Text

Amend 29 to License DPR-71,changing Tech Specs to Establish Revised Safety & Operating Limits for Cycle 3 Operation & Safety Relief Valve Pressure Settings to Provide Min Nominal Lift Setting Differential of 20 Psi
	ML19320B687
Person / Time
Site:	Brunswick
Issue date:	07/01/1980
From:	Ippolito T; Office of Nuclear Reactor Regulation
To:	;
Shared Package
ML19320B688	List: ML19320B687; ML19320B696;
References
	NUDOCS 8007140614
	Download: ML19320B687 (23)
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UNITED STATES g

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NUCLEAR REGULATORY COMMISSION 5

E WASHINGTON, D. C. 20555

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CAROLINA POWER & LIGHT COP 9ANY DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO. 1 AENDMENT TO FACILITY OPERATING LICENSE Amendment No. 29 License No. DPR-71 1.

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

A. The applications for ameiidment by Carolina Power & Light Company (the licensee) dated May 23, May 30, as supplemented June 4, and June 25,1980 comply with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Commis-sion's rules and regulations set forth in 10 CFR Chapter I; B.

The facility will operate in confonnity 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 Commission's regulations; D.

The issuance of this amendment will not be inimical to the common 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-l ifications as indicated in the attachment to this license amendment l

- and paragraph 2.C.(2) of Facility Operating Licente No. DPR-71 is hereby amended to read as follows:

(2) Technical ~ Specifications The Technical Specifications centained in Appendices A and

)

'B, as revised through Amendment No. 29, are hereby incorporated j

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

800d40 g M

2-3.

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

FOR THE NUCLEAR REGULATORY COMMISSION

-^3 Thomas 4.C Ippolito, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: July 1, 1980 o

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ATTACHMENT TO LICENSE AMENDMENT NO. 29 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 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 III/IV III/IV V/VI v/VI 3/42-1/2 3/4 2-1/2 3/4 2-5/6 3/4 2-5/6 3/4 2-7/8 3/4 2-7/8 3/4 2-9/10 3/4 2-9/10 3/4 2-11 3/4 3-41/42 3/4 3-41/42 B3/4 2-1/2 83/4 2-1/2 B3/4 2-3/4 B3/4 2-3/4 3/4 4-4 3/4 4-4 8

INDEX SAFETY LIMITS AND LIMITING SAFETY SYSTEM SETTINGS SECTION PAGE 2.1 SAFETY LIMITS Themal Power (Low Pressure or Low Flow)...................

~ 2-1 Thermal Power (High Pressure and High Flow)...............

2-1 Reactor Coolant System Pressure...........................

2-1 Reactor Vessel Water Leve1................................

2-2 2.2 LIMITING SAFETY SYSTEM SETTINGS Reactor Protection System Instrumentation Setpoints.......

2-3 f

's BASES 2.1 SAFETY LIMITS Thermal Power (Low Pressure or Low Flow)..................

B 2-1 Themal Power (High Pressure and High Flow)............... B 2-2 Reactor Coolant System Pressure...........................

B 2-8 Reactor Vessel Water Level................................

B 2-8 2.2 Limiting Safety System Settings e

Reactor Protection System Instrumentation Setpoints.......

B 2-9 k

I, BRUNSWICK - UNIT 1 III e

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f INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE I

3/4.0 APPLICABILITY............................................. ~3/4 0-1 3/4.1 REACTIVITY CONTROL SYSTEMS

[

3/4.1.1 S HUTD OWN MARG I N........................................ 3/4 1-1 3/4.1.2 R EACTIV ITY AN 0 MAL I ES................................... 3/4 1-2 3/4.1.3 CONTROL RODS Control Rod Operability................................

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

3/4 1-5 Control Rod Average Scram Insertion Times.............. 3/4 1-6 Four Control Rod Group Insertion Times................. 3/4 1-7 Control Rod Scram Accumulators.........................

