ML20236T192
| ML20236T192 | |
| Person / Time | |
|---|---|
| Site: | Brunswick  |
| Issue date: | 07/01/1998 |
| From: | Dresser T, Siphers J CAROLINA POWER & LIGHT CO. |
| To: | |
| Shared Package | |
| ML20135G046 | List: |
| References | |
| 1B21-0537, 1B21-0537-R01, 1B21-537, 1B21-537-R1, NUDOCS 9807280020 | |
| Download: ML20236T192 (34) | |
Text
{{#Wiki_filter:- - _ ___ _ - ___ _ _ ____ ENCLOSURE 1 BRUNSWICK STEAM ELECTRIC PLANT, UNIT NOS.1 AND 2 DOCKET NOS. 50-325 AND 50-324 LICENSE NOS. DPR-71 AND DPR-62 TRANSMITTAL OF C0 IRE OPERATING LIMITS REPORTS, SUPPLEMENTAL RELOAD LICENSING REPORT, AND , LOSS-OF-CGOLANT-ACCIDENT ANALYSIS REPORT -{ l i i BRUNSWICK UNIT 1, CYCLE 12 CORE OPERATING LIMITS REPORT JULY 1998
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t [- CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 1, Revision 1 l l BRUNSWICK UNIT 1, CYCLE 12 CORE OPERATING LIMITS REPORT July 1998 1 Prepared By. W -
- Date: 6/30/98 l Thomas M. Dresser Approved B . Date: 7/1/98 1
John 1TSipher[ Superintendent BWR Fuel Engineering l t
d CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report l Page 2, Revision 1 LIST OF EFFECTIVE PAGES
~ Page(s) Revision 1 33 i l
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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No, 1821-0537 B1C12 Core Operating Limits Report Page 3. Revision 1 TABLE OF CONTENTS Subject Page Number Cover..................................................................................................................I List o f E ffective Pa ges.. . . . .. . . . .. . . .. ... . . . .. . . .. .. . ... . .. . .. . .. . . .. . . . . . . . . .. ..... . ...... . .. . .. . . .. . .... . . .. . .. 2 Table o f Co nten ts. . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . .. . . . . . . List o f Tables / List o f Fig ures.. . .. ... ....... . . . ... ... .... . . ...... . .. .. .. . . .. . ........ . .... .. . . . . ..... .. . .. ... . . ... . .. 4
- I introduction & S ummary ... .... . ... .. .. . ... .. ..... ...... ..... . .... .... . . . . . . .. ..... . . .. .. . . . .. ... . ... . . ... . . . . ... ... 6 Single Loop Operation... ............................ ................................................................7 Inoperable Main Turbine Bypass System ........... ..... .......................... ............. ... .............. 7 A PL H G R L imi ts . . . .. . . . . . . . . . . . . . . . .. . . ... . . . . . . .. . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . ... . . . . .
MCPRLimits..............................................................................................................8 RBM Rod Block Instrumentation Setpoints ........... ........ ..... ................................. . .......... 8 APRM Flow Biased Simulated Thermal Power-High.. ........... ............................ ... .......... 8 APRM Control Rod Block Upscale (Flow Biased) ............ .......... .................. ... .... ... ...... 9 THI E l A Stability Sol ution .. ... ....... ..... .... . .... .. .. . . . . . . .. . .... ... .. . .. .. . .. .. .... . . . ... ... .... . ... ... . ...... . . 9 References..............................................................................................................10 i i l. i
-s,-a +-+
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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 4. Revision 1 LIST OF TABLES Table Title Page Number Table 1: M C P R L i mi ts . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 Table 2: RB M System Setpoints. ....... . ..... . . .. .. ........ ... .. . .. .. . . ... . . . . .... . . .. . . ... . 12 Table 3: Al igned Drive Flo w. . . . .. . . . . ... . . . . . . . ...... . .. . . . . .. . .. .. . . . . . . . . . . .. . . . ... .. .. . . . . . ... . . . . . .. . .. .. . ! 3 1 4 LIST OF FIGURES i 1 Figure Title or Description Page Number
- Figure 1: APLHGR Limit Versus Average Planar Exposure........ ........... .............. .. ..... 14 Figure 2: APLHGR Limit Versus Average Planar Exposure........... ........................... .... .15 Figure 3: APLHGR Limit Versus Average Planar Exposure........... ............, ................. ..16
- Figure 4
