ML20083R792
| ML20083R792 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 05/08/1995 |
| From: | Blom M, Karcher K CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML19344C916 | List: |
| References | |
| NF-2495.0015, NF-2495.0015-R, NF-2495.0015-R00, NUDOCS 9505300154 | |
| Download: ML20083R792 (17) | |
Text
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CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF 2495.0015 BIC10 Core Operating Limits Repon Page 1, Revision 0 BRUNSWICK UNIT 1, CYCLE 10 CORE OPERATING LIMITS REPORT MAY 1995 Controlled Copy i Prepared By: Michael A. Blom 1
Approved By:
[,
Date:
E' #f' Kenneth E. Karcher.
Manager BWR Fuel Analysis 9505300154 950515 PDR ADOCK 05000325 l
P PDR
W CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF-2495,0015 BlCIO Core Operating Limits Report Page 2, Revision 0 i
I LIST OF EFFECTIVE PAGES i
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CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF 2495.0015 BIC10 Core Operating Limits Report Page 3, Revision 0 INTRODUCTION AND
SUMMARY
nis report provides the values of the power distribution limits and control rod withdrawal block instrumentation setpoints for Bmnswick Unit 1, Cycle 10 as required by Technical Specification 6.9.3.1. The values of the Average Planar Linear Heat Generation Rate (APLHOR) limits, along with associated core flow and core power adjustment factors are provided as required by Technical Specification 6.9.3.1.a. He values of the Minimum Critical Power Ratio (MCPR) limits, along with associated core flow and core power adjustment factors are provided as required by Technical Specifications 6.9.3.1.b and 6.9.3.1.c. De control rod block upsce.le trip setpoints and allowable values are provided as required by Technical Specification 6.9.3.1.d.
Per Technical Specification 6.9.3.2 and 6.9.3.3, these values have been determfacd tving NRC-approved methodology and are established such that all applicable limits of the plant safety analysis are met.
Preparation of this report was performed in accordance with CP&L Nuclear Fuels Management & Safety Analysis (NFM&SA) Quality Assurance requirements as documented in Reference 1.
APLHGR LIMITS ne limiting APLHGR value for the most limiting lattice (excluding natural uranium) of each fuel type as a function of planar average exposure is given in Figures 1 through 6. These values were determined with the SAFER /GESTR LOCA methodology described in GESTAR.Il (Reference 2). Figures 1 through 6 are to be used when hand calculations are required as specified in Technical Specification 3.2.1.
De core flow and core power adjustments factors for use in Technical Specification 3.2.1 are presented in Fir,'2s 7 and 8. For any given flow / power state, the minimum of MAPLHGR(F) determined from Figure 7 and MAPLHGR(P) determined from Figure 8 is used to determine the goveming limit.
i
I CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF 2495.0015 B1C10 Core Operating Limits Report Page 4, Revision O '
MCPR LIMITS The ODYN OPTION A, ODYN OPTION B, and non-pressurization transient MCPR limits for use in Technical Specification 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 1.07 rpecified by Technical Specification 2.1.2.
s The core flow and core power adjustment factors for use in Technical Specification 3.2.2.1 are presented in Figures 9 and 10. For any given flow / power state, the maximum of MCPR(F) determined from Figure 9 and MCPR(P) determined from Figure 10 is used to determine the governing limit.
ROD BLOCK INSTRUMENTATION SETPOINTS The nominal trip serpoints and allowable values of the control rod withdrawal block instrumentation for use in Technical Specification 3.3.4 (Table 3.3.4-2) are presented in Table 2. These values were determined consistent with the bases of the ARTS program and the determination of MCPR limits with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR-II (Reference 2).
REFERENCE (s) 1)
CP&L Nuclear Fuels Management & Safety Analysis Quality Assurance File NF-2495.0015,
" Preparation of the Brunswick Unit 1. Cycle 10 (B1C10) Core Operating Limits Report (COLR),
Revision 0," (May 1995).
2)
NEDE-24011-P-A, " General Electric Standard Application for Reactor Fuel,' (latest approved version).
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CP&L Nuclear Fuels Mgmt & Sofety Analysis File: NF-2495.0015 81C10 Core Operating Umits Report Page 5 Revision 0 Figure 1 FUEL TYPE BD3238 (GE8X8EB)
AVERAGE PLANAR UNEAR HEAT GENERATION RATE APLHGR) LIMIT VERSUS AVERAGE PL(ANAR EXPOSURE 14 TI(IS FIG JRE IS Rt. t.nREO TC BY TT.CH4CLfiPECIFICATI)N 3.2,1 A.