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

3/4 1-9

.__ Control Rod Position Indication...,.................... 3/4 1-11 3/4.1.4 CONTROL ROD PROGRAM CONTROLS Rod Worth Minimizer....................................

3/4 1-14 Rod Sequ ence Control System............................ 3/4 1-15

/

Rod B l oc k Mo n i to r...................................... 3/4 1-17 3/4.1/5 STANDBY LIQUID CONTROL SYSTEM.......................... ' 3/4 1-18 3/4.2 POWER DISTRIBUTION LIMITS AVERAGE PLANAR LINEAR HEAT GENERATION RATE.............,.,3/4 2-1 A P RM S ET P 0 l NTS......................................... 3/4 2-8 l

MINIMUM CRITICAL POWER RATI0...........................

3/4 2-9 l

LIN EAR HEAT GEN ERATION RATE............................ 3/4 2-11 l

I BRUNS, WICK - UNIT 1 IV

. Amendment No. 23, 29 s

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t INDEX LIMITING CONDITIONS FOR OPERATION AND SURVEILLANCE REQUIREMENTS SECTION PAGE 3/4.3 INSTRUMENTATION T

r REACTOR PROTECTION SYSTEM INSTRUMENTATION................ 3/4 3-1 3/4.3.1 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION.....................

3/4 3-9 ls 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION. 3/4 3-30 3/4.3.4 CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION............

3/4 3-39 3/4.3.5 MONITORING INSTRUMENTATION Seismic Monitoring Instrumentation...................... 3/4 3-44 Remote Shutdown Monitoring Instrumentation..............

3/4 3-47 Post-accident Monitoring Instrumentation................

3/4 3-50 Source Range Monitors...................................

3/4 3-53 Chlorine Detection System...............................

3/4 3-54 Chloride Intrusion Moni tors............................. 3/4 3-55 Fire Detection Instrumentation..........................

3/4 3-59 3/4.3.6 ATWS RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION..... 3/4 3-62 3/4.4 REACTOR COOLANT SYSTEM

/

3/4.4.1 RECIRCULATION SYSTEM Recirculation Loops.....................................

3/4 4-1 Jet Pumps...........................r.....-..............

3/4 4-2 Idle Recirculation Loop Startup....'.....................

3/4 4-3 a:

3/4.4.2 SAFETY / RELIEF VALVES...................................'.

3/4 4-4 3/4.4.3 REACTOR COOLANT SYSTEM LEAKAGE Leakage Detection Systems...............................

3/4 4-5 Operational Leakage.....................................

3/4 4-6 t--

9 BRUNSWICK.- UNIT 1 V

Amendment No. 123,14. 29 1,

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, u INDEX LIM 5 TING CONDITIbHS FOR OPERATIud _AND SURVEILLANCE REQ i

PAGE SECTION 3/4.4-REACTORCOOLANTSYSTEM_(Cor.t.inued) 3/4 4-7 3/4.4.4.. CHEMISTRY............................................

3/4 4-10 3/ 4. 4. 5. S PECI FI C ACTIV ITY....................................

3/4.4.6 PRESSURE / TEMPERATURE LIMITS 3/4 4-13 Reactor Coolant System...............................

3/4 4-18 Re a cto r S te am 0 ome...................................

3/4 4-19 3/4.4.7 MAIN STEAM LINE ISOLATION VALVES....,................

3/4 4-20 3/4.4.8 STRUCTURAL INTEGRITY.................................

3/4.5 EMERGENCY CORE COOLING SYSTEMS 3/4 5-1 3/4.5.1 HIGH PRESSURE COOLANT INJ ECTION SYSTEM...............

3/4 5-3

~

3/4.5.2 AUTOMATIC DEPRESSURIZATION SYSTEM....................

3/4.5.3 LOW PRESSURE COOLING SYSTDIS 3/4~5-4 Core Spray Sys. tem...............................~.....

Low Pressure Coolant Injection System..............'..