- APLHGR Limit Versus Average Planar Exposure.. ............ ........... ... ..............17 Figure 5: APLHGR L,imit Versus Average Planar Exposure..... ... ... .... ..........................I 8 1
l Figure 6: APLHGR Limit Versus Average Planar Exposure.... ........ .... .. ... .......... .......19 ! 1 i ! Figure 7: APLHGR Limit Versus Average Planar Exposure.......... .. .. .. . ........ ..... ... .. 20 1 Figure 8: APLHGR Limit Versus Average Planar Exposure.. .. .. .. ... ... ... . ... . ........... 21 ; Figure 9: MAP L H G R( F) ..... . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ... . . . . . . . . . . . . . . . . . . . . .. . . . . . 2 2 Figure 10: MAP L G H R( P ) . . . .. .. . . . . . . . . . . . . . .. . . . . .. . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . .. . . . . . . . . . . . . .. . . . . . . . . . . . . . 2 3 Figure 11: MCPR(F).........................................................................................................24 Figure 12: MCPR(P).....................................................................................................25
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Umits Report Page 5, Revision 1 LIST OF FIGURES / LIST OF TABLES - CONTINUED Figure Title or Description Page Number Figure 13: PowerNiow Map Stability Regions: Norrnal T rw, Non-Setup (B 1 C12)... ... ................ .. .......................... . . .. . 26
. Figure 14: PowerNlow Map Stability Regions:
Normal T rw, Setup (B 1 C 12)...... ........ .. ........ ... ... ..... .......... .... ...... . .. 27 Figure 15: PowerNI~o w Map Stability Regions: Reduced T rw, Non-Setup (B 1 C 12)...... .. .... . ...... ........ ..... ...... ... ... .... 28 Figure 16: PowerNiow Map Stability Regions: Reduced T rw, Setup (B 1 C 12)...... .... .... ................ ................. .. .............. 29 Figure 17: EI A Setpoint Allowable Values versus Aligned Drive Flow: Normal T rw, Non-Setup (B 1 C 12)................ .................. ......................... ... 3 0 Figure 18: EI A Setpoint Allowable Values versus Aligned Drive Flow: Normal T rw, Setup (B 1 C 12)........................... ..................... ....... ....... . ..... 31 Figure 19: EI A Setpoint Allowable Values versus Aligned Drive Flow: Reduced T rw, Non-Setup (B 1 C 12).. . .............. ...... ........ .................... ..... 3 2 Figure 20: El A Setpoint Allowable Values versus Aligned Drive Flow: Reduced T rw, Setup (B 1 C 12).. .......... .............................. ..... .... .......... .... 3 3 l l-I I l i' l ! l L !
I CP&L Nuclear Fuels Mgmt. & Safety Analysis B1C12 Core Operating Limits Report Design Calc. No. 1821-0537 Page 6, Revision 1 INTRODUCTION AND
SUMMARY
This report specifies for Brunswick Unit 1, Cycle 12 ythe (THD Thermal Hydraulic Technical Specifications (ITS) 3.2.3 and 3.3.1.3, an oved anual Specification (TRMS) 3.2. This report also specifies the Exclusion Region Th onitored Region, Restricted Region and Exclusion Region are indicated on the fa er core flow map. This report specifies the " Setup"and "Non Setup"owscram Biased values of the AP Simulated Thermal Power High Allowable Valuered("byFlow ITS Biased Scram") values of the APRM Flow Biased - ock") Upscale ock as Allowab required by the TRMS 3.3. The scram and rod block er r vevalues are indicated flow domain in which the Flow Control Trip Reference cards operate. recirculation loop in operation as required one by ITS L Unit 1, Cycle 12 the limits to support operation withypass as required by ITS 3.7.6. an inoperable runswick System Main Turbin This report provides the values of the power distribution limits and control rod rawal block instrumentation setpoints for Brunswick Unit 1, Cycle ... 12 as required by ITS e values flow and core power adjustment factors , are provided a a e core the Minimum Critical Power Ratio (MCPR) limits, .along . .a.l. The values with of associat d power adjustment factors are provided as required byeITS core5.6.5.a.2. flow and core The Allowable Values for Function 2.b, APRM Flowower Biased
-High Simulated Thermal allowable values are provided as requiredn by s and ITS 5.6.5.a.4.are p Per ITS 5.6.5.b and 5.6.5.c, these values have been e determined using NRC methodology and are established such that all applicable m et.
