ML20210P461

Amend 123 to License DPR-62,changing Tech Specs to Modify Min Critical Power Ratio Values & Delete Ref to 8X8 Fuel Type in Fuel Cycle 7

ML20210P461
Person / Time
Site:	Brunswick
Issue date:	04/30/1986
From:	Muller D Office of Nuclear Reactor Regulation
To:	Carolina Power & Light Co
Shared Package
ML20210P467	List: ML20210P461 ML20210P472
References
DPR-62-A-123 NUDOCS 8605130396
Download: ML20210P461 (23)

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UNITED STATES 8

NUCLEAR REGULATORY COMMISSION o

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CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.123 License No. DPR-62 1.

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

A.

The application for amendment by Carolina Power & Light Company (thelicensee)datedDecember 20, 1985, as supplemented March 28, 1986, complies with the standards and re Energy Act of 1954, as amended (the Act)quirements of the Atomic and the Commission's rules and regulations set forth in 10 CFR Chapter I; 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 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 Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. DPR-62 is hereby amended to read as follows:

8605130396 860430 PDR ADOCK 05000324 P

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' l (2) Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.123, are hereby incorporated in the license.

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

3.

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

4 FOR THE NUCLEAR REGULATORY COMMISSION h.AV Daniel R. Muller Director BWR Project Directorate #2 Division of BWR Licensing a

Attachment:

F Changes to the Technical Specifications Date of Issuance: April 30, 1986 T'

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2

l ATTACHMENT TO LICENSE AMENDMENT NO.123 FACILITY OPERATING LICENSE N0. DPR-62 DOCKET N0. 50-324 i

Replace lthe following pages of the Appendix A Technical Specifications with the enclosed pages. The revised areas are indicated by marginal lines.

Pages 3/4 2-1 3/4 2-2 3/4 2-3 3/4 2-4 3/4 2-5 3/4 2-6 3/4 2-7 g

3/4 2-8 F

3/4 2-9 3/4 2-10 3/4 2-11 3/4 2-12 3/4 2-13 3/4 2-14 3/4 3-42 3/4 3.

B 3/4 2-1 B 3/4 2-3 B 3/4 2-5 5-1 i

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3/4.2 POWER DISTRIBUTION LIMITS i

3/4.2.1 AVERACE PLANAR LINEAR HEAT CENERATION RATE t

I LIMITING CONDITION FOR OPERATION i

3.2.1 All AVERACE PLANAR LINEAR HEAT CENERATION RATES (APLHCR's) for each

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type of fuel as a function of AVERACE PLANAR EXPOSURE shall not exceed the following limits:

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, and 3.2.1-5.

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

ACTION: With an APLHCR exceeding the limits of Figures 3.2.1-1, 3.2.1-2, I

3.2.1-3, 3.2.1-4, and 3.2.1-5, initiate corrective action within 15 minutes and continue corrective action so that APLHCR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within:

the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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g, SURVEILLANCE REQUIREMENTS

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4.2.1 All APLHCR's 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, and 3.2.1-5:

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a.

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

r b.

Within 12 nours after completion of a THERMAL POWER increase of at least 152 ot RATED THERMAL POWER, and L

c.

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

operating with a LIMITING CONTROL ROD PATTERN-for APLHCR.

b l

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i BRUNSWICK - UNIT 2 3/4 2-6 Amendment No. 707, 123 i

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i 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 (Sgg) shall be established according to the following relationship:

3 1

S $ (0.66W + 54%) T N

4 i

SRS 3 (0.66W + 42%) T l

f 1

I wnere:

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 $ 1.0). and Design TPF for:

8 x 8R fuel = 2.39 PS x 8R fuel = 2.39 i

BP8 x SR fuel = 2.39 4

}

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

equal to 25% of RATED THERMAL POWER.

m ACTION:

i

.i With S or S exceeding the allowable value, initiate corrective action within RB 15 minutes and continue corrective action so that S and S are within'the RB required limits within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or reduce THERMAL POWER,to less than 25% of j

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

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SURVEILLANCE REQUIREMENTS a

i 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:

i a.

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

Within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after completion of a THERMAL POWER increase' of at j

least 15% of RATED THERMAL POWER, and l

c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when_the reactor.is j

operating with a LIMITING CONTROL ROD PATTERN for M;TPF.