3/4 5-7 3/4 5-9 3/4.5.4 SUPPRESSION P00L.....................................

3/4.9 CONTAINMENT SYSTEMS 3/4.6.1 PRIMARY CONTAINMENT b

3/4 6-1 Primary Containment Integrity........................

3/4 6-2 1

Primary Containment Leakage..........................

1 3/4 6-4 Primary Containment Air Lock.........................

3/4 6-6 Primary Containment Structural Integrity.............

3/46-7f Primary Containment Internal Pressure................

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3/4 6-8

, Primary Containment Average Air Temperature..........

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

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LIMITING CONDITION FOR OPERATION I

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3.2.1 Al1 AVERAGE PLhiAR LINEAR HEAT GENERATION RATES (APLHGR's) for each type of fuel as a function of AVERAGE PLANAR EXPOSURE shall not

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exceed the limits shown in Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 or 2.3.1-6..

=-

APPLICABILITY: CONDITION 1, when THEPSL POWER > 25% of RATED THERMAL r

POWER.

ACTION:

With an APLHGR exceeding the limits of Figure '3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1.-5 or 3.2.1-6, initiate corrective action within 15 minutes l

1 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 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 .

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T SURVEILLANCE REQUIREMENTS i

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4.2.1 All APLHGR's shall be verified to be equal to or less than the applicable limit detennined from Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5 or 3.2.1 -

a.

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

~4.

i b.

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

)

1 c.

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

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P09ER 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 relationships:'

S < (0.66N + 54%) T gg s (0.ssw + 42%) T S

S and 5,3 are in percent of RATED THERMAL POWER, where:

W = Loop recirculation flow in percentor 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 < 1.0), and Design TPF for:

8 x 8 fuel = 2.45.

8 x 8R fuel = 2.48.

j P8 x BR fuel = 2.48.

j IPPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL P0HER.

~

ACTION:

With 5 or So, exceeding the allowable value, initiate corrective action within 15 minutes and. continue corrective action so that S and SDB are within the required limits 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 withi,n the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILEANCE'REOUIREMENTS 4.2.2 The MTPF for each class of fuel shall be detemined, 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 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 0.00333 hours 1.984127e-5 weeks 4.566e-6 months when the reactor is operating with a LIMITING CONTROL ROD PATTERN for MTPF.

,BRUNSWICX - UNIT 1 3/4 2-8

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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIfkJMCRITICALPOWERRATIO

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LIMITING CONDITION FOR OPERATION 3.'2. 3 The MINIMUM CRITICAL POWER RATIO (MCPR), as a function of core shown in Figure l

flow, shall be equal to or.. greater than MCPR x the Kf

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3.2.3-1. where..MCPR values are:

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8x8R fuel 1.24 1.30 P8x8R fuel 1.30 1.32 APPLICABILITY:

CONDITION 1, when THERMAL POWER > 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 within 15 minutes and continue corrective action so that MCPR is equal to or greater than 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 .

SNRVEILLANCEREOUIREMENTS

~~

4.2.3 MCPR, as a function of core flow, shall be detennined to be equal to or greater than the applicable limit determined from Figure 3.2.)-1:

a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL Pt31ER 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 R00 PATTERN for MCPR.,-

Beginning of Cycle 3.

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POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 4

3.2.4 All LINEAR HEAT GENERATION RATES (LHGR's), shall not exceed 13.4 kw/ft.

APPLICABILITY: CONDITION 1, when THERMAL POWER > 25% of RATED THERMAL POWER ACTION:

With the LHGR of any fuel rod exceeding 13.4 kw/ft.

initi.ite corrective 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 LHGRs shall be detemined to be equal to or less than 13.4 kw/ft:

a.

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

Within 12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months after completion of a THERMAL POWER increase of at least 15% of RATED THERMAL POWER, and 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 R00 PATTERN for LHGR.