are limits of the pla Preparation of this report was performed in accordance specified in Reference 1. rements as with Quality Assu
l f 1 I CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B'iC12 Core Operating Limits Report Page 7 Revision 1 j SINGLE LOOP OPERATION Brunswick Unit 1, Cycle 12 may operate over the entire MEOD range with Single recirculation Loop Operation (SLO) as permitted by ITS 3.4.1 with applicable limits specified in the COLR for ITS LCO's 3.2.1,3.2.2, and 3.3.1.1. The applicable limits are: LCO 3.2.1 Average Planar Linear Heat Generation Rate (APLHGR) Limits: per Reference 1, the Figuies 9 and 10 described in the APLHGR LIMITS section below include a SLO limitation of 0.8 on the MAPHGR(F) and MAPLHGR(P) multipliers. LCO 3.2.2 Minimum Critical Power Ratio (MCPR) Limits: per Reference 1, the MCPR limits described below and presented in Table I and Figures 11 and 12 apply to SLO { without modification. LCO 3.3.1.1 Reactor Protection System Instrumentation Function 2.b (Average Power Range Monitors Flow Biased Simulated Thermal Power- High) Allowable Value: per Reference 1, these limits - described below under THI EI A STABILITY SOLUTION - apply to SLO without modification. INOPERABLE MAIN TURBINE BYPASS SYSTEM Brunswick Unit 1, Cycle 12 may operate with an inoperable Main Turbine Bypass System in accordance with ITS 3.7.6 with applicable limits specified in the COLR for ITS LCO 3.2.1 and 3.2.2. One bypass valve inoperable renders the System inoperable, although the Turbine Bypass Out-of-Service (TBPOOS) analysis supports operation with all bypass valves inoperable for the l entire MEOD range and up to 110 F rated equivalent feedwater temperature reduction. The system response time assumed by the safety analyses from event initiation to start of bypass valve opening is 0.10 seconds, with 80% bypass flow is achieved in 0.30 seconds. The applicable limits are: I LCO 3.2.1 Average Planar Linear Heat Generation Rate (APLHGR) Limits: in accordance with Reference 1, TBPOOS requires a reduction in the MAPLGHR(P) limits between 25% and 30% power. This reduction is plotted on Figure 10. i LCO 3.2.2 Minimum Critical Power Ratio (MCPR) Limits: in accordance with Reference 1, TBPOOS requires a increase in the MCPR(P) multiplier between 25% and 30% power. This increase is plotted on Figure 12. TBPOOS also requires increased MCPR limits, included in Table 1.
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 8. Revision 1 APLHGR LIMITS
- The 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 I through 8. These values were determined with the SAFER /GESTR LOCA methodology described in GESTAR-II (Reference 2). Figures I through 8 are to be used only when hand calculations are required as specified in the bases for ITS 't,2.1 Normal monitoring of the APLHGR limits with POWERPLEX uses the complete set oflattices for each fuel type provided in Reference 3.
The core flow and core power adjustment factors for use in ITS 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 fro.a Figure 10 is used to determine the governing limit. MCPR LIMITS l The ODYN OPTION A, ODYN OPTION B, and non-pressurization tran ient MCPR limits for use in ITS 3.2.2 for each fuel type as a function of cycle average exposure are given in Table 1. These values were detennined 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 1.09 specified by ITS 2.1.1.2. The core flow and core power adjustment factors for use in ITS 3.2.2 are presented in Figures 11 and 12. For any given flow / power state, the maximum of MCPR(F) determined from Figure 11 and MCPR(P) determined from Figure 12 is used to determine the goveming limit. , i All MCPR limits presented in Table 1, Figure 11 and Figure 12 apply to two-recirculation pump operation and SLO without modification. RBM ROD BLOCK INSTRUMENTATION SETPOINTS
, The nominal trip setpoints and allowable values of the control rod withdrawal block instrumentation for use in ITS 3.3.2.1 (Table 3.3.2.1-1) 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
- l. GESTAR-II (Reference 2). . <
l APRM FLOW BIASED SIMULATED THERMAL POWER-HIGH The APRM Flow Biased Simulated Thermal Power-High setpoint as referenced by ITS 3.3.1.1 and TRMS 3.2 is described in the TH1 El A Stability Solution section below. l L_________________.______..__ _
