ML20101N485

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Nonproprietary Version of Suppl 1 to Power Uprate SAR for BSEP Units 1 & 2
ML20101N485
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 03/31/1996
From: Ball M, Helme R
CAROLINA POWER & LIGHT CO., GENERAL ELECTRIC CO.
To:
Shared Package
ML20013A238 List:
References
NEDO-32466-S01, NEDO-32466-S1, NUDOCS 9604080403
Download: ML20101N485 (15)
v • d • e


Text

I O GE Nuclear Energy 175 Curtner Avenue San Jose, CA 95125 NEDO-32466 Supplement 1 DRF B21-00565 Class 1 March 1996 Power Uprate Safety Analysis Report for Brunswick Steam Electric Plant Units 1 and 2 (Supplement 1)

Approved by: ( ~

l M.E.' Ball General Electric Company Approved by: -

R.E. Helme Carolina Power & Light Company

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i i l Purpose NEDC-32466P," Power Uprate Safety Analysis Report for Brunswick Steam Electric l

Plant Units 1 and 2," dated September 1995, summarized the evaluations performed tojustify uprating the licensed thermal power at Brunswick Steam Electric Plant by 5% to 2558 MWt.

l This supplement has been issued to revise specific information within NEDC-32466P. l Four changes are addressed in this supplement: l l 1. In NEDC-32466P, the short term containment analysis was based on initial drywell and l wetwell pressures of 0.75 psig. An analysis was performed with initial drywell and wetwell l pressures of 2.5 psig, and this supplement updates NEDC-32466P with the analysis results.

l See pages 4-2 and 4-11 through 4-13 of this supplement.

l 2. In NEDC-32466P, the ATWS analysis was based on a high pressure analytical limit of 1135 l l

psia. An analysis was performed with an analytical limit of 1185 psia, and this supplement  !

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updates NEDC-32466P with the analysis results. See pages 5-6,9-3 and 9-17 of this ,

i supplement. ,

i 3. Table 9-3 indicated that values of zero were assumed for certain parameters (MSIV leakage l rate, SGTS bypass fraction, and exfiltration duration during secondary containment i

( drawdown), when those parameters are actually not applicable to the Brunswick licensing l l basis contained in the Updated Final Safety Analysis Report. Table 9-3 has been revised to reflect this. See page 9-7 of this supplement.

4. As a result of the above changes, the Table of Contents required revision, and other pages ,

were renumbered to accommodate the new text. No other changes were made to these pages, j l See pages viii, xi/xii,9-3a and 9-4.

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This supplement only contains those pages from NEDC-32466P that change. Changes to the text  !

are underlined. l l

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NEDO-32466 Suppl:m:nt 1 IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT Please Read Carefully The only undertakings of the General Electric Company (GE) respecting information in this docenent are contained in the contract between Carolina Power and Light (CP&L) and GE, under Contract ZM70020000, effective October 31,1994, as amended to the date of transmittal of this document, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than CP&L, or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, GE makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document, or that its use may not infringe privately owned rights.

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NEDO-32466 SuppI; ment 1 1_

n 4

TABLE OF CONTENTS 4 (Continued)

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Page 8.5.4 Offsite Doses (Normal Operation) ................................................... .... 8-5 4

I l 9.0 REACTOR SAFETY PERFORMANCE EVALUATIONS................................ 9-1 d

9.1 Reac to r Transi ents .. . . . . . . . . .. . . .. . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .9-1 .................

9.2 Desi gn B asis Accidents.... ... ........ .... .... ...... ..... .. ........ . .. . .. .. .... . .... . . .. . . .. .. . . . . .. .. .. . . .. ... . 92 9.3 S pecial Events . . . . . . . . . . . . . . . .. .. . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . .9-3 9.3.1 Anticipated Transients Without Scram (ATWS)................................... 9-3 9.3.2 Station B lackout (S B0) . . . .. .. ........... .. .. ...... . .. ... ....... ... . .. . . ... . . . . . ....... .... . ... B 9.3.3 AppendixR........................................................................................... B 9.4 References............................................................................................................. 9-4 10.0 ADDITIONAL ASPECTS OF POWER UPRATE.............................................. 10-1 1

10.1 High Energy Line Break (HELB) ......................................................................... 10-1 l 10.1.1 Temperature, Pressure and Humidity Profiles ....................................... 10-1 l 10.1.1.1 Main Steam System Line Break .......................................... 10-1 l 10.1.1.2 High Pressure ECCS Line Break ......................................... 10-1  !

