Text

Amend 56 to License DPR-71,changing Tech Specs to Establish Revised Operating Limits for Fuel Cycle 4 & Changing Control Rod Scram Insertion Times
	ML20072L090
Person / Time
Site:	Brunswick
Issue date:	06/28/1983
From:	Vassallo D; Office of Nuclear Reactor Regulation
To:	; Carolina Power & Light Co
Shared Package
ML20072L093	List: ML20072L090; ML20072L100;
References
	DPR-71-A-056 NUDOCS 8307130021
	Download: ML20072L090 (33)
	v • d • e

_ _ _ _ _ _ _ _ _ _ _ _ _ _

UNITED STATES

.:s '-

'4 NUCLEAR REGULATORY COMMISSION j

j WASHINGTON, D. C. 20666 c

%.... /

CAROLINA POWER & LIGHT COMPANY DOCKET N0. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.56 License No. DPR-71 1.

The Nuclear Regulatory Comission (the Comission) has found that:

A.

The application for amendment by Carolina Power & Light Company (the licensee) dated May 2,1983 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; s

B.

The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission;

{

C.

There is reasonable assurance (1) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Comission's regulations; D.

The issuance of this amendment will not be iniinical to the comon defense and security or to the health and safety of the public; and E.

The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.

2.

Accordingly, the license is amended by changes to the Technical Spec-ifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-71 is hereby amended to read as follows:

(2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 56, are hereby incorporated in the licenge.

The licensee shall operate the facility in accordance with the Technical Specifications.

8307130021 830628 PDR ADDCK 05000325 p

PDR

-2 3.

This license amendment is effective as of the date of issuance.

FOR THE NU EAR REGULATORY COMMISSION A

Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: June 28,1983 ad

ATTACHMENT TO LICInSE AMENDMENT NO. 56 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 i

Replace the following pages of the Technical Specifications contained in Appendix A of the above-indicated license with the attached pages.

The changed area of the revised page is reflected by a marginal line.

Remove Insert (or delete appropriate side)

I I

II II IV IV l-4 thru l-8 1-4 thru ?-9 3/4 1-6 3/4 1-6 3/4 1-7 3/4 1-7 3/4 2-1 thru 3/4 2-11 3/4 2-1 thru 3/4 2-16 3/4 3-42 3/4 3-42 B 3/4 2-1 B 3/4 2-1 B 3/4 2-3 B 3/4 2-3

r-INDEX JDEFINITIONS SECTION PAGE 1.0 DEFINITIONS ACT10N...........................................................1-1 AVE RAG E P LANAR EXP0 SU RE.......................................... 1-1 AVE RAGE PLANAR LINEAR HEAT GENERATION RATE....................... 1-1 CHANNEL CALIBRATION.............................................. 1-1 C H ANN E L C H E CK.................................................... 1 - 1 CHANN EL FUN CT IONAL TE ST.......................................... 1 - 1 CORE ALTERATION.................................................. 1-2 CRITICAL POWER RATI0............................................. 1-2 DO S E EQ U IVA LENT I-131............................................ 1-2 E-AVE RAG E D I S I NTEG RAT ION E N E RGY................................. 1 -2 EMERGENCY CORE COOLING SYSTEM ( ECCS) RESPONSE TIME............... 1-2 FREQU ENCY N0TATIO N............................................... 1-3 IDENTIFIED LEAKAGE............................................... 1-3 ISOLATION SYSTEM RESPONSE TIME................................... 1-3 L IMIT I NG CONT RO L RO D PATT E RN..................................... 1 - 3 LINEAR HEAT GENERATION RATE...................................... 1-3 1

LOG IC S YSTEM FUNCTIONAL TE ST..................................... 1-3 MAXIMUM TOTAL PE AKING FACT 0R..................................... 1-3 MINIMUM CRITICAL POWER RATI0..................................... 1-4 ODYN OPTION 1....................................................

