Text

Proposed Tech Spec Changing Control Rod Scram Insertion Time Requirements to Complete Third Refueling Outage & Begin Cycle 4 Operation
	ML20023B728
Person / Time
Site:	Brunswick
Issue date:	05/02/1983
From:	; CAROLINA POWER & LIGHT CO.
To:	;
Shared Package
ML20023B727	List: ML20023B726; ML20023B728;
References
	NUDOCS 8305060180
	Download: ML20023B728 (30)
	v • d • e

- - --_.

INDEX DEFINITIONS SECTION PAGE 1.0 DEFINITIONS ACTI0N...........................................................1-1 AVERAGE PLANAR EXP0SURE..........................................

1-1 AVE RAGE P LANAR LINEAR H E AT G ENERATION RATE....................... 1-1 CHANNEL CALIBRATION..............................................

1-1 C H AN N E L C H E CK.................................................... 1 - 1 l

l CHANNEL FUNCTIONAL TEST..........................................

1-1 l

C O RE A LTE RAT ION.................................................. 1 - 2 CRITICAL POWER RATI0.............................................

1-2 DO S E E Q UIVA LENT I-131............................................ 1 -2 i

E-AVE RAGE D I S INTEGRAT ION E NERGY................................. 1-2 t

EMERGENCY CORE COOLING SYSTEM ( ECCS) RESPONSE T U!E............... 1-2 F RE Q U E NCY N0TATIO N............................................... 1 - 3 i

I D E NT I F I E D L E AKAG E............................................... 1 - 3 j

ISOLATION SYSTEM RESPONSE TDfE................................... 1-3 l

i L IMIT ING CONTROL R0D PATTE RN..................................... 1 -3 LINEAR HEAT GENERATION RATE...................................... 1-3 LOGIC SYSTEM FUNCTIONAL T E ST..................................... 1-3 MAXDIUM TOTE PEAKING FACT 0R..................................... 1-3 1

MIN DIUM CRIT ICAL P OWER RATI0..................................... 1 -4 l

l ODYN O7 TION A....................................................

1-4 i

l ODYN OPTION B.................................................... 1-4 i

I O P E RAB LE - O P E RAB I L ITY........................................... 1 - 4 OPERATIONAL CONDITION............................................

1-4 PHYSICS TESTS.................................................... 1-4 I

t i

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BRUNSWICK - UNIT I I

Amendment No.

8305060180 B30502 l

PDR ADOCK 05000325 PDR p

l

INDEX DEFINITIONS SECTION 1.0 DEFINITIONS (Continued)

PAGE PRESSURE BOUNDARY LEAKAGE........................................

1-4 P RIMARY CONTAINMENT I NTEGRITY.................................... 1 -5 l

RATED THERMAL P0WER..............................................

1-5 REACTOR PROTECTION SYSTE'11 RESPONSE TIME.......................... 1-5 KE FE RENC E LEVE L ZE R0............................................. 1 - 5 l

RE PO RTABLE O CCU R RENCE............................................ 1 -5 R0D DENSITY......................................................

1-6 l

SECONDARY CONTAINMENT INTEGRITY.................................. 1-6 SilUTDO WN MARG I N.................................................. 1 - 6 S P I RAL R E L0 AD.................................................... 1 - 6 SPIRAL UNL0AD....................................................

1-6 S T AGG E RE D T E ST B AS I S............................................ 1 - 7 THERMAL P0WER....................................................

1-7 T O TAL P E AK I NG F A CT0 R............................................. 1 - 7 UNIDENTIFIED LEAKAGE.............................................

1-7 l

F REQUENCY NOTATION, T ABLE 1.1.................................... 1-8 OPERATIONAL CONDITIONS, TABLE 1.2................................

1-9 l

i l

l l

l BRUNSWICK - UNIT 1 II Amendment No.

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3 / 4. 0 AP P L I CA B I L ITY................................................. 3 / 4 0 - 1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDOWN MARGIN..........................................

3/4 1-1 3/4.1.2 REACTIVITY AN0MALIES.....................................

3/4 1-2 3/4.1.3 CONTROL RODS Co n t ro l Ro d Op e ra bili ty..................................

3/4 1-3 Cont rol Rod !!aximum Scram Insertion Times................

3/4 1-5 Co nt rol Rod Ave rage Scram Inse rtion Time s................

3/4 1-6 Four Control Rod Group Insertion Times...................

