ML20129J169

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Rev 0 to 1B21-0524-9501608, Brunswick Unit 1,Cycle 11,COLR Sept 1996
ML20129J169
Person / Time
Site: Brunswick Duke Energy icon.png
Issue date: 09/09/1996
From: Siphers J, Thomas R
CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML19353D955 List:
References
1B21-0524-95016, 1B21-524-95016, IB21-0524-9501608-00, IB21-524-9501608, NUDOCS 9611060157
Download: ML20129J169 (20)
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Text

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ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 DOCKET NO. 50-325 LICENSE NO. DPR-71 TRANSMITTAL OF CORE OPERATING LIMITS REPORT, SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-COOLANT-ACCIDENT ANALYSIS REPORT BRUNSWICK UNIT 1, CYCLE 11 CORE OPERATING LIMITS REPORT I

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9611060157 961101 PDR ADOCK 05000325 P PDR

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CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 BlCl1 Core Operating Limits Report Page 1, Revision 0 BRUNSWICK UNIT 1, CYCLE 11 CORE OPERATING LIMITS REPORT September 1996 Prepared By: M /2. Date: 9 M

/'

poger L. Thomas [

Approved B : Date: 9 *)(#

i i t j- John T. Siphers l Superintendent BWR Fuel Engineering b

K: CONTROL \DOCUMEh'nBIClICOLR\COLRBIi1 WPD

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I CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 B1Cl1 Core Operating Limits Report Page 2, Revision 0 l

l LIST OF EFFECTIVE PAGES f

Revision l-18 0 l

l K:TONTROL\DOCUMENnBICll\COLR\COLRBill WPD l

l CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 BlC11 Core Operating Limits Report Page 3 Revision 0 INTRODUCTION AND

SUMMARY

l This report provides the values of the power distribution limits and control rod withdrawal block instrumentation setpoints for Brunswick Unit 1, Cycle 11 as required by Technical Specification 6.9.3.1. The values of the Average Planar 1.inear Heat Generation Rate (APLHGR) limits, along with associated core flow and core power adjustment factors te provided as required by Technical Specification 6.9.3.1.a. He values of the Minimum Critical Power Ratio l (MCPR) limits, along with associated core flow and core power adjustment factors are provided as required by Technical l Specifications 6.9.3.1.b and 6.9.3.1.c. The control rod block upscale trip setpoints and allowable values are provided as required by Technical Specification 6.9.3.1.d.

l Per Technical Specifications 6.9.3.2 and 6.9.3.3, these values have been determined using NRC-approved methodology and are established such that all applicable limits of the plant safety analysis are met.

l Preparation of'his report was performed in accordance with Quality Assurance requirements as specified in Reference 1.

APLHGR LIMITS The limiting APLHGR value for the most limiting lattice (excluding natural uranium) of each fuel type as a function of l planar average exposure is given in Figures I through 8. These values were determined with the SAFER /GESTR LOCA methodology described in GESTAR II(Reference 2). Figures 1 through 8 are to be used when hand calculations are required as specified in Technical Specification 3.2.1.

The core flow and core power adjustment factors for use in Technical Specification 3.2.1 are presented in Figures 9 and 10. For any given flow / power state, the minimum of MAPLHGR(F) determined from Figure 9 and MAPLHGR(P) determined from Figure 10 is used to determine the governing limit.

MCPR LIMITS The ODYN ONION A, ODYN ONION B, and non-pressurization transient MCPR limits for use in Technical Specifications 3.2.2.1 and 3.2.2.2 for each fuel type as a function of cycle average exposure are given in Table 1. These values were determined with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR II (Reference 2) and are consistent with the Safety Limit MCPR of 1.10 specified by Technical Specification 2.1.2.

The core flow and core power adjustment factors for use in Technical Specification 3.2.2.1 are presented in Figures 11 and 12. For any given flow / power state, the maximum of MCPR(F) determined from Figure 1I and MCPR(P) determined from Figure 12 is used to determine the governing limit.

