Text

Amends 75 & 101 to Licenses DPR-71 & DPR-62,respectively, Permitting Operation of Unit 2 for Cycle 6 & Reflecting Use of Hybrid Design Hafnium Control Rods
	ML20093C920
Person / Time
Site:	Brunswick
Issue date:	09/22/1984
From:	Vassallo D; Office of Nuclear Reactor Regulation
To:	; Carolina Power & Light Co
Shared Package
ML20093C922	List: ML20093C920; ML20093C923;
References
	DPR-62-A-101, DPR-71-A-075 NUDOCS 8410100794
	Download: ML20093C920 (23)
	v • d • e

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NUCLEAR REGULATORY COMMISSION 2 -.,,.

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f CAROLINA POWER & LIGHT COMPANY DOCKET NO. 50-325 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 1 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No. 75 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 June 26, 1984, (NLS-84-219) complies with the-standards and requirements of the Atomic Energy Act of 1954, as amended (the Act) and the Comission'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 j

Consnission; t

C.

There is reasonable assurance (i) 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 inimical to the common defense and security or to the health and safety of the public; and E.

The issuanr.e of this amendment is in accordance with 10 CFR Part 51 of the Comission'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-71 is hereby amended to read as follows:

8410100794 840922 PDR ADOCK P

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Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No. 75, are hereby incor porated 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.

FOR THE NUCLEAR REGULATORY C0t1 MISSION

~. fi '.y 1 Domenic B. Vassallo, Chief Operating Reactors Branch #2 Division of Licensing

Attachment:

Changes to the Technical Specifications Date of Issuance: September 22, 1984 4

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ATTACHMENT TO LICENSE AMENDMENT NO. 75 FACILITY OPERATING LICENSE NO. DPR-71 DOCKET NO. 50-325 Revise the Appendix A Technical Specifications as follows:

Remove Insert 5-4 5-4 l

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(BSEP-1-30)

DESIGN FEATURES 5.3 REACTOR CORE FUEL ASSMBLIES (Continued)

The initial loading shall have a maxi average enrichment of 2.35 weight percent U-235.

Reload fuel shall be similar in physical design to the initial core loading and shall have a maximum average enrichment of 2.99 weight percent U-235.

CONTROL ROD ASSMBLIES 5.3.2 The reactor core shall contain 137 control rod assemblies, each consisting of a cruciform array of stainless steel tubes containing approximately 143 inches of boron carbide, B C, powder or hafnium absorber 4

rods surrounded by a cruciform-shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTM DESIGN PRESSURE AND TM PERATURE 5.4.1 The nuclear boiler and reactor recirculation system is designed and shall be maintained:

In accordance with the code requirements specified in Section 4.2 of a.

the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements.

b.

For a pressure of 1250 psig, and 0

c.

For a temperature of 575 F.

VOLIN E 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 18,670 cubic feet.

5.5 METEOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in Figure 5.1.1-1.

-BRUNSWICK - UNIT 1 5-4 Amendment No. 75

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a CAROLINA POWER & LIGHT COMPANY DOCKET N0. 50-324 BRUNSWICK STEAM ELECTRIC PLANT, UNIT 2 AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.101 License No. DPR-62 1.

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

A.

The applications for amendment by Carolina Power & Light Company (the licensee) dated June 26, 1984, (NLS-84-219 and NLS-84-274) comply 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; 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 (i) 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:

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2.

Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.101, 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.

FOR THE NUCLEAR REGULATORY COMMISSION

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D'5menicB.Vassallo, Chief Operating Reactors Branch #2 Division of Licensing Attachment.

Changes 'a the Techn cal Specifications Date f Issuance: September 22, 1984 A

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o ATTACHMENT TO LICENSE AMENDMENT NO.101 FACILITY OPERATING LICENSE NO. DPR-62 DOCKET N0. 50-324 Revise the Appendix A Technical Specifications as indicated below.