l CP&L Nuclear Fuels Mgrnt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Lirnsts Report Page 9, Revision 1 I APRM CONTROL ROD BLOCK UPSCALE (FLOW BIASED) The APRM Control Rod Block Upscale (Flow Biased) setpoint as referenced in TRMS 3.3 is described in the THI EI A Stability Solution section below. l THI EI A STABILITY SOLUTION Implementation of the THI El A Stability Solution involves exclusion from certain areas of the power / flow map and specific restrictions for operating in other areas. Stability Regions generally protect the high power / low flow regions of the power / flow map and include: an Exclusion Region from which plant operation is excluded by the APRM Flow-biased Scram; a Restricted Region in which operation is restricted to conditions where the Boiling Boundary exists at least four feet above the bottom of the core (Fraction of Core Boiling Boundary or FCBB shall be s 1.0, per ITS 3.2.3); and a Monitored Region in which operation is allowed only if the potential for a developing THI event is monitored by the Period Based Detection System (PBDS operable, per ITS 3.3.1.3). Flow Control Trip Reference (FCTR) cards provide the ITS 3.3.1.1 automatic APRM Flow-biased Scram and the TRMS 3.2 Restricted Region Entry Alarm (RREA) using the TRMS 3.3 Flow-biased Rod Block setpoint. Because the flow signal for these automatic functions is based on drive flow and not core flow, the COLR Flow-biased Scram and Rod Block limits Allowable Values are included as functions of drive flow (Figures 17-20). These curves allow quantification of Technical Specification compliance, once the drive flow input is aligned in accordance with Table 3. However, the Stability Regions defined by these limits as well as the Monitored Region (which is not incorporated into the FCTR card) are analyzed by El A in terms of power versus core flow. Because the plant Operators are accustomed to monitoring plant j condition in terms of core flow, the COLR provides the Stability Regions on the power /(core) ! flow map for Operations in recognition of human performance factors. Day-to-day and time- I pressured responses by the Operator should use the power / flow map stability regions (Figures 13-16). These Figures define the Monitored and Restricted Regions for compliance with ITS 3.2.3 and 3.3.1.3 and TRMS 3.2. Core flow nominal trip setpoint values on Figures 13-16 correspond to the nominal trip setpoint values translated into drive flow and installed in the FCTR cards. As required by TRMS 3.2, should neither channel of the RREA be operable to protect against unintentional entry into the Restricted Region, the plant must be operated below the flow control line associated with the lowest thermal power in the Restricted Region, or continuous monitoring i of thermal power and core flow maintained by a qualified individual. l l The EI A Stability Solution provides for distinct Flow-biased Scram and Rod Block setpoints for
" Setup",as well as normal or "non-Setup" conditions (the Monitored Region boundary is unchanged by Setup). "Non-Setup" setpoints (Figures 13,15,17,19) enforce the normal l
L_____-______----____ _ _ - . . -
CP&L Nuclea Fuels Mgmt. & Safety Analysis Design Calc. No.1B21-0537 B1C12 Core Operating Limits Report - Page 10, Revision 1 Exclusion add Restricted Regions described above. Setup setpoints (Figures 14,16,18,20) are to be used only when the FCBB s 1.0, and allow operation in the Restricted Region. When operating in Setup, the Flow-biased Rod Block setpoints generally increase in power to the Flow-biased Scram or power / flow map boundaries. The Flow-biased Scram setpoint generally increases by an equivalent amount (within the power / flow map boundaries) to avoid spurious - scrams from power spikes. The inherent stability from maintaining FCBB less than onejustifies continued operation in the Restricted Region, but not in that portion of the power / flow map
.which in Setup becomes unprotected by the Flow-biased Scram. The alarm associated with the Rod Block ceases to be a RREA when in Setup, but signals to Operations a similar need to immediately move to a more stable region of the power / flow map. ~ The EI A Stability Solution generically provides for distinct Flow-biased Scram and Rod Block
- setpoints for recirculation pump SLO and two loop operation. However, for BNP the two loop
- operation (TLO) Flow-biased Scram and Rod Block setpoints, and TLO Stability Regions, are applicable for SLO without modification (Reference 1).
The EI A Stability Solution provides for distinct Flow-biased Scram and Rod Block setpoints for normal and reduced feedwater temperature conditions (" normal" and " alternate" setpoints) because the core stability decreases with feedwater temperature. Normal setpoints (Figures 13, 14,17,18) are to be used below 30% power or when feedwater temperature is within 50*F rated equivalent of nominal. The alternate setpoints (Figures 15,1619,20) are to be used above 30% rated thermal power when feedwater is reduced by more than 50'F rated equivalent (50*F * (% power /100)") in accordance with IOP-32. For example, a feedwater temperature reduction of 40"F from the design value associated with 50% power would require the attemate setpoints because this reduction would exceed the allowable reduction of 38.3 F, calculated by 50*0.50 * . REFERENCES-
- 1) BNP Design Calculation IB21-0537; " Preparation of the BlCl2 Core Operating l Limits Report," Revision 1; June 1998.
2)- NEDE-240ll-P-A; " General Electric Standard Application for Reactor Fuel,"
. (latest approved version).