KnCONTROL\DOCU MENT \B 1 Cl I COLR\COLRB I I 1.WPD l

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CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 B1Cl1 Core Operating Limits Report Page 4, Revision 0 ROD BLOCK INSTRUMENTATION SETPOINTS The nominal trip setpoints and allowable values of the control rod withdrawal block instrumentation for use in Technical Specification 33.4 (Table 3.3.4-?) m presented in Table 2. These values were determined consistent with the bases of the ARTS program and the deternunation of MCPR limits with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR-II (Reference 2).

l REFERENCEfsi l) " Preparation of the B1Cl1 Core Operating Limits Report," Design Calculation 1B21-0524-9501608, Revision A, September 1996.

2) NEDE-240ll-P-A; " General Electric Standard Application for Reactor Fuel;" (latest approved version).

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CP&L Nuclear Fuels Mgmt & Safety Anolysis DCJ 1821-0524-9501608 81C11 Core Operating Umits Report Page 5. Revision O Figure 1 i

i FUEL TYPE BD323B (GE8X8EB)

AVERAGE PLANAR UNEAR HEAT GENERATION RATE (APLHGR) UMIT VERSUS AVERAGE PLANAR EXPOSURE j 14 TlflS FIGJRE IS RET.RRED TC BY ,

T1:CH ilCAL SPECIFICATI3N 3.2,1 1 13 1'

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11 .02 13 14 18,53 12 75

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27.56 11. 37 4

33,07 10 65 38,58 9 94 4 44,09 9 23 1

49,60 8 52 7 oo,a (

4 o un

6 0 5 10 15 20 25 30 35 40 45 50 55 60 g AVERAGE PLANAR EXPOSURE (GWd/MT)

CP&L Nuclear Fuels Wgmt & Sofety Analysis DCJ 1821-0524-9501608 81C11 Core Operating Umits Report Page 6, Revision 0 l

Figure 2 FUEL TYPE BD339A (GE8X8EB) i AVERAGE PLANAR LINEAR HEAT i GENERATION RATE (APLHGR) LIMIT 1

I VERSUS AVERAGE Pl.ANAR EXPOSURE 14 -

l TitlS FIGJRE IS RETRRED TC BY TI:CH 4CL UPECIFICATl* 3.2,1 13 D%

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N 12

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FERHISS10U:

FEGlim 0F b t C PEItATli*  %

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9  :.20 11 43 k 5.51 11 .02 12 02 12 88 13.76 12 84 18.53 12, 58 8 " "" " "

27.56 11 58 33.07 10 97 38.58 10.36 44.09 9 57 (

7 49.60 oo.u 8 79 o u

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i CP&L Nuclear Fuels Mgmt & Sofety Analysis OC# 1821-0524-9501608 81C11 Core Operating Umits Report Page 7 Revision 0 Figure 3 l i

FUEL T(PE GE10-P8HXB322-11GZ-70M-150-T (GE8X8NB-3) I AVERAGE PLANAR UNEAR HEAT j GENERATION RATE (APLHGR) UMIT '

VERSUS AVERAGE PLANAR EXPOSURE l 14 i

THIS RGJRE IS RE t.nWD TC B" TI:CHilCAL UPEClRCATIM 3.2,1 13 A l, I N N

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$ 5.51 d.61 11 50 j 11 67 1 7.72 11 92 8.82 12 26 I 5 l' 11 .02 13.78 12 13 95 12

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16.53 12 89 )

22.05 12 29 (  ;

27.56 11. 67 \

7 x.u i iu vo 38.58 10 26 44.09 9 52 49.60 8 78 a 55.12 6 29 6

0 5 10 15 20 25 30 35 40 45 50 55 60 i I

AVERAGE PLANAR EXPOSURE (GWd/MT)