10.1.1.3 Reactor Core Isolation Cooling System Line Break............ 10-2 l

10.1.1.4 Reactor Water Cleanup System Line Breaks....................... 10-2 l 10.1.1.5 Control Rod Drive System Line Break.............................. . 10-2  !

10.1.2 Pipe Whip and Jet Impingement ............................................................ 10-2  :

10.1.3 Moderate Energy Line Break (MELB) .................................................. 10-2 10.2 Environmental Qualification (EQ)........................................................................ 10-3 10.2.1 EQ o f Electrical Equipment .... .. ..... .. .. . . ..... . . ..... . ..... . ..... . .... ..... ... . ..... . . . . . . . 10-3 10.2.1.1 I nside Containment ....... .. ........ . .... ...... . . . ... .. . ... . .. . . . .. . ... . .. . .. . .. . 10-3 10.2.1.2 Outside Containment ... ...... .. ..... .... . ... . .. ............ . .. . . ..... . .... . .. .. 10-3 10.2.2 EQ of Mechanical Equipment with Non-Metallic Components...... ..... 10-4 10.2.3 Mechanical Component Design Qualification............... ....................... 10-4 10.3 Re qui re d Te sti n g . . . . . . . . . . . . . .. .. . . . . .. . . . ... . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-4 10.4 Shutdown and Refueling Requirement ........................................ ........................ 10-5 10.5 Operat o r Trai ni n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . .10-5 10.6 PlantLife............................................................................................................. 10-5 VIII...

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NEDO-32466 Suppbment 1 TABLES Table Title Page 1-1 G lo s sary o f Tenns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . 1-7.......

1-2 Original and Uprated Plant Operating Conditions........... .............................. 1-10 3-1 RIPDs for Normal Conditions ............................. ...... ............ ....... ................. 3-15 3-2 RIP Ds for Upset Conditions .. . .... .. .. . ...... . . .. . . .. . . ....... .. .. .... . .. ... .. .. . .. .. . .. .. .. .. .. . . .... 3-16 3-3 RIPDs for Faulted Conditions........................................................ ................. 3-17 3-4 Summary of Maximum Stresses and Locations for Reactor Internals at 105% Power Uprate ........................... ...... ....................... .... . ...... 3-18 3-5 Fatigue Usage Factors of Limiting Components ............... ......... . ........... .... 3-19 4-1 Containment Performance Results...... ............. ....... .. ... .... ........................ 4-11 5-1 Analytical Limits for Setpoints ........ ............ ................................................... 5-6 6-1 Uprated Plam Slectrical Design Characteristics ............................................. 6-10 6-2 Fuel Pool Co o l i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .6-11 6-3 Effluent Discharge Comparison...................... ... ............. .......... ......... ........ 6-12 9-1 Parameters Used for Transient Analysis ......................................... ............... 9-5 9-2 Transient Analysis Results for Power Uprate...... ................... ...................... 9-6 9-3 Assumptions for Loss-of-coolant Accident ................................... ................ 9-7 9-4 Assumptions for LOCA Control Room Dose................................................. 9-9 9-5 Assumptions for Main Steam Line Break Accident ...... ................................ 9-10 9-6 Assumptions for Control Rod Drop Accident ................................................ 9-11 9-7 Assumptions for Fuel Handling Accident......... ............................................. 9-12 9-8 LOCA Radiological Consequences ..................... ...... . ... . ................... ....... 9-13 9-9 MS LB A Radiological Consequences ............................................................. 9-14 9-10 FHA Radiological Consequences .. ........................ .... .................. ..... ......... 9-15 9-11 CRDA Radiological Consequences . .. ................... .................. ............... .... 9-16 9-12 MSIV Closure Results - ATWS Peak Vessel Pressure Evaluation . 9-16a 10-1 Additional Aspects of Power Uprate ..... .............. . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10-7 11-1 Technical Specifications Affected By Power Uprate.............................. ....... 11-4 xi/xii