The changed area is indicated by vertical line.

i Remove Insert 3/4 2-2 3/4 2-2 3/4 2-3 3/4 2-3 3/4 2-4 3/4 2-4 3/4 2-5 3/4 2-5 3/4 2-6 3/4 2-6 3/4 2-7 3/4 2-7 3/4 2-8 3/4 2-8 3/4 2-9 3/4 2-9 3/4 2-10 3/4 2-10 3/4 2-13 3/4 2-13 3/4 2-15 3/4 2-15 3/4 3-42 3/4 3-42 3/4 3-82 3/4 3-82 B3/4 2-3 B3/4 2-3 5-1 5-1 5-4 5-4 I

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(BSEP-2-35)

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POWER DISTRIBUTION LIMITS 3/4.2.2 APRM SETPOINTS LIMITING CONDITION FOR OPERATION The flow-biased APRM scr' n trip setpoint (S) and rod block trip set 3.2.2 a

s point (SRB) shall be established adcording to the following relationship:

S j( (0.66W + 54%) T SRB I- (0.66W + 42%) T s.

s 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 <fl.0), and Design TPF for: 8 x 8 fuel = 2.43 8 x 8R fuel = 2.39 P8 x 6R fuel = 2.39 BP8 x 8R fuel = 2.39 1

z APPLICABILITY:

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

equal to 25% of RATED THERMAL POWER.

i 1

ACTION:

f With S or S exceeding the allowable value, initiate corrective a'ction within RB 15 minutes and continue corrective action so that S and SRB 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 within the next 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months .

SURVEILLANCE REOUIREMENTS 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 bours, 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.

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BRUNSWICK - UNIT 2 3/4 2-8 Amendment No. 101

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-(BSEP-2-35)

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._ POWER' DISTRIBUTION LIMITS

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-3/4.2.3 MINIMUM CRITICAL POWER RATIO I'

LIMITING-CONDITION FOR OPERATION 3.2.3.1 The MINU4mi CRITICAL POWER RATIO (MCPR), as a function of core flow, shown in shall be equal to or greater than the MCPR limit times-the Ef Pigure 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 /aie listed below:

1

(

1.

MCPR for 8 x 8 fuel = 1.25 2.

MCPR for 8 x BR fuel = 1.26

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3.

MCPR for~P8 x 81 fuel = 1.28 4.

.MCPR for BP8 x 8R-fuel = 1.28 b.

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

1.

FCPR for 8 x 8 fuel = 1.36 l

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MCPR for 8'x 8R fuel = 1.37 g

3.

MCPRifor P8'x 8R fuel = 1.40 4.

MCPR for BP8 x 8R fuel = 1.40

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c.

BOC to EOC minus 2000 MWD /t with ODYN 0PTION B analyses in ef fect and '

i the end-of-cycle recirculation pump trip.'mystem inoperable, the MCPR f

l limits are listed below:

(

l' 1.

MCPR for 8 x 8. fuel = 1.23 i

2.

MCPR for 8 x 8R fuel = 1.24 3.

MCPR foc"P8 x 8R fuel = 1.24 s

4.

MCPR for BP8 x 8R fuel = 1.24 y

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 licits are listed below:

1..

MCPR for 8 x 8 fual = 1.24 y

2.

MCPR fde 8 x 8R fuel = 1.25

(

3.

MCPR for PS x 8R fuel = 1.28 4.

MCPR for.BP8 x SR fuel = 1.28 APPLICABILITY:

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

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<s n1 BRUNSWICK - UNIT 2-3/4 2-9 Amendment No.101

A (BSEP-2-35)

POWER DISTRIBUTION LIMITS LIMITING CONDITION FOR OPERATION (Continued)

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 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 .

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 1.ast 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 in a LIMITING CONTROL ROD PATTERN for MCPR.