- 3) NEDC-31624P, Supplement 1;" Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 1 Reload 11 Cycle 12," Revision 4, April 1998.
l I CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 11, Revision 1 Table I
' MCPR Limits Non-pressurization Transient MCPR Limits Exposure Range Fuel Type ' BOC-EOC GE8x8NB-3 1.26 gel 3 1.26 1
.l Pressurization Transient MCPR Limits: I Main Turbine Bypass System Operable
)
Normal and Reduced Feedwater Temperature MCPR Option Fuel Type - Exposure Range: Exposure Range: BOC to EOFPC-2205 EOFPC-2205 mwd /MT to l mwd /MT EOC A GE8x8NB 1.40 1.46 gel 3 1.40 1.46 8 GE8x8NB-3 1.35 1.38 l< GE13 1.35 1.38 l Pressurization Transient MCPR Limits:
- p. Inoperable Main Turbine Bypass System j.
j Exposure Range: BOC to EOC MCPR Option FuelType ' Normal Feedwater Feedwater Temperature Temperature Reduction A- GE8x8MB 3 1.49 1.50 GE13 - 1.49 1.50 B GE8x8NB-3 1,41 1.42 GE13 - 1.41 1.42
. nis Table is referred to by Improved Technical Specifications 3.2.2,3.4.1, and 3.7.6.
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 12, Revision 1 Table 2 RBM System Setpoints Setpoint Trip Setpoint Allowable Value Lower Power Setpoint (LPSP') 27.0 s 29.0 Intermediate Power Setpoint (IPSP') _ 62.0 564.0 High Power Setpoint (HPSP') 82.0 s 84.0 Low Trip Setpoint (LTSPD s 115.1 s 115.5 Intermediate Trip Setpoint (ITSPh s 109.3 s 109.7 High Trip Setpoint (HTSPh s 105.5 s 105.9 to s 2.0 seconds s 2.0 seconds Setpoints in percent of Rated Thermal Power. b Serpoints relative to a full scale reading of 125. For example, s 115.1 means s 115.1/125.0 of full scale. This Table is referred to by Improveo Technical Specification 3.3.2.1 (Table 3.3.2.1-1). i I
l l CP&t. Nuclear Fuels Mgmt. & Safety Analysis . Design Calc. No.1B21-0537 B1C12 Core Operating Limits Report Page 13. Revision 1 Table 3 - ; Aligned Drive Flow . l- The Scram and Rod Block trip setpoints are provided by Flow Control Trip Reference L (FCTR) cards. The FCTR cards have their drive flows calibrated each cycle by
- l. OPT-50.10,"APRM FCTR Card Drive Flow Alignment". The. calibration " aligns" the current cycle drive flow to the drive flow used when the EI A flow mapping L solution was developed for BNP. The COLR presents the BICl2 Scram and Rod l
' Block trip setpoints as a function of aligned drive flow. This Table 3 provides an equation for deriving the aligned drive flow from the FCTR card input drive flow signal:
100.005 A" - 30.2946 A'" + 69.7104 W5 g8* l 69.7104 -(A'" - A") where: Wois the aligned drive flow to be used for Figures 17 through 20 W3is the input drive flow signal A" and A'" are the current values for the FCTR card aligiunent i ! . This Table supports Improved Technical Specifications 3.2.3 and 3.3.1.1 and Technical Requirements Manual Specifications 3.2 and 3.3. l l l I' . , 1 E- - l
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page14 Revision 1 Figure 1 Fuel Type GE10-P8HXB322-11GZ-70M-150-T (GE8x8NB-3) Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 13.0 , - This Figure is Referred To By 12.0 Improved Technical Specification 3.2.1 11.0
- Exposure Limit h (GWd/MT) (kW/ft)
.x 0.00 10.84
- 10.0 -
0.22 10.88 -- 1.10 Permissible h; 2.20 10.95 11.08 Region of i g 3.31 11.21 Operation ' O 9.0 - 4.41 11.35 - _ 3 n. 5.51 11.50 6.61 11.67 4 7.72 11.92 8.82 12.26 8.0 - 9.92 12.61 l 11.02 12.95 l 13.78 13.12
~
16.53 12.89 4 7.0 - 22.05 12.29 l 27.56 11.67 33.07 10.96 38.58 10.26 44.09 9.52 6.0 - 49.60
~ ~~~
8.78 55.12 6.29 5.0 0 5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT) l L--_-_-___-_____-_________---__-__-_--__-------. . - _ _ _ _ _ - - - - - - - - - - _ - - - - - - - - - - - - - -- - - - . - . - - - - - - - - - - - - - - - -
