization events l b

M4 p',.. is described in NE00-24154(3) and the program used in non-pressurization events 1%

i

. 1s described in NEDO-10802(2) The outputs of this program alon wfth the I _d l initial MCPR form the input for further analyses of the thermall limiting bundle

] h; with the single channel transient thermal hydraulic TASC code described in 4 The principal result of this evaluation is the reduction in g & NEDE-25149I4) .

+e MCPR caused by the transient.

c o

.g h The purpose of the MCPR g and MCPRp ;f Ti;;r;; 3.2.3-1 :nd 3.2.3-2 is to g.1 define u.. operatingmrw limits at other than rated core flow and power conditions. 1

'gj

, a. 6.. .,....a s u._, . . s . _ _ .6. .... ,4.ma w on 4, . 6. io m. ,,3u,

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mh WhDb i

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[e mUnI4 m me.h1 an[ ,. at At a C The'MCPR s f ,

.,f o1 are established to protect the core from inadvertent core flow increases such I g,jthatthe99.9%MCPRlimitrequirementcanbeassured.j P fr,- CThe reference core flow increase event used to establish the MCPR g is a 6

1. sp g5neutron f l flux overshoot typothesized slow flow runout to maximum, that does not result in a scram from exceedi the APRM neutron flux-high level (Table 2.2.1-1 4 I

,g item 2).gighJyhsbasis.the PRg curvesgnerated from a series of steady p*

f. T.S state a flow core thermal conditions hydraulic along calculations the steepest performed flow control line. at Thi;several

77;;pcore rd; t;plower the and 4 10"A et;;;;JW 1h ;entrel lia; 'riger; " 3/4 2.5-2). In the actual calcula-lO5%s tions a conservative highly steep generic representation of the '^'" ^^^ - " --

skM % flow control line has been used. Assumptions used in the original calculations

.//:s of this generic flow control line were consistent with a slow flow increase od/M transient duration of several minutes: (a) the plant heat balance was assumed GRAND GULF-UNIT 1 8 3/4 2-4 ANrAcMswr M.

INSERT I The maximum runout flow value is dependent on the existing setting of the core flow limiter in the Recirculation Flow Control System. Two flow rates have been considered, 102.5% core flow and 107.0% core flow (for Increased Core Flow operation).

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i l l l POWER DISTRIBUTION LIMITS BASES _

MINIltJM CRITICAL POWER RATIO (Continued) I l

i to be in equilibriu , and (b) core menon concentration was assumed to be constant. l l The generic flow control line is used to define several core power / flow states l l' _

at which to perform steady-state core themal-hydraulic evaluations. jxe.se d "JT jo:ML' s The first state fanalyzed corresponded to the maximum core power at maxi-

~

l eum core flow G46.-55 of rated) after the flow runout. Several evaluations were perfomed at this state iterating on the normalized core power distribu-tion input until the limiting bundle MCPR just exceeded the safety limit *

[lll l

, Specification (2.1.2). Next, similar calculations of core MCPR perfomance 4 '

, were determined at other power / flow conditions on the generic flow control M9

! line, assuming the same nomalized core power distribution. The result is a j%

definition of the MCPRgperformance requirement such that a flow increase event q l

) -

to maximum (h t."4) will not violate the safety limit. (The assumption of con- $_, l

stant power distribution during the runout power increase has been shown to be 3 j conservative. Increased negative reactivity feedback in the high power limiting bundle due to doppler and voids would reduce the limiting bundle relative power I

, in an actual runout.)  :

I l

i The MCPR,is established to protect the core from plant transients other I

than core flow increase including the localized rod withdrawal error event.

b Core power dependent setpoints are incorporated (incremental control rod with- a

, drawal limits) in the Rod Withdrawal Limiter (RWL) System Specification (3.3.6).  ?

l I t g

These setpoints allow greater control rod withdrawal at lower core powers where core thermal margins are large. However, the increased rod withdrawal requires

[,

I jJ higher initial MCPR's to assure the MCPR safety limit Specification (2.1.2) is

• i

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

• eents that sunoort the RWL system are presented in GESSAR II. Annendix 15B. 1 j 3 verity of other (core-wide) transients at off-r s i .