3/4 1-7 Co n t ro l Ro d Sc rma Ac cumul a t o r s...........................

3/4 1-8

)

Control Rod Drive Coupling...............................

3/4 1-9 Co ntrol Rod Position Indication..........................

3/4 1-11 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer......................................

3/4 1-14 Rod Sequence Control Sys tem..............................

3/4 1-15 l

Ro d B l o ck Mo n i t o r........................................

3/4 1-17 3/4.1.5 STANDBY LIQUID CONTROL SYSTEM...w........................

3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT G ENERATION RATE...............

3/4 2-1 3/4.2.2 APRM SETP0INTS...........................................

3/4 2-9 3/4.2.3 MINIMUM CRIT ICAL POWER RATI0.............................

3/4 2-10 3/4.2.4 LINE AR HEAT G ENE RATION RATE..............................

3/4 2-16 l

l l

l i

l BRUNSWICK - UNIT 1 IV Amendment No.

l

DEFINITIONS MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO IMCPR) shall be the smallest CPR which exists in the core.

ODYN OPTION A ODYN OPTION A shall be analyses which refer to minimum critical power ratio limits which are determined using a transient analysis plus an analysis uncertainty penalty.

ODYN OPTION B ODYN OPTION B shall be analyses which refer to minimum critical power ratio limits determined using a transient analysis which includes a requirement for 20% scrma insertion times to reduce the analysis uncertainty penalty.

OPERABLE - OPERABILITY A system, subsystem, train, component, or device shall be OPERABLE or have OPERABILITY when it is capable of performing its specified function (s).

Implicit in this definition shall be the assumption that all necessary attendant instrumentation, controls, normal and emergency electric power sources, cooling or seal water, lubrication or other auxiliary equipment that are required for the system, subsystem, train, component, or device to perform its function (s) are also capable of performing their related support function (s).

OPERATIONAL CONDITION An OPERATIONAL CONDITION shall be any one inclusive combination of mode switch position and average reactor coolant temperature as indicated in Table 1.2.

PHYSICS TESTS PHYSICS TESTS shall be those tests performed to measure the f undamental nuclear characteristics af the reactcr core and related instrumentation and are 1) described in Section 13 of the FSAR, 2) authorized under the provisions of 10 CFR 50.59, or 3) otherwise approved by the Commission.

PRESSURE BOUNDARY LEAKAGE PRESSURE BOUNDARY LEAKAGE shall be leakade through a non-isolatable f ault in a reactor coolant system component body, pipe wall, or vessel wall.

BRUNSWICK - UNIT 1 1-4 Amendment No.

l i

DEFINITIONS PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY shall exist when:

a.

All penetrations required to be closed during accident conditions are either:

1.

Capable of being closed by an OPERABLE containment automatic isolation valve systes, or 2.

Closed by at least one manual valve, blind flange, or deactivated automatic valve secured in its closed position, except as provided in Table 3.6.3-1 of Specification, 3.6.3.1, or b.

All equipment hatches are closed and sealed.

c.

Each containment air lock is OPERABLE pursuant to Specification 3.6.1.3.

d.

The containment leakage rates are within the limits of Specification 3.6.1.2.

e.

The sealing mechanism associated with each penetration (e.g., welds, bellows or 0-rings) is OPERABLE.

RATED THERMAL POWER RATED THERMAL POWER shall be a total reactor core heat transfer rate to the reactor coolant of 2436 MWT.

REACTOR PROTECTION SYSTEM RESPONSE TIME REACTOR PROTECTION SYSTEM RESPONSE TIME shall be the time interval f rom when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is ar bitrarily set at 367 inches above the vessel zero point. This REFERENCE LEVEL ZERO is approximately mid point on the top fuel guide and is the single reference for all specifications of vessel water level.

REPORTABLE OCCURRENCE A REPORTABLE OCCURRENCE shall be any of those conditions specified in Specifications 6.8.1.8 and 6.9.1.9.

BRUNSWICK - UNIT 1 1-5 Amendment No.

z e

s DEFINITIONS ROD DENSITY s

R0D DENSITY shall be the nuaber of control rod notches insert 2d as a f raction of the total number of notches.

All rods fully inserted is equivalent to 10bt.

ROD DENSITY.

3 SECONDARY CONTAIN'IENT INTZGRITY SECONDARY CONTAINMENT INTEGRITY shall exist when:

a.

All automatic reactor building ventilation system isolation valves or dampers are OPERABLE or secured in the isolated posittun, b.