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1.10 11 44 3
9 2.20 11 Ct
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27,56 11 37 33,07 10 65 38,58 9 94 44,09 9 23 49.60 8 52
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10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)
CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF-2495,0015 81C10 Core Operating umits Report Page 6 Revision 0 Figure 2 FUEL TYPE BD339A (GE8X8EB)
AVERAGE PLANAR LINEAR HEAT GENERATION RATE APLHGR) LIMIT VERSUS AVERAGE PL(ANAR EXPOSURE 14 THIS F1GJRE IS RET.RREC TC E T1:CH41CJL!iPECIFICATI3N 3.2,1 1
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10 15 20 25 30 35 40 45 50 55 60 l
AVERAGE PLANAR EXPOSURE (GWd/MT)
i CP&L Nuclear Fuels Mgmt & Safety Anotysis Rie: NF-2495.0015 B1C10 Core Operating umits Report Page 7. Revision 0 Figure 3 FUEL TYPE GE10-P8HXB322-11GZ-70M-150-T (GE8X8NB-3)'
AVERAGE PLANAR UNEAR HEAT GENERATION RATE (APLHGR) UMIT 1
l VERSUS AVERAGE PLANAR EXPOSURE 14
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THIS FIGJRE IS Rt. t.nEO TC B1 TI:CH4CL!PECIFICATDN 3.2,1 13 g
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5.51 11 50 E.61 11 67 7.72 11 92
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g 11,02 12 95 13,78 13 12 16.53 12 89 22,05 12 29 27.56 11 67 7
x,v f iv 9, 38.58 10 26 44,09 9 52 49,60 8 78 3
55.12 6 29 6
0 5
10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)
CP&L Nuclear Fuels Mgmt & Sofety Analysis File: NF-2495.0015 01C10 Core Operating Umits Report Page 8, Revision 0 Figure 4 FUEL TYPE GE10-P8HXB324-12GZ-70M-150-T (GE8X8NB-3)
AVERAGE PLANAR UNEAR HEAT VERSUS AVERAGE P(APLHGR) UMIT GENERATION RATE LANAR EXPOSURE 14 THIS RGJRE IS Rt. tME0 TC BV TI:CHilCAL !M:CIRCA1DN 3.2,1 13-I A
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10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)
CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF-2495.0015 B1C10 Core Operating umits Report Page 9. Revision 0 Figure 5 FUEL TYPE GE10-P8HXB320-11GZ-100M-150-T (GE8X8N8-3)
AVERAGE Pl.ANAR UNEAR HEAT GENERATION RATE (APLHGR) UMIT VERSUS AVERAGE PLANAR EXPOSURE 14 TitlS FIGJRE IS Rt. r.niE0 TC BV T1:CH41CJL IPECIFICAT13N 3.2,1 13-
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CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF-2495.0015 B1C10 Core Operating Limits Report Page 10, Revision 0 Figure 6 FUEL T(PE GE10-P8HXB346-12GZ-100M-150-T (GE8X8NB-3)
AVERAGE PLANAR LINEAR HEAT GENERATION RATE (APLHGR) UMIT VERSUS AVERAGE PLANAR EXPOSURE 14-THIS FIG JRE IS RETRREE TC BY T1:CH ilCALSPECIFICATI3N 3.2,1 f w 12
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5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)
6 CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF 2495.0015 BIC10 Core Operating Limits Report Page 11, Revision 0 Figure 7 Flow - Dependent MAPLHGR Limit, MAPLHGR (F) 1.0
~
oS >*
s g*
gs g N
+
g%
0.9 s
0.8
{
MAPLHGRtF).MAPFACp + MAPLHGR ;p 3
MAPLHGRSTD = 5TANDARD MAPLHGR UMITS M PFACp(Fle MINIMUM (1.0. ApW /100 + 8p)
C 0.7 W = % RATED CORE FLOW C
i O.6 MAXIMUM CORE FLOW
(% toted Ap Sp 102.5 0.8784 0.4881 0.5 107.0 0.8758 0.4574 112.0 0.8807 0.4214 117.0 0.8888 0.3828 0.4 I
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30 40 50 00 70 30 90 100 110 CORf FLOW (% retodi This figure is referred to by Technical Specification 3.2.1
CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF-2495.0015 BIC10 Core Operating Limits Report Ptge 12, Revision 0 Figure 8 Power - Dependent MAPLHGR Limit, MAPLHGR (P) 1.0 1 0.9 Y
I O.8
=
0 2e MAPLHGRtP) = MAPFACp'MAPLHGRSTD N
O.7 s 50% CORE FLOW MAPLHGRgio STANDARD MAPLHGR LIMITS FOR 25% >P. NO THERMAL LIMITS MONITORING
=
E REQUIRED l
o, FOR 25%sP 30%:
MAPFACp = 0.585 + 0.005224 (P-30%)
g FOR s 50% CORE FLOW l
MAPFACp = 0.433 + 0.005224 (P-30%)
I
> 50% CORE FLOW FOR >50% CORE FLOW 0.5 l
s FOR 30% sP.