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1 NEDO-32466 Suppl:mont 1 1

l 4.1.1 Containment Pressure and Temperature Response l

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NEDO-32466 Suppl:m:nt 1 Table 4-1 CONTAINMENT PERFORMANCE RESULTS Parameter Current Rated Power Uprated Power Limit Peak Drywell 49.4 (UFSAR) 40d* 62 Pressure (psig) 36.8*

Peak Bulk Pool 205 (UFSAR) 201 220 Temperature ('F) 197 (current method)

Drywell 283* 283 340**

Temperature ( F)

  • Mark I LTP method.
  • The drywell temperature design limit is reported in the UFSAR as 281*F. However, drywell liner stresses and equipment in the drywell can actually withstand temperatures of up to 340 F, i

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.. NEDO-32466 Supplement 1 l

t ORYELL PRESSURE-PSIG

  • E TELL PRESSURE-PSIG i

60.

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10.

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20.

w E  :

0. 'I'

O. 12. 24 56. 48.

TIME-SECONDS Figure 4-1. DBA-LOCA Short-Term Drywelland Wetwell Pressure Response (102% of UpratedPower) l 4-12

NEDO-32466 Suppl 1 ment 1 DRYELL TE W .~0E3.F E TELL TEW .-K).F 450.

500. .i i , , i e

8, k * '

ig 150. _

5 -

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O. 10. 20. 50. 40.

TIME-SECONDS Figure 4-2. DBA-LOCA Short-Term Drywell and Wetwell Temperature Response (102% of UpratedPower)

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NEDO-32466 Suppl: ment 1 Table 5-1 ANALYTICAL LIMITS FOR SETPOINTS Analytical Limits Parameter Current Power Uprate APRM Basis Calibrated to 2436 MWt Calibrated to 2558 MWt (rated power) (rated power)

APRM Simulated Thermal Power 117 % 117 %

Scram Clampedm APRM Neutron Flux Scram 121 % 121 % l Vessel High Pressure Scram 1086 psia 1111 psia High Pressure ATWS RPT 1135 psia 1181 psia Safety / Relief Valve Settings (psig)* 1116 1164 1126 1174 1136 1185 Turbine First-Stage Scram Bypass 30.0 % 30.0 % l Pressure (% Power)  !

Main Steam High Flow 140 % 140 %

Main Steam High Flow (Unit 2 only 40 % 40 %

Modes 2 and 3)

NOTE:

(1) Analytical Basis - No credit taken for flow bias.

(2) Includes tolerances.

5-6

. NEDO-32466 S:rppl: ment 1 9.3 Special Events 9.3.1 As.ticipated Transients Without Scram (ATWS)

The plant's parameters for power unrate meet the four criteria. except that the ATWS high oressure setooint will be increased by 50 psi.

Per Reference 2. the Inadvertent MSIV Closure (MSIVC) event is the limiting ATWS event. The most limiting initial flow condition is the lowest core flow allowed at 100% power.

i.e.. the point on the highest flow control line. For the Brunswick power uprate analysis. this is the point at 100% of the unrate power (105% of 2436 MWt) and 81% core flow. Therefore. RPV integrity was analyzed for the MSIVC event. with the initial conditions noted above. to determine the effects of the ATWS high pressure setooint increase. The key inouts to the analysis are:

e Reactor oower at 2558 MWt: i e Reactor dome pressure at 1048 osia:

. SRV onening setooints increased by 25 psi. with the tolerance increased to 3%:

. ATWS high oressure anaMical limit increased by 50 osi to 1170 psig: and

. One SRV assumed out-of-service (the SRV with the lowest setooint).

The results of this analysis are shown in Table 9-12 and demonstrate that the calculated geak vessel oressure is below the ASME limit of 1500 nsig.

Based on the generic analysis in Reference 2 and the Brunswick soecific analysis described above. nower unrate will not result in any ATWS accentance criteria from being exceeded.