+a

+

i BRUNSWICK - UNII 2 3/4 2-7

• i Amendment No. JWJ, j

j

123, i

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POWER DISTRIBUTION LIMITS 3/4.2.3 MINIMUM CRITICAL POWER RATIO 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 Kg shown in 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.

ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:

1.

MCPR for 8 x 8R fuel = 1.31 2.

MCPR for P8 x 8R fuel = 1.33 3.

MCPR for BP8 x 8R fuel = 1.33 b.

EOC minus 2000 MWD /t to EOC with ODYN OPTION A analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:

1.

MCPR for 8 x 8R fuel = 1.41 l[

2.

MCPR for P8 x 8R fuel = 1.44 3.

MCPR for BP8 x 8R fuel = 1.44 i

BOC to EOC minus 2000 MWD /t with ODYN OPTION B analyses in effect and c.

the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:

1.

MCPR for 8 x 8R fuel = 1.29 2.

MCPR for P8 x 8R fuel = 1.29 3.

MCPR for BP8 x 8R fuel = 1.29 d.

EOC minus 2000 MWD /t to EOC with ODYN OPTION B analyses in effect and the end-of-cycle recirculation pump trip system inoperable, the MCPR limits are listed below:

1.

MCPR for 8 x 8R fuel = 1.29 2.

MCPR for P8 x 8R fuel = 1.32 3.

MCPR for BP8 x 8R fuel = 1.32 APPLICABILITY: OPERATIONAL CONDITION 1 when THERMAL POWER is greater than or equal to 25% RATED THERMAL POWER BRUNSWICK - UNIT 2 3/4 2-8 Amendment No. J$J,

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LIMITING CONDITION l'OR OPERA

• ION (Continued) j AC* ION:

i With MCPR. as a function of core flow, less than the applicable limit determined from FigurU3.2.3-1 initiate corrective action within 15 minutes snd restore MCPP. to yithin the applicable limit within 4. hours or reduco THERMAL POWER to.less 'thari 25% of' RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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3URVEILLANCEREQUIkEMENTE;,

I. 2.3.1 MCP'n';' ai a fur.c't. ion of core flow. shall be determined to be equal to s

or greater t:isQthe-applicable limit, determined from Figure 3.2.3-1:

a.

At least once pe 24' hours.

s, a

b.

Within 12' hours after3 completion of a THERMAL POWER increase of at least ISO of RATED THERMAL POWER, and s

'c c.

Initially.ano at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when the reactor is l

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operating in a LIMITING CONTROL ROD PATTERN for MCPR.

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

LIMITING C03DITION FOR OPERATION 3.2.3.2 For the OPTION B MCPR limits listed in specification 3.2.3.1 to be usea, the evele average 20% scram time (t

) shall be less than or equal to the Option 'B scram time limit ( B ' "

and T are determined as ve B

ollows:

n

?

N.

t=1 t t

=

,y, n

N.

T L

L i=1 Surveillance test number,

- =

s n = Numoer of surveillance tests performed to date in the cycle (including BOC),

N: = Numoer of' rods tested in the ich surveillance test, and

{=Averagescramtimetonotch36forsurveillancetest i f~

-g 1.65 ( n N.}

(),where:

a +

=-

[

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

th Nt = Number of rods tested in the i surveillance test Ni = Number of rods tested at BOC, u = 0.834 seconds (mean value for statistical scram time distribution from de-energization of scram pilot valve solenoid to ptekup on notch 36),

a = 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.

i i

i BRUNSWICK - UNIT 2 3/4 2-10 Amendment No. g),

123,

J POWER DISTRIBUTION LIMITS LIMITINC CONDITIONS FOR OPERATION (Continued)

ACTION:

Jithin twelve hours after determining that t is greater than T, the

perating Limit MCPRs shall be either:

a.

Adjusted for each fuel type such that the operating limit MCPR is the maximum of the non pressurization transient MCPR operating Limit (from Table 3.2.3.2-1) or the adjusted pressurization transient MCPR operating limits, where the adjustment is made by:

T

- T fMCPR

- MCPR I

MCPR

= MCPR

+

=-

B ption A option B#

ad j.usted option B where: :3 = 1.05 seconce, control rod average scram insertion time limit to notch 36 per Specification 3.1.3.3, t

t f~

MCPRoption A = Determined from Table 3.2.3.2-1

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

The OPTION A MCPR limits listed in Specification 3.2.3.1.