g l
MAPFACP = 1.0 + 0.005224 (P-100%)
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0.4 20 25 30 40 50 60 70 80 90 100 POWER (% tated) i i
This figure is referred to by Technical Specification 3.2.1
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i CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF-2495.0015 BIC10 Core Operating Limits Report' Page 13, Revision 0 i
e 1
Table 1 MCPR Limits Non-oressurization Transient MCPR Limits GE8x8NB-3/GE8x8EB Exposure Range: BOC10 to EOC10 1.25 Pressurization Transient MCPR Limits GE8x8NB-3/GE8x8EB MCPR - Option A 1.31 Exposure Range: BOC10 to EOC10-2000 mwd /MT 1.32 Exposure Range: EOC10 2000 mwd /MT to EOC10 4
GE8x8NB-3/GE8x8EB MCPR - Option B Exposure Range: BOC10 to EOC10-2000 mwd /MT 1.24 Exposure Range: EOC10-2000 mwd /MT to EOC10 1.28 4
This table is referred to by Technical Specifications 3.2.2.1 and 3.2.2.2
CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF 2495.0015 B1C10 Core Operating Limits Report Page 14. Revision 0 Figure 9 Flow - Dependent MCPR Limit, MCPR (F)
FOR WC (% RATED CORE FLOW) <40%
FOR WC (% RATED CORE FLOW) k40%
MCPRIFl=(ApW /100 + 87)
MCPR(Fl= MAX 11.20. ApW /100+8pl C
C
'l1 + 0.0032 (40-W l3 C
7 MAX FLOW Ap By 117.0
- 0.632 1.80fr 112.0
- 0.602 1.747 1.6 107.0
- 0.586 1.697 102.5
- 0.571 1.655 1.5 MAXIMUM FLOW RATE = 117.0%
~
112.0%
107.0 %
102.5%
1.4 1.3 -
1.2 -
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l 20 30 40 50 60 70 80 90 100 110 120 CORE FLOW (% retodi This figure is referred to by Technical Specificatios 3.2.2.1
CP&L Nuclear Fuels Mgmt & Safety Analysis File: NF.2495.0015 B1C10 Core Operating Umits Report Page 15 Revision 0 I
l Figure 10 l
l Power - Dependent MCPR Limit, MCPR (P) l 9L_MC_M_ -
RATED MCPR MULTIPLIER (K )
P r
['-l-V' 2.3 e
l
> 50% CORE FLOW OPERATING UMIT MCRP (P) = K,
- OPERATING LIMIT MCPR (100) 2.2 -
l FO't P < 25%:
NO THERMAL LIMITS MONITORING REQUIRED 8' 2.1 l
l NO LIMITS SPECIFIED 0
g l
FOR 25% 5 P s P,,,u:(P,,,u = 30%)
l K, - MAXIMUM OF 1.481 OR K 2.0 y
I g s 50% CORE FLOW K, = (K,, + 0.02 (30%. P)] / OLMCPR(100) i a.
1.9 l
l K,, =
1.9 FOR s 50% CORE FLOW 2.2 FOR > 50% CORE FLOW
- 1.8 FOR 30% s P < 45%:
K, = 1.28 + 0.0134 (45% P) 1.5 FOR 45% s P < 60%:
==
K, = 1.15 + 0.00887 (60% P) l g
FOR 60% s P:
l g
K, = 1.0 + 0.00375 (100% P) 1.3 -
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l W i.i l
l 1
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1.0 I
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I t
f t
20 25 30 40 50 60 70 80 90 100 1
i P vPAss POWER (% rated) 8 This figure is referred to by Technical Specification 3.2.2.1 1
l 1
1
M File: NF 2495.0015 CP&L Nuclear Fuels'Mgmt & Safety Analysis Page 16, Revision 0 BlC10 Core Operating Limits Report Table 2 RBM System Setpoint Setpoint Trio Setooint Allowable Value 27.0 5 29.0 Lower Power Setpoint (LPSP')
62.0 s 64.0 Intermediate Power Setpoint(IPSP')
82.0 s 84.0
' High Power Setpoint (HPSP')
s 115.5 s 115.1 low Trip Setpoint (LTSP*)
Intermediate Trip Setpoint (ITSP6) s 109.3
$ 109.7 s 105.5 s 105.9 High Trip Serpoint (HTSP6) s 2.0 seconds s 2.0 seconds to Setpoints in percent of Rated %ermal Power.
115.1/125.0 of full Setpoints relative to a full scale reading of 125. For example, s 115.1 means s scale.
His table is referred to by Technical Specification 3.3.4 (Table 3.3.4 2) g
i ENCLOSURE 2 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NO.1 NRC DOCKET NO. 50-325 OPERATING LICENSE NO. DPR-71 TRANSMITTAL OF CORE OPERATING LIMITS REPORT, SUPPLEMENTAL RELOAD LICENSING REPORT, AND LOSS-OF-ACCIDENT ANALYSIS REPORT SUPPLEMENTAL RELOAD LICENSING REPORT 24A5159, REVISION O
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