1 1

I CP&L Nuclear Fuels Mgmt & Safety Anolysis DCJ 1B21-0524-9501608

) 81C11 Core Operating umits Report Page 8. Revision 0 f

1 Figure 4 j

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! FUEL TYPE GE10-P8HXB324-12GZ-70M-150-T (GE8X8NB-3)

AVERAGE PLANAR UNEAR HEAT i

GENERATION RATE (APLHGR) UMIT VERSUS AVERAGE PLANAR EXPOSURE 14 1 THIS FIGJRE IS RETRRED TC BV TI:CHilCAL. SPECIFICATim 3.2,1 13 f

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11 .02 13.78 12 96 13 17

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38.58 10 27 44.09 9 52 49.60 8 78 55.12 6 27 3 6

0 5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)

CP&L Nuclear Fuels Mgmt & Safety Analysis DCJ 1821-0524-9501608 B1C11 Core Operating Umits Report Page 9, Revision 0 Figure 5 FUEL TYPE GE10 -P8HXB320-11GZ-100M-150-T (GE8X8NB-3)

AVERAGE PLANAR UNEAR HEAT GENERATION RATE (APLHGR) UMIT VERSUS AVERAGE PLANAR EXPOSURE 14 THIS FlOJRE IS RETRRED TC BV T1:CHiiCJL!iPEdiFir.ATDN 3.2,1 13 12 J[ \'

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$ 5.51 E.61 11 53 11 68 7.72 11 84 8

11 .02 12 33 13.78 12 52 18.53 12 40 22.05 11 93 L 27.56 11. 44 \

7 x.u i iv ca 38.58 10 26 44.09 9 47 49.60 8 68 55.12 6 15 a 6

0 5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuels Mgmt & Safety Analysis DCJ 1821-0524-9501608 j

B1C11 Core Operating Umits Report Page 10, Revision 0 i

Figure 6 i

FUEL TYPE GE10-P8HXB346-12GZ-100M-150-T AVERAGE PLANAR LINEAR HEAT (GE8X8NB-3)

GENERATION RATE (APLHGR) LIMIT VERSUS AVERAGE PLANAR EXPOSURE 14

) TlilS FIGJRE IS RETRRED TC BV j TI:CH ilCJLliPECIFICATl* 3.2.1 13 4

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l CP&L Nuclear Fuels Mgmt & Sofety Analysis 1

B1C11 Core Operating Umits Report DCJ 1821-0524-9501608 Page 11, Revision 0 l

l Figure 7 FUEL TYPE GE13-P9DTB380-11G5.0A-100T-146-T (GE13) ,

AVERAGE PLANAR LINEAR HEAT GENERATION RATE l VERSUS AVERAGE P(APLHGR) LIMIT LANAR EXPOSURE 14 THIS FIGIJREl 15 REFE RRED TO Erf TECHMIC/L 3PtCifICATION,5.2 1 13 1

% I f  %

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AVERAGE PLANAR EXPOSURE (GWd/MT)

.. - . . - - . =- - .- .- ._ - - . - . - . . . . _-- . - -

CP&L Nuclear Fuels Mgmt & Sofety Anolysis DCJ 1821-0524-9501608 B1C11 Core Operating Umits Report Page 12 Revision 0 Figure 8 FUEL TYPE GE13-P9DTB380-10G5.0A-100T-146T (GE13) l AVERAGE PLANAR LINEAR HEAT i GENERATION RATE (APLHGR) LIMIT  !