SURVEILLANCE REQUIREMENTS 4.2.3.2 The values of I and r shall be cetermined and compared each time a scram time test is perfEImed.

kherequirement for the frequency of scram time testing shall be identical to Specification 4.1.3.2.

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

123,

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TABLE 3.2.3.2-1 E

h TRANSIENT OPERATINC LIMIT MCPR VALUES R

l TRANSIENT FUEL TYPE h

8x8R P8x8R BP8 x 8R s

M NONPRESSURIZATION TRANSIENTS BOC + EOC 1.29 1.29 1.29 PRESSURIZATION TRANSIENTS 2

MCPR MCPR MCPR MCPR MCPR HCPR A

B A

B A

B w BOC + EOC - 2000 1.31 1.17 1.33 1.17 1.33 1.17 L

" EOC - 2000 + EOC 1.41 1.29 1.44 1.32 1.44 1.32 1

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r POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT CENERATION RATE LIMITING CONDITION FOR OPERATION j

3.2.4 The LINEAR HEAT CENERATION RATE (LHCR) shall not exceed 13.4 kw/ft for i X 3R. PB X 8R, and BPS x BR fuel assemblies.

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

ecual to 25% of RATED THERMAL POWER.

i ACTION:

[

With the LHCR of any fuel rod exceeding the above limit, initiate corrective action within 15 minutes and continue corrective action so that the LHCR is within the limit within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. or reduce THERMAL POWER to less than 25% of RATED THERMAL POWER within the next 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />.

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SURVEILLANCE REOUIREMENTS 4.2.4 LHCRs shall be determined to be equal to or less than the limit:

a.

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

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

c.

Initially and at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> when ti tctor is operating on a LIMITING CONTROL ROD PATTERN fo.

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E CONTROL Roll WITIIDRAWAL 15 LOCK INSTRUMENTATION SLTPolNTS 5*

TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT AI.I.OUAlli.E _ VAI.UE I

E 1.

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

Upscale (Flow Biased)

$ (0.66W + 42%)

T*

$ (0.66W + 42%)

T*

y MTPF MTPF w

b.

Inoperative NA NA c.

Downscale

> 3/125 of full scale

> 3/125 of full scale d.

Upscale (Fixed) 3 12% of RATED TilERMAl. POWER 312%ofRATEDTilERMALPOWER 2.

ROD HLOCK HONI'IUR ( C51-RBM-Cll. A,II) a.

Upscale

<- (0.66W + 39%)

T*

< (0.66W + 39%)

T*

b.

Inoperative 5A MTPF UA MTPF w

c.

Downscale 3 3/125 of full scalc 3 3/125 of full scale w

3.

SOURCE RANCE MONITORS (C51-SRM-K600A,B,C,D) 5-a.

Detector not full in NA NA 5

5 b.

Upscale

$ 1 x 10 cps

$ 1 x 10 cp, c.

Inoperative NA NA d.

Downscale 3 3 cps 3 3 cps 4.

INTERMEDIATE RANCE MONITORS ( C$ 1-I RM-K601 A, B,C, D, E, F,C,II) a.

Detector not. full in-NA NA b.

Upscale

$ 108/125 of full scalc

$ 108/125-of full scale c.

I nope ra t i ve NA NA d.

Downscale 3 3/125 of full scale 3 3/125 of full scale 5.

SCRAM DISCllARGE VOLUME (Cl2-LS!!-N013E)

Of a.

Water I.evel liigh 5 73 gallons 5 73 gallons a

5 T=2.39 for 8 x 8R. fuel.

T=2.39 tor PB x BR luel.

u T=2.39 for HP8 x BR luel.

5?

.I[,.-

1 INSTRUMENTATION i

END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM INSTRUMENTATION-LIMITING CONDITION FOR OPERATION i

f 3.3.6.2 The end-of-cycle recirculation pump trip (EOC-RPT) system instrumentation channels shown in Table 3.3.6.2-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint l

column of Table 3.3.6.2-2 and with the END-OF-CYCLE RECIRCULATION PUMP TRIP SYSTEM RESPONSE TIME as shown in Table 3.3.6.2-3.

j APPLICABILITY: OPERATIONAL CONDITION 1, when THERMAL POWER is greater than or equal to 30% of RATED THERMAL POWER.*

ACTION:

L With an end-of-cycle recirculation pump trip system instrumentation a.

channel trip setpoint less conservative than the value shown in the j

Allowable Values Column of Table 3.3.6.2-2, declare the chan 4el l

inoperable until the channel.is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

[

[

t b.