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BRUNSWICK-UNIT 1 3/4 2-11 Amendment No. 73, 29 5

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TABLE 3.3.4-1 (Continued)

CONTROL ROD WITHDRAWAL BLOCK INSTRUMENTATION NOTE When THERMAL POWER exceeds the preset power level of the RWM and RSCS.

a.

The minimum number of OPERABLE CHANNELS may be reduced by one for up to 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months in one of the trip systems for maintenance and/or testing except for Rod Block Monitt function.

b.

This function is bypassed if detector is reading > 100 cps or the IRM channels are on range 3 or higher.

c.

This function is bypassed when the associated IRM channels are on J

range 8 or higher, d.

A total of 6 IRM instruments must be OPERABLE.

e.

This function is bypassed when the IRM channels are on range 1.

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9 488

TABLE 3.3.4-2 CONTROL R00 WITHDRAWAL BLOCK INSTRUMENTATION SETPOINTS

.Q{g[ 3 g 5-TRIP FUNCTION AND INSTRU'ENT NUMBER TRIP SETPOINT ALLOWABLE VALUE ..Fr. (i-. M 5 i y 1. APRM (C51-APRM-CH.A.B C.D.E.F) h si c ...,kn:s ~ v. =. I iip 1 [' a. Upscale (Flow Blased) < !' 66 H + 42%) RTPT- <(0.,66.W+42%)RWF T* T* b. Inoperative NA NA ~' c. Downscale > 3/125 of full scale > 3/125 of full scale d. Upscale (Fixed) i 12% of RATED THERMAL POWER i12% of RATED THERMAL POWER -. v;.. -.. -c 1 2. RODBLOCKMONITOR(C51-RBM-CH.A,B) a. Upscale < (0.66W + 41%) T* < (0.66 W + 41%) T* b. Inoperative NA HIPT WA MTPF c. Downscale > 3/125 of full scale > 3/125 of full scale. 1 i + 3. SOURCEPANGEMONITORS(C51-SRM-K600A,B,C,0) (l ' [l a. Detector not full in NA NA. ~ Y 5 5 O b. Upscale < 1 x 10 cps < 1 x 10 c.ps c. Inoperative RA NA d. Downscale > 3 cps > 3 cps 4. ' INTERMEDIATE RANGE MONITORS (C51-IRM-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 g d. Downscale > 3/125 of full scale > 3/125 of full scale 4 T=2.43 for 8 x 8 fuel. g T=2.48 for 8 x 8 R fuel. T=2.48 for P8x8R fuel. z P N i l' e e.

~ 1 1 3/4.2 POWER DISTRIBUTION LIMITS i 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 effects of i fuel pellet densification. 3/4.2.1 AVERAGE PLANAR LINEAR HEAT GENERATION RATE i This specification assures that the peak cladding temoerature 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) fo11bing 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 a assembly. The peak clad temperature is calculated assuming a LHGR for the highest powered rod which is equal to or less than the design LHGR corrected for densification. This LHGR times 1.02 is used in the heatup code along with the exposure dependent steady state gap conductance and rod-to-rod local peaking factor. The Technical Specification APHGR 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 and 3.2.1-6. l The calculational procedure used to establish the APLHGR shown on Figure 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, 3.2.1-5, and 3.2.1-6 is based l on a loss-of-coolant accident analysis. The analysis was performed r using General Electric (GE) calculational models which are consistent K 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. Differences in this analysis compared to previous analyses perfomed with Reference 1 are: (1) The analyses 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, and 3.2.1-6, (2) Fission product decay is computed assuming an energy release rate of 200 MEV/ Fission; (3) Pool boiling is assumed after nucleate boiling is lost.during the flow stagna-tion period; (4) The effects of core spray entrainment and cou'nter-current 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. U, 29 ,.y. __.g

y m .u 5 w -.3 p.. Bases Table B 3.2.1-1 SIN IFICANT INPUTS PARAMETERS TO THE ~ LOSS-OF-COOLANT ACCIDENT ANALYSIS ~ ? ' g FOR BRUNSWICK-UNIT 1 .J 1 . ;. : : c..;:. ;.:

.m..: ..,

.:. a...l... - ? ' y ' *r p s. c : - e ~ m Plant Parameters; c.;c g Core Thennal Power.................... 2531 Mwt which corresponds ~ 105t of rated steam flow

l 6

Vessel Steam Output.............. 10.96 x 10 Lbm/h which corresponds to 105% of rated steam flow Vessel Steam Dome Pressure......... 1055 psia o Recirculation Line Break Area for Large Breaks a. Discharge 2.4 ft (DBA);1.9ft2 (80% DBA) g b. Suction 4.2 ft Number of Drilled Bundles 560 Fuel Parameters:- t PEAK TECHNICAL INITIAL SPECIFICATION DESIGN MINIMUM LINEAR HEAT AXIAL CRITICAL FUEL BUNDLE GENERATION RATE PEAKING PCWER FUEL TYPES GEOMETRY (kw/ft) FACTOR RATIO ** l All 8x8 13.4 1.4 1.2 l-A more detailed list of input to each model and'its source is presented in ^' Section II of Reference 1. e

This power' level meets the, Appendix K requirement of 102%.

/ 2

- To account for the 2% uncertainty in bundle power required by Appendix K, the SCAT calculation is performed with an MCPR of 1.18 (i.e.,1.2 divided by 1.02) for a bundle with an initial MCPR of 1.20.

y BRUNSWICK - UNIT 1 B 3/4 2-2 'AmendmentNo723(Correction) s g. , ~. .s

r POWER DISTRIBUTION LIMITS BASE 1 3/4.2.2 APRM SE7 POINTS f '~ The fuel cladding integrity safety limits of Specification 2.1 were based on a TOTAL PEAKING FACTOR of 2.45 for 8 x 8 fuel and 2.48 for 8 x 8R and P8 x 8R fuel. The scram setting and rod bicch.ft:nctions of l 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 fomula in this specification when the combination of THERMAL POWER and peak flux indi-cates a TOTAL PEAKING FACTOR greater than 2.45 for 8 x 8 fuel and 2.48 for 8 x 8R and P8 x 8R fuel. The method used to detemine the design l TPF shall be consistent with the method used to detemine the MTPF. 3/4.2.3 MINIMUM 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 Safety git MCPR of 1.07, and an analysis of abnormal operational transients For any abnomal operating tran-sient analysis evaluation with the initial condition of the reactor being at the steady state operating limit, it is required that the resulting MCPR does not decrease below the Safety Limit MCPR at any time during the transient assuming instrument trip setting 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 eyaluated were loss of flow, increase in pressure and power, positive reactivity insertion, and coolant temperature decrease. The limiting transient which determines the required stea'y state'MCPR d limit is the turbine trip with failure of the' turbine by pass. This transient yields the largest A 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 the analysis of abnomal operational <tran-a sients an initial fuel bundle MCPR was detemined. This parameter is based on the bundle flow calculated by a GE multi-channel steady gte flow distribution model as described in Section 4.4 of NED0-20360 and on core parameters shown in Reference 3, response to Items 2 and 9. ...a E t BRUNSWICK - UNIT 1 B 3/4 2-3 Amendment No.' D, 29 r 's ' ~ /.