VERSUS AVERAGE PLANAR EXPOSURE 14 THiSI%IJREl 11 ht.r ERRED TO Erf TECHPllCJL:iPE CIFICAT10N.5.2 1 13 gA 1 l

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. l CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524 9501608 BlCl1 Core Operating Limits Report Page 13, Revision 0 l

l 1 Figure 9 i

l Flow - Dependent MAPLHGR Limit, MAPLHGR (F) i  !

l 1.0 -

o

/s *.%*4*

0.9 -

g i 0.s -

l . MAPLHORtF)=MAPFACp MAPLHGRSTD I

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u MAPLHGMSTD = STANDARD MAPLHOR UMITS I MAPFACptF)= MINIMUM (1.0, ApW g,7 C /100+Spl l

' " We = % RATED CORE FLOW ANO Ar. Op ARE FUEL TYPE DEPENOENT CONSTANTS GIVEN BELOW:

0.6 -

MAXIMUM CORE FLOW

(% rated) Ap GF 0.5 - 102.s 0.e7s4 0.44e1 107.0 0.8754 0.4574 112.0 0.6807 0.4214 117.0 0.6884 0.3828 0.4 I I I I I I I 30 40 50 60 70 80 90 100 110 CORE FLOW (% rated) i 1

His figure is referred to by Technical Specification 3.2.1 {

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CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 B1C11 Core Operating Limits Report j Page 14, Revision 0  :

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Figure 10 Power - Depender1t MAPLHGR Limit, MAPLHGR (P) 1.0 I -

t 0.9 N

g 0.8 -

U I

j MArtHGR(P)=MAPFACP MApLwGRsTo 0.7 ~ s 50% CORE FLOW MAPLHGRgyp = STANDARD MAPLHGR LIMITS FOR 25% > P: NO THERMAL LIMITS MONITORING E REQUIRED l-0.6 - 1 FOR 25%sP<30%:

g MAPFACp = 0.545 + 0.005224 (P- 30%)

, FOR s 50% CORE FLOW 8

MAPFACp = 0.433 + 0.005224 (P- 30%)

1 > 50% CORE FLOW FOR > 50% CORE FLOW 0.5 -

3 g e FOR 30% s P:

l MAPPACp = 1.0 + 0.005224 (P- 100%)

1 0.4 I I I I I I I 20 25 30 40 50 60 70 80 90 100 POWER (% retodi r This figure is referred to by i Technical Specification 3.2.1 i

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1 CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 BICII Core Operating Limits Report Page 15, Revision 0 Table 1 MCPR Limits Non-pressurization Transient MCPR Limits l

FuelType MCPR Limit Exposure Range GE8x8NB-3/GE8x8EB 1.29 BOC11-EOC11 GE13 1.29 BOCl1-EOClI l I

l Pressurization Transient MCPR Limits MCPR Option A l FuelType MCPR Limit Exposure Range GE8x8NB-3/GE8x8EB 135 BOCil to EOCll-2205 mwd /MT l GE13 1.42 BOClI to EOCl1-2205 mwd /MT s ^Pg , E@" Igg @f; ' # 'gi,>, Eie ygel i

GE8x8NB-3/GE8x8EB 135 EOCll 2205 mwd /MT to EOC11 l l

GE13 1.48 EOCll-2205 mwd /MT to EOCil i

MCPR Option B FuelType MCPR Limit Exposure Range GE8x8NB-3/GE8x8EB 1.28 BOCil to EOCl12205 mwd /MT gel 3 137 BOCll to EOCl1-2205 mwd /MT

,j , QlfN , j, qllil%% , ^;; }$f lg L gy '

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GE8x8NB-3/GE8x8EB 131 ECCll 2205 mwd /MT to EOC11 GE13 1.40 EOCll 2205 mwd /MT to EOCll

'Ihis table is referred to by Technical Specifications 3.2.2.1 and 3.2.2.2 .