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Umits Report Page 15, Revision 1 Figure 2 Fuel Type GE10-P8HXB320-11GZ-100M-150-T (GE8x8NB-3) Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 13.0 - This Figure is Referred To By 12.0 improved Technical ._ Specification 3.2.1 11.0
$ Exposure Limit 3 (GWdMT) (kW/ft) 6 10.0 -
0.00 10.81 - b 0.22 10.87 E J 1.10 10.97 2.20 11.10 E 3.31 11.24 g 9.0 - 4,43 3 j,39 . . J 5.51 11.53 Permissible k 6.61 11.68 Region of 7.72 11.84 Operation 8.0 - 8.82 12.00 -- 9.92 12.16 11.02 12.33 13.78 12.52 16.53 12.40 7.0 - 22.05 11.93 ~ ~ ~~-~ 27.56 11.44 33.07 10.85 38.58 10.26 6.0 -- 44.09 9.47 - 49.60 8.68 55.12 6.15 5.0 0 5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)
I l i CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 16, Revision 1 Figure 3 Fuel Type GE10-P8HXB348-12GZ-100M-150-T (GE8x8NB-3) l Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 13.0 t This Figure is Referred To By 12.0 - Improved Technical _. Specification 3.2.1 11.0
- Exposure Limit l
- 10.0 (GWd/MT) (kW/ft) 0.00 0.22 10.92 10.95 1.10 11.04 Permissible a 2.20 11.21 Region of 3.31 11.41 Operation 9.0 4.41 11.85 ~
5.51 11.90 d 6.61 12.08 . 4 7.72 12.26 l 8.82 12.47 , 8.0 9.92 12.71 ; 11.02 12.94 13.78 12.86 16.53 12.73 7.0 22.05 . 12.05 27.56 11.33 33.07 10.69
- 38.58 10.06
, 44 09 9.41 6.0 49.00 8.76 - - 55.12 6.46 i
- . 5.0 0 5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT)
I i CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537
C12 Core Operating Limits Report Page17, Revision 1 Figure 4 i
Fuel Type GE10-P8HXB346-12GZ-100T-150-T (GE8x8NB-3) Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 13.0 _ This Figure is Referred To By 12.0 - Improved Technical _ Specification 3.2.1 11.0 - -- - Exposure Umst a (GWd/MT) (kWMt) 0.00 10.54 l
.w 0.22 10.57
' ^
- 10.0 ~
1.10 10.66 . . 2.20 10.82 Permissible a 3.31 11.02 Region of g 4.41 11.25 Operation g 9.0 5.51 11.49 -- - 6.61 11.66 N 7.72 11.84 4 8.19 11.90 8.82 11.94 8.0 _ 9.92 12.03 l i 11.02- 12.12 i 13.78 12.10 l
. 16.53 11.75 7.0 21.64 11.17 - - -
22.05 11.12 1 27.56 10.75 38.58 10.28 48.26 10.21 l 6'0 49.60 9.88' . 53 ?S 8.57 l 56/ I 6.33 I' 5.0 0 5 10 15 20 25 30 35 40 45 50 55 60 AVERAGE PLANAR EXPOSURE (GWd/MT) L L --- - _ _ - - - _ _ - - - . - - - - - - - - - - - - - - - - - - - - - - - - _
t I CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 18, Revision 1
- Figure 5 l Fuel Type GE13-P9DTB380-11G5.0-100T-146-T (GE13) i Average Planar Linear Heat Generation Rate (APLHGR) Limit
- Versus Average Planar Exposure i
14.0 j l 13.0 - - - i This Figure is Referred To By 12.0 Improved Technical _ Specification 3.2.1 11.0 , -- Exposure Limit ---- --- (GWd/MT) (kW/ft) { 0.00 10.94 g 0.22 10.99 6 10.0 --. 1.10 11.11 E 2.20 11.29 5 3.31 11.49 "3 4.41 11.70 cc 5.51 11.90 g 9.0 6.61 12.07 g 7.72 12.24 8.82 Permissible 11.02 12.61 Region of ) 8.0 _ 13.78 12.66 Operation 16.53 12.39 19.29 12.03 22.05 11.66 7.0 27.56 10.93 33.07 10.21 38.58 9.51 44.09 8.82 6~0 49.60 8.15 55.12 7.48 60.63 6.80 63.68 6.41 5.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 AVERAGE PLANAR EXPOSURE (GWd/MT) l I'
_ - , , _ _ _ _ - _ _ - _ , - _ _ _ - -- - - - - - - - - - - - - ' - " - - - ' ' ' - '-~ ~ CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 19 Revision 1 Figure 6 Fuel Type GE13-P9DTB380-10G5.0-100T-146-T (GE13) Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 13.0 This Figure is Referred To By 12.0 improved Technical . _ _ Specification 3.2.1 Exposure U mst 11.0 (GWd/MT) (kWM) 0.00 11.17 0.22 11.20 3{ 1.10 11.32 E 10.0 2.20 11.49 t:- 3.31 11.68 5 4.41 11.88 3 5.51 12.03 E 6.61 12.14 0 9.0 7.72 12.25 5 8.82 12.36 I k 9.92 11.02 12.47 12.59 . Permissible 8.0 13.78 12.65 Region of 16.53 12.39 .' 19.29 12.03 Operation 22.05' 11.67
. 27.56 10.94 7.0 33.07 10.22 38.58 9.52 44.09 8.83 49.60 8.15 6.0 55.12 7.48 60.63 6.80 63.56 6.43 5.0 l 0 5 10 15 20 25 30 35 40 45 50 55 60 65 l AVERAGE PLANAR EXPOSURE (GWd/MT)