BRUNSWICK - UNIT 2 3/4 2-10 Amendment No. 101

(BSEP-2-35)

TABLE 3.2.3.2-1 TRANSIENT OPERATING LIMIT MCPR VALUES TRANSIENT FUEL TYPE i

8x8 8x8R P8x8R BP8 x 8R g

NONPRESSURIZATION TRANSIENTS BOC + EOC 1.23 1.24 1.24 1.24 TURBINE TRIP / LOAD REJECT WIT 110lTI BYPASS MCPR MCPR CPR MCPR M R MCPR CPR A

B A

B

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BOC + EOC - 2000 1.25 1.08 1.26 1 08 1.28 1.09 1.28 1.09 EOC - 2000 + EOC 1.36 1.24 1.37 1.25 1.40 1.28 1.40 1.28-FEEDWATER CONTROL FAILURE MCPR MCPR MCPR MCPR MCPR MCPR MCPR MCPR A

B A

B BOC + EOC - 2000 1.17 1.11 1.17 1.11 1.17 1.11 1.17 1.11 EOC - 2000 + EOC 1.17 1.11 1.17 1.11 1.17 1.11 1.17 1.11

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(BSEP-2-3 5)

POWER DISTRIBUTION LIMITS 3/4.2.4 LINEAR HEAT GENERATION RATE LIMITING CONDITION FOR OPERATION 3.2.4 The LINEAR HEAT GENERATION RATE (LHGR) shall not exceed 13.4 kw/f t for 8 X 8, 8 X 8R, P8 X 8R, and BP8 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 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 .

4 SURVEILLANCL REQUIREMENTS 4.2.4 LHGRs shall be determined to be equal to or less than the limit:

l a.

At least once per 24 hcurs, 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 4

c.

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

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l ERUNSWICK - UNIT 2 3/4 2-15 Amendment No.101

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CONTROL ROD WITi!De1AWAL BLOCK INSTRUMENTATION SETPOINTS TRIP FUNCTION AND INSTRUMENT NUMBER TRIP SETPOINT ALLOWABLE VALUE E

1.

APRM (C51-APRM-Cll. A,B,C,D,E,F)

U a.

Upscale (Flow Biased)

< (0.66W + 42%)

T*

< (0.66W + 42%)

Y*

w MTPF MTPF b.

Inoperative NA NA c.

Downacale

> 3/125 of rull scale

> 3/125 of full scale d.

Upscale (Fixed)

< 12% of RATED T2iERMAL POWER

< 12% of RATED TilERMAL POWER l

2.

ROD BLOCK MONITOR ( C 51-RBM-Cil. A, B )

a.

Upscale

< (0.66W + 39%)

T*

< (0.66W + 39%)

T*

b.

Inoperative NA MTPF NA MTPF c.

Downscale

> 3/125 of full scale

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SOURCE RANGE MONITORS (C51-SRM-K600A,B,C,D) u a.

Detector not full in NA NA L

b.

Upscale i1x 105 cps i 1 x 105 cps c.

Inoperative NA NA d.

Downscale

> 3 cps 1 3 cps 4.

INTEP. MEDIATE RANGE MONITORS (C51-IRM-K601A,B,C,D,E,F,G,li) a.

Detector not full in NA NA b.

Upscale

< 408/125 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 DISCilARGE VOLUtlE (Cl2-LSil-N011E) a.

Water Level liigh 1 73 gallons 1 73 gallons B.

a T=2.43 for 8 x 8 fuel.

T=2.39 for 8 x SR fuel.

T=2.39 for P8 x 8R fuel.

T=2.39 for BP8 x oK fuel.

S S

(BSEP-2-35)

INSTRUMENTATION l

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

APPLICABILITY:

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

.l' ACTION:

a.

With an end-of-cycle recirculation pump trip system instrumentation channel trip setpoint less conservative than the value shown in the Allowabie Values Column of Table 3.3.6.2-2, declare the channel inoperable until the channel is restored to OPERABLE status with the channel setpoint adjusted consistent with the Trip Setpoint value.

b.

With the number of OPERABLE channels one less than required by the Minimum OPERABLE Channels per Trip System requirement for one or both trip systems, place the inoperable channel (s) in the tripped condition within one hour.

c.

With the number of OPERABLE channels two or more less than required by the Minimum OPERABLE Channels per Trip System requirement for one trip system and:

1.

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

2.

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

d.