1-4 O DY N O P T I O N B.................................................... 1 - 4 OPE RABLE - O PE RAB I L ITY........................................... 1 -4 O P E RAT ION AL C0 ND IT ION............................................ 1 -4 PHYSICS TESTS.................................................... 1-4 l

e l

BRUNSWICK - UNIT 1 1

Amendment No. 56 y_

I INDEX DEFINITIONS SECTION 1.0 DEFINITIONS (Continued)

PAGE P RE S SU RE BOUNDARY LE AKAG E........................................ 1-4 PRIMARY CONTAINMENT I NTEGRITY.................................... 1-5 l

RATED TilERMAL P0WER.............................................. 1-5 RE ACTOR PROTECTION SYSTEM RESPONSE TIME.......................... 1-5 REFERENCE LEVEL ZER0............................................. 1-5 l

REPORTABLE OCCURRENCE............................................ 1-5 ROD DENSITY...................................................... 1-6 l

S ECONDARY CGNTAINMENT I NTEGRITY.................................. 1 -6 SilUTDOWN MARGIN.................................................. 1-6 SPIRAL REL0AD.................................................... 1-6 SPIRAL UNL0AD.................................................... 1-6 S T AGG E RE D T E ST B AS I S............................................. 1 - 7 Tile RMAL P 0 WE R.................................................... 1 - 7 TOTAL P E AK ING F ACT0 R............................................. 1 - 7 UNIDENTIFIED LEAKAGE............................................. 1-7 F REQUENCY NOTATION, T ABLE 1.1.................................... 1-8 OPERATIONAL CONDITIONS, TABLE 1. 2................................ 1-9 l

t BRUNSRICK - UNIT I II Amendment No. 57,56

INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REOUIREMENTS SECTION PAGE 3/4.0 APPLICABILITY................................................. 3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS 3/4.1.1 SHUTDORN MARGIN..........................................

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

3/4 1-2 3/4.1.3 CONTROL RODS Co n t rol Rod Op e ra bili t y..................................

3/4 1-3 Control Rod Maximum Scram Insertion Times................

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

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

3/4 1-7 Co n t rol Ro d Sc ram Ac cumul a t o rs...........................

3/4 1-8 Control Rod Drive Coupling...............................

3/4 1-9 Cont rol Rod Position Indication..........................

3/4 1-11 3/4.1.4 CONTROL R0D PROGRAM CONTROLS Ro d Wo r t h Mi n imi z e r......................................

3/4 1-14 Rod Sequence Control System..............................

3/4 1-15 Ro d B l o c k Mo n i t o r........................................

3/4 1-17 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 G ENERATION RATE...............

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

3/4 2-9 3/4.2.3 MINIMUM CRITICAL POWER RATI0.............................

3/4 2-10 3/4.2.4 LINEAR HEAT GENE RATION RATE..............................

3/4 2-16 l

l BRUNSWICK - UNIT 1 IV Amendment No. 23, 27, 56

I DEFINITIONS MINIMUM CRITICAL POWER RATIO The MINIMUM CRITICAL POWER RATIO (MCPR) 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 pa-forming 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.

PilYSICS TESTS PilYSICS TESTS shall be those tests performed to measure the fundamental nuclear characteristics of the reactor 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. 56

+

l

_ DEFINITIONS PRIMARY CONTAINMENT INTEGRITY PRIMARY CONTAINMENT INTEGRITY shall exist when:

a.

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

i 1.

Capable of being closed by an OPERABLE containment automatic isolation valve system, or 1

1 2.'

Closed by at least one manual valve, blind flange, or l

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.

I c.

Each containment air lock is OPERABLE pursuant to Specification j

3.6.1.3.

d.-

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

The sealing mechanism associated with each penetration (e.g., welds, e.

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 tima interval f rom when the monitored parameter exceeds its trip setpoint at the channel sensor until de-energization of the scram pilot valve solenoids.

4 REFERENCE LEVEL ZERO The REFERENCE LEVEL ZERO point is arbitrarily 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 i

. 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 5-Amendment No. 52, 56 3

m

DEFINITIONS ROD DENSITY ROD DENSITY shall be the number of control rod notches inserted as a f raction of the total number of notches. All rods fully inserted is equivalent to 1007.

ROD DENSITY.

SECONDARY CONTAINMENT INTEGRITY SECONDARY CONTAINMENT INTEGRITY shall exist when:

a.

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

The standby gas treatment system is OPERABLE pursuant to Specification 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.

SHUTDOWN MARGIN SHUTDOWN MARGIN shall be the amount of reactivity by which the reactor would be subcritical assuming that all control rods capable of insertion are fully inserted except for the analytically determined highest worth rod which is assumed to be fully withdrawn, and the reactor is in the shutdown condition, cold, 68"F, and Xenon f ree.