With the number of OPERABLE channels'one less than required by the

(

Minimum OPERABLE Channels per Trip System requirement for one or both j

trip systems, place the inoperable channel (s) in the tripped condition within one hour.

l c.

With the number of OPERASLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one j

trip system and l

1.

If the operable channels consist of one turbine control valve f

channel and one turbine stop valve channel, place both inoperable channels in the tripped. condition within one hour.

j l

2.

If the inoperable channels include two turbine control valve channels or two turbine stop valve channels, declare the trip system operable.

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d.

With one trip system inoperable, restore the inoperable trip system i

to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or take the ACTION required by i

Specification 3.2.3.

e.

With both trip systems inoperable, restore at least one trip system to OPERABLE status within one hour or take the ACTION required by Specification 3.2.3.

• During the current cycle operation, the end-of-cycle recirculation pump trip i

(EOC-RPT) system will be inoperable (manually bypassed); therefore, Specification 3.3.6.2 above does not apply. The provisions of Specification 3.0.4 are not applicable.

BRUNSWICK - UNIT 2 3/4 3-82 Amendment No. JSJ,

123,

]

3/4.2 POWER DISTRIBUTION LIMITS BASES The specifications of this section assure that the peak cladding

emperature following the postulated design basis loss-of-coolant accident will not exceed tne 2200*F limit specified in the Final Acceptance Criteria i

iFAC) issued in. June 1971 considering the postulated effects of fuel pellet densification.

3.2.1 AVERACE PLANAR LINEAR HEAT CENERATION 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 pean cladding temperature (PCT) following a postulated loss-of-

^

coolant accident is primarily a function of the average heat generation rate ci 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 cladding temperature is calculated assuming a LHCR for the highest powered rod which is equal to or less than the design LHGR corrected for densification.

g-This LHCR times 1.02 is used in the heatup code along with the exposure-F dependent steady state gap conductance and rod-to-rod local peaking - f actor.

The Technical Specification APHCR is this LHCR 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, and 3.2.1-5.

The calculational procedure used to establish the APLHCR shown on i

Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5 is based on a loss-of-l coolant accident analysis. The analysis.was performed using'Ceneral Electric (CE) 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.

Differences 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 MAPLHCR shown in l

Figures 3.2.1-1, 3.2.1-2, 3.2.1-3, 3.2.1-4, and 3.2.1-5;'(2) Fission product l

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' stagnation period; and (4) The effects 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.

I e

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

123,

a 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 2.39 for 8 x 8R P8 x 8R, and BP8 x 8R fuei. The scram setting and rod block functions of the APRM instruments must ce aajusted to ensure that the MCPR does not become less than 1.0 in the

earaded situation.

The scram settings and rod block settings are adjusted in accorcance with the formula in tnis specification when the combination of THERMAL POWER and peak flux indicates a TOTAL PEAKING FACTOR greater than 2.39 for 8 x 8R P8 x 3R. and BP8 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 enanze. The method used to determine tne design TPF shall be consistent witn ene metnod useo to cetermine the MTPF.

3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel f_

claddingintegrttySateg{)LimitMCPRof1.07,andananalysisofabnormal operational transients.

For any abnormal operating transient 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 an 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.

Unless otherwise stated in cycle specific reload analyses, 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 a MCPR.

Prior to the analysis of abnormal operational transients an initial fuel bundle MCPR was determined. This parameter is based on the bundleflowcalculatedbyaGEmultichannelsteggystateflowdistribution model as described in Section 4.4 of NEDO-20360 and on core parameters shown in Reference 3, response to Items 2 and 9.

BRUNSWICK - UNIT 2 B 3/4 2-3 Amendment F e. J0J,

123,

i i I.

POWER DISTRIBUTION LIhITS F

I i-BASES MINIMUM CRITICAL POWER RATIO (Continued)

I

?ce operation in the automatic flow control mode, the same procedure was employed except the initial power distribution was established.such that the.