.it:.}: ^ + .~ f? ~ . - ~. l ^ POWER DISTRIBUTION LIMITS s -BASES 7-J' L-

~ -

^ a ,a MINIMUM CRITICAL POWER RATIO (Continuedl22-. I-4 1= evaNation.of a given transient beg' Ins with the system" initial par- ~ Th N ameters shown in Attachment 5 of Reference 6 that are input to a GE-core h dynamic behavior. transient computer program described in NEDO-10802(5). '~ Also,~ the void reactivity coefficients that were input to the transient calculational procedure are based on a new rethod of calculation temed NEV which provides a better agreement between the calculated and plant instrument power distributions. The outputs of this program along with the initial MCPR fom the input for further analyses of the themally limiting bundle with the single channel transient thermal hydraulic SCAT code described in NED0-20566(1). The principal result of this evaluation i is the reduction in MCPR caused by the transierft. The purpose of the K, factor is to define operating limits at other than rated flow conditions. At less than 100". flow the required MCPR factor., Speci-is the. product of the operating limit MCPR and the Kf fically, the K factor provides the required themal margin to protect f The most limiting transient initiated against a flow increase transient. from less than rated flow conditions is the recirculation pump speed up caused by a motor-generator. speed control failure. ~7 ,,C' lu ~ factors" For operation in the automatic ' flow control mode, the K[2.3 will not be assure that the operating limit MCPR of Specification 3 violated should the most limiting transient occur at less than rated factors assure that the flow. In the manual flow control mode, the Kf Safety Limit MCPR will not be violated should the most limiting transient ~' occur at less than rated flow. ,I \\ The K factor values shown in Figure 3.2.3-1 were developed generically f The K which are applicable to all BWR/2, BWR/3, and BWR/4 reactors. fact thennal power at rated core flow... L.: facto'rs were calculated such For the manual flow control mode, the K f f that the maximum flow state (as limited by the pump scoop tube set.. i l point).and the corresponding core power (along the rated flow cont'rol line), the limiting bundles re. ative power was adjusted until.the l MCPR was slightly above the Safety Limit. Using this relative bundle l power, the MCPR's were calculated at different points along the. rated The ratio flow control line corresponding to different core flows. of the MCPR calculated at a given point of core flow, divided by the operating ' limit MCPR detemines the K. l f l EU.:::SL:!CK UNIT 1 B 3/4 2-4 9 e, e = =e* ah y me = e

e
f g

g i

REACTOR COOLANT SYSTEM 3/A.4.2 SAFETY / RELIEF VALVES LIMITING CONDITION FOR OPERATION 3.4.2 The safety valve function of all reactor coolant system safety / relief valves shall be OPERABLE with lift settinos within + 1% of the followino values.*# ~ l 4 Safety-relief valves 91105 osig. 4 Safety-relief valves 91115 psig. 3 Safety-relief valves 01125 psig. APPLICA51ILITY: CONDITIONS 1, 2 and 3. ACTION: With the safety valve function of one,fety valve function of a. safety / relief valve inocerable, restore the inoperable sa the valve to OPERABLE status within 31 days or be in at least HOT SHilTDONN within the next 12 hours and in COLD SHUTOOWN within the following 24 hours. p b. With the safety valve function of two safety / relief valves inoperable, restore the inoperable safety valve function of at least one of the valves to OPERABLE status within 7 days or be in at least HOT SHUTDONN within the next 12 hours and in COLD SHUTDOWN within the following 24 hours. With the safety valve function of more than two safety / relief c. valves inoperable, be in at least HOT SHUTDOWN within 12 hours and in COLD SHUTDOWN within the next 24 hours. SURVEILLANCE REnUIREMEMTS 4.4.2 The safety valve function of each of the above required safety / relief valves shall be demonstrated OPERABLE by verifying that the 3ellows on the safety / relief valves have integrity, by instrumentation indication, at least once Der 24 hours. 'The lift setting pressure shall correspond to ambient c:;nditions o.f the valves at nominal operating temperature and pressure.

From Spring,1980 until the maintenance outage in Sept.,1980, the safety-relief valve lift settings shall be arranged such that each safety-relief valve pair has a-minimum nominal lift setting differential of 20 psi and shall be within + 1% of the following values:

2 Safety-relief valves 01095 psig 3 Safety-relief valves @ 1105 psig 3 Safety-relief valves 91115 psig 3 Safety-relief valves @ 1125 psig 1RUNSWICK - UNIT 1 3/4 4-4 Amendement No. H, 29 --}}

ML19320B687

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