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.. NEDO-32466 Supplem1nt 1 9.3.2 Station Blackout (SBO)

Plant response and coping capabilities for a station blackout (SBO) event are impacted slightly by operation at the uprated power level due to the increase in the operating temperature  !

of the primary coolant system, increase in the decay heat, and increase in the main steam safety relief valve setpoints. There are no changes to the systems and equipment used to respond to an ]

SBO, nor is the required coping time changed. j l

The Brunswick response to a postulated SBO is to utilize the HPCI System with DC augmentation from the other unit to cope for four hours. Emergency diesel-generator and station f battery performance are adequate for response to an SBO with power uprate operation.

The following areas contain equipment necessary to mitigate the station blackout event: i Control, Diesel-Generator Basement and 480V EBUS Switchgear Rooms j HPCI and RCIC Equipment Room l ECCS Pipe Tunnel j

  • Containment  !

The temperature increases in the Control, Diesel-Generator Basement and 480V EBUS  !

Switchgear Rooms are not affected by power uprate. The HPCI and RCIC equipment room  ;

temperatures and the ECCS pipe tunnel temperature will increase; however, significant margin exists to operability limits so that the operability of equipment in these locations is not affected.

  • Suppression pool temperature increases about 4 F and containment pressure about 2 psi, but the  !

l increases are small enough to not affect equipment operability. I Also, the condensate water requirement increases less than 3%; however, the current I condensate storage tank design ensures that adequate water volume is available.  !

l 9.3.3 Appendix R i 1

l Brunswick calculations associated with 10CFR50 Appendix R have been reviewed for l

l the 105% power uprate condition. The conclusions of the review are that operation of the plants l at the 105% power level will not affect the ability of the safe shutdown systems to perform their l 1

intended function, and the minimum systems and equipment required for safe shutdown at power uprate do not change. l 1

h 9.~3Q I

, NEDO-32466 Suppl 1 ment 1 9.4 References i I. GE Nuclear Energy, Generic Guidelines For General Electric Boiling Water Reactor Power Uprate, Licensing Topical Report NEDO-31897, Class I (Non-proprietary), February 1992; and NEDC-31897P-A, Class III (Proprietary), May 1992.

2. GE Nuclear Energy, Generic Evaluations ofGeneral Electric Boiling Water Reactor Power Uprate, Licensing Topical Report NEDC-31984P, Class III (Proprietary), July 1991;

! NEDO-31984, Class I (Non-proprietary), March 1992; and Supplements.

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NEDO-32466 SuppI2m:nt 1 Table 9-3 ASSUMPTIONS FOR LOSS-OF-COOLANT ACCIDENT Power (MWt) 2558 Power Multiplier Factor 1.02 Initial Inventory Fractions in Containment Atmosphere (%)

Noble gases 100 lodines 25 Primary Containment Leak Rate (%/ day) 0.5 MSIV Leakage Rate (%/ day) N/.A  !

Fraction of Containment Leakage Which  ;

Bypasses SGTS (%) N/A  :

Holdup in Secondary Containment None .

Duration of Exfiltration During  !

Secondary Containment Drawdown (min.) N/A  !

SGTS Iodine Filter Efficiency (%) 99 l ECCS Leakage in Secondary Containment None l Release Height (m) 100 i

Site Boundary Distance (m)3 914,3219 l X/Q at EA Boundary (sec/m ) l 0-2 hours 2.0E-5 j 3

X/Q at LPZ (sec/m )

0-8 hours 8.8E-6 j 8-24 hours 3.8E-6 l 24-96 hours 1.1 E-6

  • 96-720 hours 3.5E-7 Thyroid Inhalation DCF (rem /Ci)

I-131 1.49E+6 I-132 1.43 E+4 1-133 2.69E+5 I-134 3.73E+3 I-135 5.60E+4

  • RG 1.3 value is more conservative than the UFSAR value.

9-7 l

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.. NEDO-32466 SuppI:m:nt 1 i

Table 9-12 l

MSIV CLOSURE RESULTS -

ATWS PEAK VESSEL PRESSURE EVALUATION [

Parameter Cait 1 2 3 Initial Core Flow (% Rated) JDQ 31 81 Number of Functional SRVs 11 11 1D l Peak Neutron Flux (% Rated) 324 112 312 l Peak Ave. Heat Flux (% Rated) 152 l_43 142 I

l Peak Vessel Pressure (nsiM 131fi 1102 14.21

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Accentance Criteria (psis 15.00 110D 15HQ l

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