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

Except f or two diagonal fuel bundles around each of the four SRMs, the fuel in the interior of the core, spumetric to the SRMs, is loaded first.

SPIRAL UNLOAD A SPIRAL UNLOAD is a core unload performed by first removing the f uel f roa the outermost control cells (four bundles surrounding a control blade). Unioading continues in a spiral f ashisn by removing fuel f rom the outermost periphery to the interior of thu core, symmetric about the SRMs, except for two diagonal fuel bundles around each of the four SRMs, 3RUNSWICK_- UNIT 1 1-6 Amendment No. /S2, 47, 52, 56

DEFINITIONS STAGGERED TEST BASIS A STAGGERED TEST BASIS shall 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 or.her designated component at the beginning of each subinterval.

THERMAL POWER THER:!AL POWER shall be the total reactor core heat transfer rate to the reactor coolant.

TOTAL PEAKING FACTOR The TOTAL PEAKING FACTOR (TPF) shall be the ratio of local LHGR for any specific location on a fuel rod divided by the average LHGR associated with the fuel bundles of the same type operating at the core average bundle power.

UNIDENTIFIED LEAKAGE UNIDENTIFIED LEAKAGE shall be all leakage uhich is not IDENTIFIED LEAKAGE.

22, 56 BRUNSWICK - UNIT 1 1-7 Amendment No.

1 1

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 once per 366 days.

R At least once per 18 months (550 days).

S/U Prior to each reactor startup.

N.A.

Not applicable.

BRUNSWICK - UNIT 1 1-8 Amendment No. 56

TABLE 1.2 OPERATIO!1AL CONDITIONS OPERATIONAL MODE SWITCil AVERAGE COOLANT CONDITIONS POSITIONS TEMPERATURE 1.

POWER OPERATION Run Any temperature 2.

STARTUP Startup/ilot Standby Any temperature 3.

Il0T SilUTDOWN Shutdown

> 212'F 4.

COLD SilUTDOWN Shutdown

< 212*F 5.

REFUELING

Re f uel* *

< 212*F

Reactor vessel head unbolted o r removed and fuel in the vessel.***

- See Special Test Exception'3.10.3.

- - See Special Test Exception 3.10.1.

BRUNSWICK - UNIT 1 t_9 Amendment No. 56

REACTIVITY CONTROL SYSTE!!S CONTROL ROD AVERAGE SCRAM INSERTION TIMES LIMITING CONDITIONS FOR OPERATION 3.1.3.3 The average scram insertion time of all OPERABLE control rods f rom the fully withdrawn position, baied on de-energization of the scram pilot valve solenoids as time zerc, shall not exceed any of the following:

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

3.37 APPLICABILITY: OPERATIONAL CONDITIONS 1 and 2.

ACTION:

With the average scram insertion time exceeding any of the above limits, be in at least il0T SilUTDOWN 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.

!)RUNSWICK - UNIT I 3/4 1-6 Amendment :o. 56

(

t I

REACTIVITY CONTROL SYSTE:!3 I

FOUR CONTROL ROD GROUP SCRAM INSERTION TIMES LI:!ITING 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 I

APPLICABILITY:

OPERATIONAL CONDITIONS I and 2.

l ACTION:

With the average scram insertica ti cs of control etds exceeding the above limits, operation may continue 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 control rods with slow scram insertion times.

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

SURVEILLANCE REOUIREMENTS 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.

BRUNSWICR - UNIT 1 3/4 1-7

\\mendment No. 56

3/4.2 POWER DISTRIBUTION LIMITS 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE t

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 following limits:

a.

During two recirculation loop operation, the limits are shown in l

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.

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

2 ACTION:

4 I

With an APLilGR 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 THERIIAL POWER to less than 25% of RATED TilEPJ1AL POWER within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLA! CE REQUIRE!!ENTS 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, 3.2.1-5, 3.2.1-6, or 3.2.1-7:

l 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 !AL POWER increase of at Icast 15% of RATED TilERMAL 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.

a A

BRUNSWICK - UNIT 1 3/4.2-1 Amendment No. 35, 56

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1 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 (SltB) shall be established according to the following relationship:

S j( (0.66W + 54%) T SRB I-(0.66W + 42%) T 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 obtained for any class of fuel in the core (T j[ 1.0), and Design TPF for:

8 x 8 fuel = 2.43 8 x 8R fuel = 2.39 P8 x 8R' fuel = 2.39 APPLICASILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 25% of RATED THERMAL POWER.