__-__m____.________.________-____________________ _ _ _ . _ _ _ _ _ _ _ _ . _ m w..
l l CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 20, Revision 1 Figure 7 Fuel Type GE13-P9DTB403-5G6.0/7G5.0-100T-146-T (GE13) Average Planar Linear Heat Generation Rate (APLHGR) Limit ! Versus Average Planar Exposure j 14.0 13.0 _ This Figure is Referred To By 12.0 Improved Technical __. Specification 3.2.1 11.0 - Exposure Umrt - _ (GWd/MT) . (kW/ft) g: ' O.00 10.65 3 0.22 10.72 6 10.0 _. 1.10 10.85 _ _ _ h 2.20 11.00 3 3.31 11.12 3 4.41 11.25 GC 5.51 11.38 Permissible g 9.0 6.61 11.52 Region of , g 7.72 11.66 Operation i 4 8.82 11.81 j 9.92 11.95 ' 8.0 11.02 12.05 13.78 12.04 16.53 11.97 19.29 11.79 22.05 11.54 7.0 27.56 11.02 33.07 10.44 38.58 9.69 , 44.09 8.98 l 6.0 49.60 8.30 . _ _ _ _ _ 55.12 7.64 60.63 7.00 64.48 6.54 l 5.0 - 0 5 10 15 20 25 30 35 40 45 50 55 60 65 l AVERAGE PLANAR EXPOSURE (GWdTdT)
~.
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 21. Revision 1 Figure 8 Fuel Type GE13-P9DTB403-7G6.0/7G5.0-100T-146-T (GE13)
- Average Planar Linear Heat Generation Rate (APLHGR) Limit
} Versus Averag'e Planar Exposure l 14.0 13.0 - . _ l This Figure is Referred To By 12.0 , , ,
- - - - Improved Technical _
Specification 3.2.1 11.0 - Exposure Limit (GWd/MT) (kW/ft) O.00 10.44 I':!E 10.0 0.22 1.10 10.51 10 61 . , _ i b 2.20 10.74 i 3 3.31 10.88 ! "3 4.41 11.02 01: 5.51 11.17 0 9.0 6.61 11.32 3 7.72 11.48 Permissible g 8.82 11.62 Region of 1 8.0 13.78 11.86 16.53 11.86 13.29 11.76 22.05 11.54 7.0 27.56 11.02 ' -- 33.07 10.49 38.58 9.85 44.09 9.13 6.0 49.60 8.43 _ 55.12 7.73 60.63 7.03 64.29 6.56 5.0 0 5 10 15 20 25 30 35 40 45 50 55 60 65 AVERAGE PLANAR EXPOSURE (GWd/MT) o
l l. l-l CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 81C12 Core Operating Limits Report Page 22, Revision 1 p Flow-Dependent MAPLHGR Limit, MAPLHGR(F) 1.10 This Figure is Referred To By Improved Technical Specifications - Two Loop Operation Limit 3.2.1 and 3.4.1 1.00 - - - - Max Flow = 102.5% J 107 % ! 112 % ) 0.90 ~~ ~ - ~ ' - M7% { I Single Loop Operation Limit k 0.80 -- - a x
- u. !