The standby gas treatment system is OPERABLE pursuant to Specificatioa 3.6.6.1.

c.,

At least one door in each access to the reactor building is closed.

d.

The sealing mechanism associated with each penetration-(e.g., welds, bellows or 0-rings) is OPERABLE.

O w-SHUTDOWN MARGIN SHUTD04N MARGIN shall be 'the amount of reactivity by which the reactor would be suberitical assuming 9 hat all control rods capable of insertion ar+1 fully inserted except for t'he analytically determined highest worth rod wht-h is assumed to be fully withdrawn, and. the reactor is in the shutdown con'dition,

's cold, 68*F, and Xenon f ree.

N SPIRAL RELOAD

^

b A SPIRAL RELOAD is the reverse of a SPIRAL UNLOAD.

Except for two diagonal fuel bundles around each of the four SRMs, the fuel ih the interioc Of the

_f l

core, synmetric to the SRMs, is loaded first.

i SPIRAL UNLOAD l

A SPIRAL UNLOAD is a core unload performed by first removing the fuet f rom the j

autermost control cells (four bundles surrounding a control blade).~ Unloading l

continues in a spiral f ashion by removing fuel f rom the outermost periphery to l

the interior of the core, symmetric about the SRMs, except for'tyo diagonal' l

fuel bundles around each of the f our SKMs.

N i

3RUNSWICK - UNIT 1 l-6 Amendment No.

v e

4.,

3 r

i DEFINITIONS STAGGERED TEST bag r

A STAUGERED TEST BASIS chall consist of:

a.'

A test schedule for n systems, subsystems, trains or other designated components obtained by dividind the specified test interval into n equal subintervals.

b.

The testing of one system, subsystem, train or' ither desidnated

.a t.omponent at the beginning of each subinterval; I

THERMAL POWER THERMAL POWER ci all be the total reactor" core heat t ransf er rate to the reactor coolant,- '

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TOTAL PEAKING FAC'10R The TOTAL PEAKING FACT 0l' (TPC) shall be the ratio of local LHGR for any specffic location"on,a fuel rod divided by the averapj LHGR associated with the fuel bundle,{ of te e same t.fpe operating; st the core average bundle power.

w.

UNIDENTIFIED LEAYAGE

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UNIDENTIFIED.iEAKAGE shall be all leakage which is not IDENTIFIED LEAKAGE.

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BRUNSWICK - UNIT 1 1-7 Amendoent No.

t-l I

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TABLE 1.1 FREQUENCY NOTATION NOTATION FREQUENCY S

At least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

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

W At least once per 7 days.

M At least once per 31 days.

Q At least once per 92 days.

SA At least once per 134 days.

A At least 01cc per 366 days.

R At least op.ce pct 13 months (550 days).

S/U Prior to each reactor startup.

N.A.

Not applicable.

BRUNSWICK - UNIT 1 1-8 Amendment No.

TABLE 1.2 OPERATIONAL CONDITIONS OPERATIONAL MODE SWITdi AVERAGE COOLANT CONDITIONS POSITIONS TEMPERATURE 1.

POWER OPERATION Run Any temperature 2.

STARTUP Startup/ Hot Standby Any temperature 3.

HOT SHUTDOWN Shutdown

> 212* F 4.

COLD SHUTDOWN Shutdown

< 212*F 5.

REFUELING

Re f uel*

j( 212* F

Reactor vessel head unbolted or removed and f uel in the vessel. ***

- See Special Test Exception 3.10.3.

- - See Special Test Exception 3.10.1.

BRUNSWICK - UNIT 1 t-9 Amendment No.

.=

i REACTIVITY CONTROL SYSTEMS 9

CONTROL ROD AVERAGE SCRAM INSERTION TIMES LIMITING CONDITIONS FOR OPERATION 3.1.3.3 The average scram insertion time of all OPERABLE control rods from the fully withdrawn position, based on de-energization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

f Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 46 0.31 36 1.05 26 1.82 6

3.37 APPLICABILIIY:

OPERATIONAL CONDITIONS 1 and 2.

ACTION:

- With the average scram insertion time exceeding any of the above limits, be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

-SURVEILLANCE REOUIREMENTS 4.1.3.3 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement l

4.1.3.2.

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6 BRUNS'JICR - UNIT l 3/4 1-6 Amendment No.