ACTION:

With S or S exceeding the allowable value, initiate corrective action within 15minutesa!dcontinuecorrectiveactionsothatSandS are within the R

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

SURVEILLANCE REQUIREMENTS -4.2.2 The MTPF for each class of fuel shall be determined, the value of T 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 , 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 MIPF.

l

' BRUNSWICK - UNIT 1 3/4 2-9 Anendment No. /J,/$,2$,56 l

i

POWER DISTRIBUTION LUitTS j

3/4.2.3 MINIMUM CRITICAL POWER RATIO 4

' LIMITING CONDITION FOR OPERATION 4

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

Figure 3.2.3-1 with the following MCPR limit adjustments:

Beginning-of-cycle (B0C) L to end-of-cycle (E0C) minus ' 2000 MWD /t with a.

ODYN 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 f uel = 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:

r 1.

MCPR for 8 x 8 fuel = 1.37

-2.

MCPR for-8 x 8R fuel = 1.38 3.

MCPR for P8 x 8R fuel = 1.41 4

c.

BOC to E0C minus 2000 MWD /t with ODYN OPTION B analyses in effect, 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 ECC with ODYN OPTION B analyses in ef fect, 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 = 1.29 APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or ' equal to 25% RATED THERMAL POWER ACTION:

With MCPR, as a fun _ tion of core flow, less than the applicable limit i

determined from Figure 3.2.3-1 initiate corrective action within 15 minutes

~

and restore MCPR to within the applicable 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 .

BRUNSWICK - UNIT 1

'3/4 2 Amendaent No. 23, 56

POWER DISTRIBUTION LIMITS SURVEILLANCE REQUIREMENTS 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 , 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.

BRUNSWICK - UNIT 1 3/4 2-11 Amendment No. 23, 29, 56

POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO (ODYN OPTION B)

LIMITING CONDITION FOR OPERATION 3.2.3.2 For the OPTION B MCPR 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" and T are determined as B

ave B

follows:

n I

NT i=1 gg

, where r,y, =

g i

{

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 t

surveillance test, and t = Average scram time to notch 36 for surveillance test i N

1/2 I

= p + 1. 6 5 ( n

)

(a), where:

T B

N

[

i=1 i = Surveillance test number n = Number of surveillance tests performed to date in the cycle (including B0C),

th Ng = Number of rods ' tested in the'i surveillance test N1 = Number of rods tested at BOC, p = 0.834 seconds (mean value for statistical 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 distribution).

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

BRUNSWICK - UNIT 1 3/4 2 Amendment -No. 56 1

-e 9

-e

POWER DISTRIBUTION LIMITS

)

LIMITING CONDITIONS FOR OPERATION (Continued) q ACTION:

- Within twelve hours af ter determining that T is greater than T,

the operating limit MCPRs shall be either:

Adjusted for each fuel 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:

ave - B

~"

adjusted " "

option B +

T Ption A option B A ~ B where: T =1.05 seconds, control rod average scram insertion time limit to notch 36 per Specification 3.1.3.3, MCPR

= Determined f rom Table 3.2.3.2-1, option A option B = Determined from Table 3.2.3.2-1, or, MCPR b.

The OPTION A MCPR limits listed in Specification 3.2.3.1.

SURVEILLANCE REOUIREMENTS 4.2.3.2 The values of T and T shall be determined and compared each time a scram time test is perioImed.

She requirement for the f requency.of scran.

' time testing shall be identical to Specification 4.1.3.2.

l t

B

{

I' E.

3/4 2-13 Amendment No. 56 BRUNSWICK - UNIT'1 l:

. ~

.E TABLE 3.2.3.2-l'

.E iE TRANSIENT OPERATING LIf!IT MCPR VALilES 5n.