With one trip system inoperable, restore the inoperable trip system to OPERABLE status within 72 hours8.333333e-4 days 0.02 hours 1.190476e-4 weeks 2.7396e-5 months or take the ACTION required by 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 Cycle 6 operation, the end-of-cycle recirculation puSp trip (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.101

(BSEP-2-35)

POWER DISTRIBUTION LIMITS i

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 and 2.39 for 8 x 8R,

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P8 x 8R, and BP8 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 and 2.39 for 8 x 8R, P8 x 8R, 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 change.

The method used-to determine the design TPF shall be consistent with the method used to determine the MTPF.

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3/4.2.3 MINIMUM CRITICAL POWER RATIO The required operating limit MCPRs at steady state operating conditions as specified in Specification 3.2.3 are derived from the established fuel cladding integrity Safe operationaltransients.(()LimitMCPRof1.07,andananalysisofabnormal 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.

I To assure that the fuel cladding integrity Safety Limit is not exceeded during any anticipated abnormal operational transient, the most limiting traasients 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 l

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.

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 abnormal operational transients an initial fuel bundle MCPR was determined. 7his parameter is based on the bundle flow calculated by a GE multichannel gadystateflowdistributionmodelasdescribedinSection4.4 of NEDO-20360 and on core parameters shown in Reference 3, response to Items 2 and 9.

BRUNSWICK - UNIT 2 8 3/4 2-3 Amendment No.101

(BSEP-2-35) 5.0 DESIGN FEATURES 5.1 SITE

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

LOW POPULATION ZONE 5.1.2 The low population zone shall be as shown in Figure 5.1.2-1.

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

For the purpose of effluent release calculations, the boundary for atmospheric releases is the SITE BOUNDARY and the boundary for liquid releases is the SITE BOUNDARY prior to dilution in the Atlantic Ocean.

5.2 CONTAINMENT CONFIGURATI0t1 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 5.2.2 The primary containment is designed and shall be maintained for:

a.

Maximum internal pressure 62 psig.

b.

Maximum internal temperature:

drywell 300* F Suppression chamber 200* F c.

Maximum external pressure 2 psig.

5.3 REACTOR CORE FUEL ASSEMBLIES 5.3.1 The reactor core shall contain 560 fuel assemblies.

The 8 x 8 fuel assemblies contain 63 fuel rods and the 8 x 8R, P8 x 8R, BP8 x 8R fuel assemblies contain 62 fuel ro'ds.

All fuel rods shall be clad with Zircaloy 2.

The nominal active fuel length of each fuel rod shall be 146 inches for 8 x 8 fuel assemblies and 150 inches for 8 x 8R, P8 x 8R, and BRUNSWICK - UNIT 2 5-1 Amendment No.101

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(BSEP-2-35)

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I DESIGN FEATURES 5.3 REACTOR CORE.

FUEL ASSEMBLIES (Continued)

BP8 x 8R fuel assemblies. The initial core loading shall have a maximum average enrichment of 2.47 weight percent U-235.

Reload fuel shall be'similar

'in physical design to the initial core loading and shall have a maximum

- average enrichment of 2.99 weight percent U-235.

CONTROL RD:1 ASSEMBLIES 2

5.3.2 The reactor core shall contain 137 control rod assemblies, each

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consisting of a cruciform array of stainless steel tubes containing approximately 143 inches of boron carbide, B C, powder or hafnium absorber 4

rods surrounded by a cruciform-shaped stainless steel sheath.

5.4 REACTOR COOLANT SYSTEM I

I DESIGN PRESSURE AND TEMPERATURE 5.4.1 The nuclear boiler and reactor recirculation system is designed and shall be maintained:

In accordance with the code requirements specified in Section 4.2 of a.

the FSAR, with allowance for normal degradation pursuant to the applicable Surveillance Requirements, b.

For a pressure of 1250 psig, and c.

For a temperature of 575 F.

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VOLUME 5.4.2 The total water and steam volume of the reactor vessel and recirculation system is approximately 3,670 cubic feet.

5.5 METdOROLOGICAL TOWER LOCATION 5.5.1 The meteorological tower shall be located as shown in ' Figure 5.1.1-1.

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Amendment No.101 BRUNSWICK ~- UNIT'2 5-4

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ML20093C920

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