, g limited by the o setdown the e simulated thermal U

.g power-high scram trip se <on (3.2.2), the rod withdrawal error is

• 9 hing transient and estab1@T1e**, -irements ---

Mapgser j i At THERMAL POWER levels less than or equal to 25% of RATED THERMAL POWER, Jeh l the reactor will be operating at minimum recirculation pump speed and the modera en j ter void content will be very small. For all designated control rod patterns Mar E which any be employed at this point, operating plant experience indicates that 1%

l the resulting MCPR value is in excess of requ'roments by a considerable margin.

Ouring initial start-up testinji of the plant a MCPR evaluation will be made at m of RATED THERMAL POWER evelwitheinInumrecirculationpumpspeed, j The MCPR margin will thus be demonstrated such that future MCPR evaluation be-low this power level will be shown to be unnecessary. The daily requirement for calculating MCPR when THERMAL POWER is greater than or equal to.25% of RATED

THERMAL POWER is sufficient since power distribution shifts are very slow when

there have not been significant power or control rod changes. The requirement

! to calculate MCPR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER in-l crease of at least 15% of RATED THERMAL POWER ensures thermal limits are met i .

INSERT 7T (either 102.5% for Rated Core Flow operation of 107% of rated for -

Increased Core Flos operation).

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.

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

l j INSERT "J" For core power below 40% of RATED THERMAL POWER, where the EOC-RPT and the j

reactor scrams on turbine stop valve closure and turbine control valve fast  !

closure are bypassed, separate sets of MCPR limits are provided for high and t P low core flows to account for the significant sensitivity to initial core flows.

For core power above 40% of RATED THERMAL POWER, bounding power dependent MCPR limits were developed.

1 1

J12 MISC 86030401 - 3

i i

POWER 015TRIBUT10N LIMITS l BASES NINIMUM 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 80D PATTERN exists ensures that MCPR will be known following a change in THERMAL POWER or power shape, that could place 1 operation exceeding a thermal limit.

3/4.2.4 LINEAR HEAT GENERATION RATE This specifiestion assures that the Linear Heat Generation Rate (LNGR) in any rod is less than the design linear heat generation even if fuel pellet densification is 0)stulated.

, The daily requirement for calculatin0 LNGR when THERMAL POWER is greater i than er equal to 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 LHCR within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after the completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER ensures thermal limits are met after power distribution shifts while still i

a11otting time for the power distribution to stabilize. The requirement for calculating LNGR after initially determining a LIMITING CONTROL R00 PATTERN exists ensures that LNGR will be known following a change in THERMAL POWER er power shape that could place operation exceeding a thermal limit.

References:

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

l 2. R. 8. Linford, Analytical Methods of Plant Transient Evaluations for the GE BWR, February 1973 (NEDD-10802).

I 3. Qualification of the One Dimensional Core Transient Model for l Boiling Water Reactors, NE00-24154, October 1978.

! 4. TASC 01-A Computer Program for The Transient Analysis of a Single

! Channel. Technical Description, NEDE-25149, January 1980.

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GRAND GULF-UNIT 1 5 3/4 2-7 dAENbAr#TAlp, l

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f INSTRUNENTATION

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) 8ASES RECIRCULATION PUMP TRIP ACTUATION INSTRUMENTATION (Continued) feature will function are closure of the turbine stop valves and fast closure of the turbine control valves.

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

Each EOC-RPT system may be manually bypassed by use of a keyswitch which is Y administratively controlled. The manual bypasses and the automatic Operating [

Bypass at less than 40% of RATED THERMAL POWER are annunciated in the control room. ywy Q jj  ;

The EOC-RPT system response time is the time assumed in the analysis l between initiation of valve motion and complete suppression of the electric _4 arc, i.e., 190 ms, less the time allotted from start of motion of the stop valve j or turbine control valve until the sensor relay contact supplying the input to i the reactor protection system opens, i.e., 70 ms, and less the time allotted for breaker arc suppression determined by test, as correlated to manufacturer's test results, i.e., 50 ms, and plant pre-operational test results.