'E

' TRANSIENT FIJEL TYPE 8x8 8x8R P8x8R' E

H NONPRESSURI7.ATION TRANSIENTS BOC.+ EOC 1.21 1.25 1.25 TURBINE: TRIP / LOAD REJECT WITF10UT BYPASS MCPR ItCPR i!CPR itCPR

!!CPit itCPR g

g g

B w

BOC + EOC.- 20001 1.26 1.08 1.27 1.08 1.28 1.09 L"

EOC - 2000 + EOC 1.37 1.25 1.38 1.26 1.41 1.29 I'

I l

FEEDWATER CONTROL FAILURE' MCPR MCPR MCPR ffCPR MCPR MCPR A

g A

g g

g BOC + EOC - 2000 l.21 1.15 1.22 1.16 1.23 1.17 EOC - 2000 + EOC 1.33 1.26 1.34 1.27 1.36 1.29

.g' E

.if

l i'

1

\\i oit i

LO R

I

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T 0

O N\\

9 C

W O

L f

\\s C

I T

1 s

M Ns n

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/

N 0

N 7

1 EL l

O 3

AG

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(

R W

O 2

TI PT O

T 3

EM L

C C

0 F

A E

l Cf N\\A 1 0 1 6

F R

AP N

5000 E

U O

J O

R R

g G

L I

2727 O

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\\

M RB

= =

l l

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0 N

h TO TW 5

LTAO ONI L l

RI TF

\\ \\Q J Pl GTC i

CEU SS W

OE) o N

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TO L

f APT U3 I

1 \\

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0 8

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1 U

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afM " "

Iu U 3.4 5, E

[

f!l

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 K 8, 8 X 8R, and P8 X 8R fuel assemblies.

APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or 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 REOUIREMENTS 4.2.4 LHGR shall be deter.11ned to be equal to or less than the limit:

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 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 on a LIMITING CONTROL ROD PATTERN for LHGR.

l I

l l

BRUNSWICK - UNIT'1 3/4 2-16 Amendment No. 56 l

~,

TABLE 3.3.4-2 CONTROL ROD UITrIDRAWAL BLOCK INSTRUI!ENTATION SETPOINTS

?

N TRIP FUNCTION AND INSTRUMENT NUf1BER TRIP SETPOINT ALLOWABLE VALUE

.5 Q

1.

APRM (C51-APR!l-CII. A,B,C,D,E,F) a.

Upscale (Flow Biased)

< (0.66W + 42%)

T*

< (0.66W + 42%)

T*

MrPF KrPF qh' b.

Inoperative NA NA H

c.

Downecale

> 3/125 of full scale

> 3/125 of full scale H

d..

Upscale'(Fixed)

[ 12% of RATED tiler!!AL POWER

[ 12% of RATED tiler!!AL POWER 2.

ROD BLOCK MONITOR (C51-RBM-Cll. A,B) a.

Upscale

< (0.66W + 41%)

T*

< (0.66W + 41%)

T*

KfPF MTPF

,b.

Inoperative NA NA

> 3/125 of full scale

> 3/125 of full scale c.

Downscale 3.

SOURCE RANGE !!ONITORS (C51-SRt!-K600A,B,C,D) a.

Detector not f ull in NA

^

5 5

R b.

' Upscale

< 1 x 10 cps

< 1 x 10 cps e

c.

Inoperative NA NA d.

Downscale

> 3 cps

> 3 cps 4.

INTERMED[ ATE RANGE ilONITORS ( C51-I RM-K601 A, B,C,D, E, F,G,II) a.

Detector not full in NA NA b.

Upscale

< 108/125 of full scale

< 108/125 of full scale c.

Inoperative NA Na d.

Downscale

> 3/125 of full scale

> 3/125 of full scale 5.

SCRAM DLSCilARGE VOLUME (Cll-LSil-N013E) l a.

Water Level liigli 1 73 gallons

< 73 gallons

T=2.4 3 for 8x8 f uel T=2.39 for 8x8R fuel 8

T=2.39 for P8x8R fuel S

e o

i

3/4.2 POWER DISTRIBITTION LIMITS BASES The specifications of this section assure that the peak cladding temperature followind 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-ef-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 for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

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

The Technical 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 o n 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 f uel 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 after 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. 23. 29, 56 l

m

'M

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 8R 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 8R fuel, and 2.39 for PS x 8R fuel.

This adjustment may be accomplished by increasing the APRM gain and thus reducing the slope and intercept point of the flow 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(({ 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 Saf ety 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 not 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 l'argest 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 MCPR was i

determined. This parameter is based on the bundle flow calculated by a GE multichannel ggyndy state flow distribution model as described in Section 4.4 of NEDO-20360 and on core parameters shown in Reference 3, response to i

Items 2 and 9.

i 4

l BRUNSWICK - UNIT 1 8 3/4 2-3 Amendment No. 23, 29, 56 i

l l

l

. _ _-

ML20072L090

Text

Attachment:

= =

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