1

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REACTIVITY CONTROL SYSTEM 3 FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LIMITING CONDITION FOR OPERATION 3.1.3.4 The average scram insertion time, from the fully withdrawn position, for the three fastest control rods in each group of four control rods arranged in a two-by-two array, based on deenergization of the scram pilot valve solenoids as time zero, shall not exceed any of the following:

Position Inserted From Average Scram Inser-Fully Withdrawn tion Time (Seconds) 46 0.33 36 1.12 26 1.93 6

3.58 APPLICABILITY:

OPERATIONAL CONDITIONS 1 and 2.

l ACTION:

With the average scran insertion times of control rods exceeding the above limits, operation may coatinue and the provisions of Specification 3.0.4 are not applicable provided:

The control rods with the slower than average scram insertion times a.

are declared inoperable, b.

The requirements of Specification 3.1.3.1 are satisfied, and The Surveillance Requirements of Specification 4.1.3.2.c are c.

performed at least once per 92 days when operation is continued with three or more contt.91 rods with slow scram insertion times.

Otherwise, be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months .

SURVEILLANCE REQUIREMENTS 4.1.3.4 All control rods shall be demonstrated OPERABLE by scram time testing from the fully withdrawn position as required by Surveillance Requirement l

4.1.3.2.

BRUNSWICK - UNIT 1 3/4 1-7

\\mendment No.

~-

f 1

3/4.2 POWER DISTRIBUTION LIMITS i

3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.1 All AVERAGE PLANAR LINEAR llEAT GENERATION RATES ( APLHGRs) for each type of fuel as. a function of AVERAGE PLANAR EXPOSURE shall not exceed the j

following limits:

i' a.

During two recirculation loop operation, the limits are shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, or 3.2.1-7.

APPLICABILITY: OPERATIONAL CONDITION 1, when THER24AL POWER is greater than or equal to 257, of RATED THERMAL POWER.

ACTION:

With an APLHGR exceeding the limits of Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, j

3.2.1-4, 3.2.1-5, 3.2.1-6, or 3.2.1-7 initiate corrective action within 15 l

minutes and continue corrective action so that APLHGR is within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THER!!AL POWER to less than 25% of RATED THEPJ1AL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 4.2.1 All APLHGRs shall be verified to be equal to or less than the applicable limit determined from Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, i.

3.2.1-5, 3.2.1-6, o r 3.2.1-7:

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a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter completion of a THER:iAL POWER increase of at

.least 15% of RATED THERMAL POWER, and c.

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is

)

operating with a LIMITING CONTROL ROD PATTERN for APLHGR.

BRUNSWICK - UNIT I 1/4 2-1 Amendment No.

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POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION 3.2.2 The flow-biased APRM scram trip setpoint (S) and rod block trip set point (SRB) shall be established according to the following relationship:

S j[ (0.66W + 54%) T.

SRB I-(0.66W + 42%) T i

where:

S and S are in percent of RATED THERMAL POWER.

RB W = Loop recirculation flow in percent of rated flow, T = Lowest value of the ratio of design TPF divided by the MTPF 1

obtained for any class of fuel in the core (T f_ 1.0), and Design TPF for:

8 x 8 fuel = 2.43 8 x 8R fuel = 2.39 l

P8 x 8R fuel = 2.39 APPLICA3ILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or l

equal to 25% of RATED THERMAL POWER.

ACTION:

With S or S exceeding the allowable value, initiate corrective actt.sq within RB 15 minutes and continue corrective action so that S and S are within the.

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required limits within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to fess than 25% of R

RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

I SURVEILLANCE REQUIREMENTS 4.2.2 The MTPF for each class of fuel shall be determined, the value of T l

calculated, and the flow biased APRM trip setpoint adjusted, as required:

a.

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

b.

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

Initially and at least once per 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL R0D PATTERN for MTPF.

1 i

t

~

BRUNSWICK - UNIT 1 3/4 2-9 Amendment No.

4 2

er

+- ---

,,pr,,,,t-v-S-.y,,

y-rp.-g-ogm7ww,er eewy,em_3,._

w

,., _,.,,9..,.

c.,

.-p,

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_.,y m.

,,,,..,_,_y,..

T POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO l

LIMITING CONDITION FOR OPERATION 3.2.3.1 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core flow, shall be equal to or greater than the MCPR limit times the K shown in f

i Figure 3.2.3-1 with the following MCPR limit adjustments:

)

Beginning-of-cycle (BOC) to end-of-cycle (EOC) minus 2000 MWD /t with a.