MCPR was equal to the operating limit MCPR at rated power and flow.

The se factors shown in Figure 3.2.3-1 are conservative for the Ceneral I'.ectric Plant operation with 8 x 8R fuel assembly types because the operating f

• imit MCPRs of Specification 3.2.3 are-greater than the original 1.20 operating limit MCPR used for the generic derivation of K.

g t-At core thermal power levels less than or equal to 25%, the reactor will be operating at minimum recirculation pump speed and the moderator void content i

vill be very small. For all designated control rod patterns which may be employed at this point, operating plant experience indicated that the t

}

resulting MCPR value is in excess of requirements by a considerable margin.

With this low void content, any inadvertent core flow increase would only l

jl piace operation in 1 more conservative mode relative to MCPR. During initial;

[

start-up testing of the plant, a MCPR evaluation will be made at 25% thermal g,

power level with minimum recirculation pump speed. The MCPR margin will thus be demonstrated such that future MCPR evaluation below this power level will i

be shown to be unnecessary. The daily requirement for calculating MCPR above j

25% rated thermal power is sufficient since power distribution shifts are very

[

slow when there have not been significant power or control rod changes. The i

i l

requirement for calculating MCPR when a limiting control rod pattern is I

approached ensures that MCPR will be known following a change in power or j

power. shape, regardless of magnitude that could place operation'at a thermal Limit.

i i

i 3.2.4 LINEAR HEAT CENERATION RATE i

The LHCR specification assures that the linear. heat generation rate in any

(

rod is less than the design linear heat generation even if fuel pellet l

densification is postulated. The power spike penalty-specified is based on j

the analysis presented in Section 3.2.1 of the CE topical report NEDM-10735 j'

Supplement 6, and assumes a linearly increasing variation in axial gaps l

between core bottom and top,-and assures with a 95% confideace that no more than one fuel rod exceeds the design linear heat generation rate due to power s p iki n g.'.

l I

i i

e i

i i

i-BRUNSWICK - UNIT 2 8 3/4 2-5 Amendment No.g),

123, v

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y.-,-.yy.,,-mym

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,,,,,_,,,,._,,,,,,v..,-y-.,,,,y.

..,,,--,-,-.y.,,,,,,,~,c,%,---

9 5.0 DESIGN FEATURES 5.1 SITE j

EXCLUSION AREA f.;.; The exclusion area shall be as shown in Figure 5.1.1-1.

LC*4 POPULATION ZONE J.

3.1.2 The low population :one shall be as shown in Figure 5.1.2-1.

i SITE BOUNDARY 4

4 5.1.3 The SITE BOUNDARY shall be as shown in Figure 5.1.3-1.

For the purpose l

of effluent release calculations, the boundary for atmospheric releases is the 3:TE BOUNDARY and the bouncarv for Liquid releases is the SITE BOUNDARY prior o dilution in the Atlantic Ocean.

5.2 CONTAINMENT i

4 CONFIGURATION t

l 5.2.1 The PRIMARY CONTAINMENT is a steel-lined, reinforced concrete structure composed of a series of vertical right cylinders and truncated cones which form a drywell. This drywell is attached to a suppression chamber through a series of vents. The suppression chamber is a concrete, steel-lined pressure vessel in the shape of a torus. The primary containment has a minimum free air volume of 288,000 cubic feet.

DESIGN TEMPERATURE AND PRESSURE c

I 5.2.2 The primary containment is designed and shall be maintaineo for:

a.

Maximum internal pressure.62 psig.

l l

b.

Maximum internal temperature: drywell 300*F i

Suppression chamoer 200*F i

Maximum external pressure 2 psig.

i c.

i 5.3 REACTOR CORE 4

i FUEL ASSEMBLIES

)

5.3.1 The reactor core shall contain 560 tuel. assemblies.

The 8 x 8R, j

P8 x 8R, BP8 x dR fuel assemblies contain 62 fuel rods.

All fuel rods shall be clad with Zircatoy 2.

The nominal' active fuel length of each fuel rod

)

shall be 150 inches for 3 x 8R, P8 x 8R, and I

i t

BRUNSWICK - UNIT 2 5-1 Amendment No. JgJ,

123,

. _. _ -- _ _ __