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

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

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

3/4.3.6 CONTROL R00 BLOCK INSTRUMENTATION l The control rod block functions are provided consistent with the require-ments of the specifications in Section 3/4.1.4, Control Rod Program Controls and Section 3/4.2 Power Distribution Limits. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

The OPERA 8ILITY of the control rod block instrumentation in OPERATIONAL CONDITION 5 is to provide diversity of rod block protection to the one rod-out

} interlock.

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

-- ,,,,.-.-w-_,.-_.,,,.__..--_.%., - _

y , -7,-_,___.,r ,_,yw__y , , _, , , , ___ , , , _

INSERT K The automatic bypass setpoint is feedwater temperature dependent due to the subcooling changes that affect the turbine first stage pressure-reactor power relationship. For RATED THERMAL POWER operation with feedwater temperature greater than or equal to 420*F. an allowable setpoint of JE 26.9% of control valve wide openThis turbine first stage setpoint is also pressure is provided for the bypass function.

applicable to operation at less than RATED THERMALThePOWER allowablewith the setpoint is correspondingly lower feedwater temperature.

reduced to 6.22.5% of control valve vide open turbine first stage pressure for RATED THERMAL POWER operation with a feedwater temperature between 370*F and 420*F.

Similarly, the reduced setpoint is applicable to operation at less that RATED THERMAL POWER with the correspondingly lower feedwater temperature.

INSTRUMENTATION l 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. y 3

% ,3.1o Alsurgon Ft.o r Montroerne ru+reuseu r4now IN $ERT* *b" v

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GRAND GULF-UNIT 1 B 3/4 3-7 AMenoMent No,

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I I

INSERT "D" RASES 3/4.3.10 KEUTRON FLUX MONITORING INSTRUMENTATION 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.

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 values. Therefore, actions taken to 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.

0 $

I J14 MISC 86020502 - 7

l I

1 3/4.4 REACTOR COOLANT SYSTEM BASES i 3/4.4.1 RECIRCULATION SYSTEM 1 asar 4. -

be 9*.4. I Operation with one reactor core coolant recirculation loo p rd ib i t:d =t i ' cr n: h:t im . . ... ,,- . . . . -...-- . . ...- - .p i noperabl aios n; eIGP Te- e

r

ti r 5= inr p r fr ;;d, : =l at:d =d d;traind t; b; en-Etd h. .

sa%w lear opemMos,-

An inoperabla 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 5 the core; thus, the requirement for shutdown of the facility with a jet pump ~'

inoperable. Jet pump failure can be detected by monitoring jet "

formance on a prescribed schedule for significant degradation.J / pump per-ecirculation -

loop flow mismatch limits are in compliance with ECCS LOCA analysis design a criteria. The limits will ensure an adequate core flow coastdown from either i recirculati'on loop following a LOCA = g

= hsur 'c'(usa entsne*h e 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 a prior to startup of an idle loop. The loop temperature must also be within f

'l l 50*F of the reactor pressure vessel coolant temperature to prevent thermal shock i*

to the recirculation pump and recirculation nozzles. Since the coolant in the 1

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 p l 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 c,onsistent with the rod pattern I

' and recirculation pump speed. The recirculation flow control system consists 1 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.

l

' The " motion inhibit" signal causes hydraulic power unit shutdown and hydraulic I 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 1011% 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 adequate to ensure that the flow control valves remain OPERABLE and not so frequent as to cause excessive wear on the system components.

.L cases whore % a0v=.M limi+, c4%at k mMaed, conhved oped. [ ,,, , ,

is pe%%ed w % one looph operad6.s .

GRAND GULF-UNIT 1 B 3/4 4-1 AMsuoMasr o.

l

\ _ _ _ _ . . _ . - - .

. l 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.

l J14 MISC 86020502 - 6

_ - _ _ _ - _\