ODYP OPTION A analyses in effect, the MCPR limits are listed below:

1.

MCPR for 8 x 8 fuel = 1.26 2.

MCPR for 8 x 8R fuel = 1.27 3.

MCPR for P8 x 8R fuel = 1.28 b.

EOC minus 2000 MWD /t to EOC with ODYN OPTION A analyses in ef fect, the MCPR limits are listed below:

~

1.

MCPR for 8 x 8 f eci = 1.37 2.

MCPR for 8 x 8R fuel = 1.38 3.

MCPR for P8 x 8R fuel = 1.41 c.

BOC to EOC minus 2000 MWD /t with ODYN OPTION B analyses in etfect, the MCPR limits are listed below:

1.

MCPR for 8 x 8 fuel = 1.21 2.

MCPR for 8 x 8R fuel = 1.25 3.

MCPR for P8 x 8R fuel = 1.25 d.

EOC minus 2000 MWD /t' to EOC with ODYN OPTION B analyses in effect, the MCPR limits are listed below:

i 1.

MCPR for 8 x 8 fuel ' l.26 2.

MCPR for 8 x 8R fuel = 1.27 3.

MCPR for P8 x 8R fuel = 1.29 APPLICABILITY:

OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or ectual to 25% RATED THERMAL POWER ACTION:

With MCPR, as a function of core flow, less than the applicable limit determined from Figure 3.2.3-1 initiate corrective action wichin 15 minutes and restore MCPR to within the applicable limit withir. 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , f

dRUNSWICK - UNIT 1 3/4 2-10 Amendment No.

4 y

,e-+m m-

,v

-g

---,,,,__oy

~-,,-y,,

,--.,y

,w,-,..,ynm.,.,

m-

,_.-.,..FT

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2 4

4 POWER DISTRIBUTION LIMITS SURVELLLA!!CE REQUIREMENTS

}

l 4.2.3.1 MCPR, as a function of core flow, shall be determined to be equal to or greater than the applicable limit determined from Figure 3.2.3-1:

a.

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

b.

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

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

L i

d 4

1 1-f a

4

)

L t

4 i

i i

A 4

i BRUNSWICK - UNIT 1 3/4 2-11 Amendment No.

= m ev--

._+w-----r

,g w+

e v w-gv e r ww-

,,.-err---

m

,em-w--re-t-

- = ~ -.

,e-,,.e.-r--.-r-----

4.---,--

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POWER DISTRI3UTION LIMITS 3/4.2.3 i INIMUM CRITICAL POWER RATIO (ODYN OPTION B)

LIMITING CONDITION FOR OPERATION 3.2.3.2 For the OPTION B !!CPR limits listed in specification 3.2.3.1 to be used, the cycle average 20% scram time ( T

) shall be less than or equal to the Option B scram time limit ( T ), where"Y" ave ""

B

8 3

follows:

n

[

T i=1 ii

, where Tave "

n N

[

i=1 i = Surveillance test number, n = Number of surveillance tests performed to date in the cycle (including BOC),

N = Number of rods tested in the ith surveillance test, and g

T = Average scram time to notch 36 for surveillance test i i

N 1/2

(,13')

(c), where:

T = p + 1.65 3

T" i=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including SOC),

th N = Number of rods ' tested in the i surveillance test 1

N1 = Number of rods tested at BOC, p = 0.834 seconds (mean value f or s tatistical scram time distribution f rom de-energi:ation of scram pilot valve solenoid to pickup on notch 36),

o = 0.059 seconds (standard deviation of the above statistical distrioution).

APPLICABILITY: OPERATIONAL CONDITION 1, when THEPJ1AL POWER is greater than or equal to 25% RATED THERMAL POWER.

BP.UNSWICK - UNIT 1 3/4 2-12 Amendment No.

POWER DISTRIBUTION LIMITS LIMITING CONDITIONS FOR OPERATION (Continued)

ACTION:

Within twelve hours af ter determining that T is greater than T ave B'

operating limit MCPRs shall be either:

Adjusted for each f uel type such that the operating limit MCPR a.

is the maximum of the non pressurization transient MCPR operating limit (f rom Table 3.2.3.2-1) or the adjusted pressurization transient MCPR operating limits, where the adjustment is made by:

T T

T

-T pti n A ptionB) adjusted "

option B +

B where: T =1.05 seconds, control rod average scram insertion A

time limit to notch 36 per Specification 3.1.3.3, option A = Determined f rom Table 3.2.3.2-1, M R MCPR

= Determined from Table 3.2.3.2-1, or, option B b.