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CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 BlCll Core Operating Limits Report Page 16. Revision 0 Figure 11 l

l Flow - Dependent MCPR Limit, MCPR (F) l l

l FOR WC (% RATED CORE FLOW) <40% FOR WC (% RATED CORE FLOW) a40%

MCPRIP)=(ApWC /100+ 87) MCPR(F)= MAX (1.20 ApWC /100+ Spl

_ '*I1 + 0.0032 (40-WC ll MAX FLOW Ap Sp I

117.0 - 0.632 1.809 l 112.0 - 0.602 1.747 1.6 -

107.0 - 0.586 1.697 102.5 - 0.571 1.655

]

l 1.5 -

MAXIMUM FLOW RATE = 117.0%

C 112.0% 1

[ -107.0%

g 102.5%

i.4 -

l l

! 1.3 =

1.2 -

i,i l I I I I I I I l 20 33 40 50 60 70 80 90 100 110 120 CORE FLOW (% reteel This figure is referred to by

Technical Specification 3.2.2.1 t

l J K:\ CONTROL \DOCUMEN 1 Cl I COLR\COLRB 111.WPD

l CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-9501608 BICI1 Core Operating Limits Report Page 17. Revision 0 '

Figure 12 l

Power - Dependent MCPR Limit, MCPR (P) l l

r ; RATED MCPR MULTIPUER IKpl r I 2.3 -

$ l OPERATING UMIT MCPR(P) = K,

  • OPERATING UMIT MCPR(f 00) 250% CORE FLOW 2.2 -

l FOR P < 25%: NO THERMAL UMITS MONITORING REOUIRED l l NO UMITS SPECIFIED 2.1 -

l l FOR 25% s P s P  :(Pew = 30%)

E K, = MAX 1 MUM OF 1.481 OR Kg, 2.0 -

l l s 50% CORE FLOW Ka, = [4 + 0.02(30% P)] / OLMCPR(100)

a. 1.9 -  ! K,=

x 1.9 FOR s 50% CORE FLOW l l l 2.2 FOR > 50% CORE FLOW l p 1.8 - ' "# *

,> K, = 1.28 + 0.0134(45% P) 1.5 -

L, FOR 45% s P < 60%:

w l l K, = 1.15 + 0.00867(60% 4 P)

! FOR 60% s P:

= l l l 1.3 .

K, = 1.0 + 0.00375(100% P) g l

2 I 31.2 -

g l l l 8 I 1.1 -

l l l 8 I I I I I I l1.020 25 30 I

40 I

50 60 I I i 70 80 90 100 PBYPASS POWER (% retodl "Ihis figure is referred to by Technical Specification 3.2.2.1 KTONTROL\DOCUMEN11B I C I liCOLR\COLRB I i 1.WPD

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CP&L Nuclear Fuels Mgmt & Safety Analysis DC # 1B21-0524-950188 BICI1 Core Operating Lirnits Report Page 18 Revisior,0 l \

i Table 2 I

I l RBM System SetpointS I 1

1 Sg,tqtugt Trin Setnnint Allowable Value i i

Lower Power Setpoint (LPSP') 27.0 s 29.0 t

Intermediate l'ower Setpoint (IPSP') 62.0 s 64.0 High Power Setpoint (HPSP') 82.0 s 84.0 Low Trip Setpoint (LTSP') s 115.1 s 115.5 Intermediate Trip Setpoint (ITSP') s 109.3 s 109.7  !

High Trip Setpoint (HTSP') I s 105.5 s 105.9 to s 2.0 seconds s 2.0 seconds I 1

I

  • I Setpoints in percent of Rated Thermal Power.

Setpoints relative to a full scale reading of 125. For example, s 115.1 means s 115.1/125.0 of full scale.

This tableis referred to by Technical Specification 3.3.4 (Table 3.3.4-2) 9 l'

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4 ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 l DOCKE f NO. 50-325 LICENSE NO. DPR-71 TRANSMITTAL OF CORE OPERATING LIMITS REPORT, ,

SUPPLEMENTAL RELOAD LICENSING REPORT, AND j LOSS-OF-COOLANT-ACCIDENT ANALYSIS REPORT l i l

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SUPPLEMENTAL RELOAD LICENSING REPORT l FOR BRUNSWICK STEAM ELECTRIC PLANT UNIT 1  !

l RELOAD 10, CYCLE 11 ,

24A5376, REVISION O l

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