The OPTION 1 MCPR limits listed in Specification 3.2.3.1.

i f

SURVEILLANCE REOUIREMENTS 4.2.3.2 The values of T and T shall be determined and compared each time a scram time test is periofmed. khe raz;uirement for the f requency of scram time testing shall be identical to Spectiication 4.1.3.2.

BRUNSUICK - UNIT 1 3/4 2-13 Amendment No.

TABLE 3.2.3.2-1 ri TRANSIENT OPERATING. lit!IT MCPR VALIJES 5

TRANSIENT FilEL TYPE 8x8 8x8R P8x8R U

w NONPRESSURI7.ATION TRANSIENTS BOC + EOC 1.21 1.25 1.25 TURRINE TRIP / LOAD REJECT WITIIOG"T WPASS MCPR lICPR

!!CPR

. MCPR ItCPR A

g g

g 4

B w

BOC + EOC - 2000 1.26 1.08 1.27 1.08 1.28-1.09 EOC - 2000 + EOC 1.37 1.25 1.38 1.26 1.41 1.29 FEEDNATER CONTROL FAILURE MCPR MCPR

!!CPR MCPR MCPR MCPR A

B A

g A

S BOC + EOC - 2000 1.21 1.15 1.22 1.16 1.23 1.17 g

EOC - 2000 + EOC 1.33 1.26 1.34 1.27 1.36 1.29 0

8-en

.C

1.11 f.

1.3 UNAC CEPTAI iLE OPfRATla l s

\\

1.2 N

AUTOM iTIC f LOW CO JTROL NNN N

N

/

ND N

N NNNN N

nr NNNN N

5 Z

NNDkN u

IW llAL FI ' M CGJTROL SCnP TUll: SEriOINT

ALIBRiTTION

/

\\

POS I T IGJl :) SUC6 THAT

\\

FLOW: LAX = 102.5':

%s N

N s

=

107.01

=

112.07

=

117.0';

b M

M A

i,0 lu 50 f;ll 70 30 90 100 i

N

}r-j CORE FLOW (O K FACTOR g

FIGURE 3.2.3-1

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kw/f t for 8 X 8, 8 X 8R, and P8 X 8R fuel assemblies.

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

ACTION:

With the LHGR of any fuel rod exceeding the above limit, initiate corrective l

action within 15 minutes and continue corrective action so that the LHGR is within the limit within 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months , or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REQUIREMENTS 4.2.4 LHGR shall be determined to be equal to or less than the limit:

a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months af ter completion of a THERMAL POWER increase of at least 15% of RATED THER:fAL POWER, and c.

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

l BRUNSWICK - UNIT 1 3/4 2-16 Amendment No.

TABLE 3.3.4-2 4

CONTROL ROD UITHDRAWAL BLOCK INSTRUtIENTATION SETPOINTS kh TRIP FUNCTION AND INSTRUMENT nut 181:R TRIP SETPOINT ALLOWABLE VALUE N

p, 1.

APRM (CSI-APR!!-Cil. A,B,C,D,E,F) a.

Upscale (Flow Biased)

< (0.66W + 42%)

T*

< (0.66W + 42%)

T*

lfrPF MTPF h

b.

Inoperative NA NA H

c.

Downscale

> 3/125 of f ull scale

> 3/125 of full scale H

d.

Upscale (Fixed)

< 12% of RATED tiler 11AL POWER

< 12% of RATED TilERllAL POWER 2.

ROD llLOCK MONITOR (CS I-RBM-Cl!. A,B) a.

Upscale

< (0.66W + 41%)

T*

< (0.66W + 41%)

T*

MfPF MTPF b.

Inoperative NA NA c.

Downscale

> 3/125 of full scale

> 3/125 of f ull scale 3.

SOURCE RANGE t10NITORS (C51-Slui-K600A,B,C,D) a.

Detector not f ull in NA

^

5 5

R b.

Upscale

< 1 x 10 cps

< 1 x 10 cps c.

InoperatEve NA NA d.

Downscale

> 3 cps

> 3 cps w

4.

INTERMED[ ATE RANCE !!ONITORS ( C51-I RM-K601 A, B, C,D, E, F,G,il) a.

Detector not full in NA NA b.

Upscale

< 108/125 of f ull scale

< 108/125 of full scale c.

Inoperative NA NA d.

Downscale

> 3/125 of f ull scale

> 3/125 of full scale 5.

SCRAP! DISCllARCE VOLUME (Cll-LS11-N013E) l a.

Water Level Iligli

< 73 gallons

< 73 gallons l

k

T=2.4 3 for 8x8 f uel i

T=2.31 for 8x8R fuel T=2.N for P8x8R inel

s et

.O

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding temperature following the postulated design basis loss-of-coolant accident will not exceed the 2200 F limit specified in the Final Acceptance Criteria (FAC) issued in June 1971 considering the postulated ef fects of fuel pellet densification.

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 limit specified in 10 CFR 50, Appendix K.

The peak cladding temperature (PCT) following a postulated loss-of-coolant accident is primarily a function of the average heat generation rate of all the rods of a fuel assembly at any axial location and is dependent only secondarily on the rod-to-rod power distribution within an assembly. The peak clad temperature is calculated assuming the LHGR foe the highest powered rod which is equal to or less than the design LHGR corrected for densification.

This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor.

The Technical Specification APLHGR is this LHGR of the highest powered rod divided by its local peaking factor. The limiting value for APLHGR is shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, or 3.2.1-7.

l The calculational procedure used to establish the APLHGR shown on Figures j

3. 2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, and 3.2.1-7 is based on a loss-of-coolant accident analysis. The analysis was performed using General Electric (GE) calculational models which are consistent with the requirements of Appendix K to 10 CFR 50.

A complete discussion of each code employed in the analysis is presented in Reference 1.

Dif ferences in this analysis compared to previous analyses performed with Reference 1 are:

(1)

The analysis assumes a fuel assembly planar power consistent with 102% of the MAPLHGR shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, 3.2.1-6, and 3.2.1-7 (2) Fission product decay is computed assuming an energy release l

rate of 200 MEV/ Fission; (3) Pool boiling is assumed af ter nucleate boiling is lost during the flow stagnation period; (4) The ef fects of core spray entrainment and countercurrent flow limitation as described in Reference 2, are included in the reflooding calculations.

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.

BRUNSWICK - UNIT 1 B 3/4 2-1 Amendment No.

X-

_y LR

~

POWER DISTRIBUTION LIMITS BASES 3/4.2.2 APRM SETPOINTS The fuel cladding integrity Safety Limits of Specification 2.1 were based on a TOTAL PEAKING FACTOR of 2.43 for 8 x 8 fuel, 2.39 for.8 x 8R fuel, and 2.39 for P8 x BR fuel. The scram setting and rod block functions of the APRM instruments must be adjusted to ensure that the MCPR does not become less than 1.0 in the degraded situation.

The scram settings and rod block settings are adjusted in accordance with the formula in this specification when the combination of THERMAL POWER and peak flux indicates a TOTAL PEAKING FACTOR greater than 2.43 for 8 x 8 fuel, 2.39 for 8 x 83 fuel, and 2.39 for P8 x 8R fuel. This adjustment nay be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the ficw referenced APRM high flux scram curve by the reciprocal of the APRM gain change. The method used to determine the design TPF shall be consistent with the method used to determine the MTPF.

3/4.2.3 MINIME4 CRITICAL POWER RATIO The required operating limit MCPR's at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Saf((y Limit MCPR of 1.07, and an analysis of abnormal operational transients For any abnormal operating transient analysis evaluation with the initial condition of tne 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 as given in Specification 2.2.1.

To assure that the fuel cladding integrity Safety Limit is net exceeded during any anticipated abnormal operational transient, the most limiting transients have been analyzed to determine which result in the largest reduction in CRITICAL POWER RATIO (CPR). The type of transients evaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease.

The limiting transient which determines the required steady state MCPR limit is the turbine trip with failure of the turbine bypass.

This transient yields the largest MCPR. When added to the Safety Limit MCPR of 1.07, the required minimum operating limit MCPR of Specification 3.2.3 is obtained. Prior to analysis of abnormal operational transients, an initial fuel bundle hCPR was determined. This parameter is based on the bundle flow calculated by a GE multichannel ggyady state flow distribution model as described in Section 4.4 of NED0-20360 and en core parameters shown in Reference 3, response to Items 2 and 9.

BRUNSWICK - UNIT 1 B 3/4 2-3 Amendment No.

ML20023B728

Text

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