b u. g 0.70 --- x i g MAPLilGR(F) = MAPFACr
- MAPLHGRsro MAPLilGRsip = Standard MAPLHGR Limits MAPFACr(F) = Minimum (1.0, ArWe/100+Br) 0.60 We = % Rated Core Flow Ar And Br Are Fuel Type Dependent Constants Given Below:
Mas Core Flow (% Rated) Ar Br 0.50 102.5 0.6784 0.4861 .. 107.0 0.6758 0.4574 112.0 0.6807 0.4214 117.0 0.6886 03828 l 0.40
-- 30 40 50 60 70 80 90 100 110 Core Flow (% Rated) i
l CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 23, Revision 1 Power-Dependent MAPLHGR Limit, MAPLHGR (P) ' 1.00 This Figure is Referred To By f improved Technical Specifications _. 3 2.1, 3.4.1 and 3.7.6 _ _ . _ . Two Loop Operation Limit 0.90 --- _ T O i 4 l I 0.80 __.___- - _-_ _ _ _... . _., il a j Single Loop Operation Limit n e [ 0.70 i l Core Flow 5 50% MAPLHGR(P) = MAPFAC,
- MAPLHGRsto l
Turbine Bypass WLHGRsTo = Standard MAPLHGR Limits 0.60 f
- n. ,
e TIsrbine liyp~as~s 's
~
For P < 25% : No Thermal Limits Monitoring Required e 8 Inoperable e For 25% $ P < 30%: s - -e ***-- For Core Flow 5 50% & Turbine Bypass Operable, 8 MAPFAC, = 0.585 + 0.005224 (P-30%)
, For Core Flow s 50% & Turbine Bypass inoperable.
l MAPFAC, = 0.567 + 0.0128 (P-30%) For Core Flow > 50%, Core Flow > 50% MAPFAC, = 0.433 + 0.0052 (P-30%) Turbine Bypass -- For P > 30%: J
)( Operable or MAPFAC,= 1.0 + 0.005224 (P-100%) 1 Inoperable 0.40 l 20 25 30 35 40 45 50' 55 60 65 70 75 80 85 90 95 100 Power (% Rated)
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 B1C12 Core Operating Limits Report Page 24, Revision 1 p Flow-Dependent MCPR Limit, MCPR(F) 1.80 For We (% Rated Core Flow) < 40%, MCPR(F) = ( A,Wc/100+ Br)'[ l +0.0032 (40-Wc)] 1.70 __- For We (% Rated Core riow) > 40%, MCPR(F) = Max (1.20, A,Wc/100+Br) Max Core Flow 1.60 - _ (% htal) Ar B, __ 102.5 0.571 1.655 107.0 - 0.586 1.697 112.0 - 0.602 1.747 117.0 - 0.632 1.809 1.50 --
- a.
N 1.40 --- - -- This Figure is Referred To By improved TechnicalSpecification Max Flow = 117% 3 2.2,3.4.1 and 3.7.6 1.30 -- - - - - - _ - 102.5 % 1.20 1.10 20 30 40 50 60 70 80 90 100 110 120 j Core Flow (% Rated) l { l < i
CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 1821-0537 ; B1C12 Core Operating Limits Report Page 25, Revision 1 Figure 12 ; Power - Dependent MCPR Limit, MCPR (P) 3.30 OLMCPR 3.20 --
; ; R'ated MCPR Multiplier (Kp) ;
3.10 -- -
- - - - - - ~ - -
3.00 ---\ l-f \ l50Tore
% Flow >Operating Limit MCPR(P) = Kp* Operating Limit MCPR(100)
M 2.90 v \ g Turbine BypassFor l- P < 25% : I 2.80 Jogable _ l- No Thermal Limits Monitoring Required h 2.70
"$" 2.60 Core Flow > _
For 25% $ P5 PaypAss : Where PsypAss = 30%
! 50 % K, = Maximum of 1.481 or Keu.
E A 2.50 Turbine Bypass For Core Flow s 50% & Turbine Bypass Operable, o Operable Keep = [1.90 + 0.02 (30% - P)] / OLMCPR(100) 3 2.40 y -- For Core Flow > 50% & Turbine Bypass Operable O Kpt, = [2.20 + 0.02 (30% - P)] / OLMCPR(100)
. 2.30 m, For Core Flow 550% & Turbine Bypass inoperable, g
- 2.20 Ket, = [1.96 + 0.072 (30% - P)] / OLMCPR(100) 1 For Core Flow > 50% & Turbine Bypass Inoperable '
2.10 Tore Flow < 50%l p=[.
+ 0.05 (30% - @ ONR@) i g-l Turbine Bypass g-2.00 1 inoperable -
L For 30% $ P < 45% : K, = 1.28 + 0.0134 (45% P) 1.90 Al For 45% ~< P < 60% : g .80 - Core Flow < 50% - K, = 1.15 + 0.00867 (60% P) j Turbine Bypass 1.70 - g Operable po, P > ~ 60% - f 1.60 Ke = 1.00 + 0 00375 (100% - P)
.t I li. 1.50 . , ,
E
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] 1.20 -
Specification 322,34.1,3.7.6 -- i is ! at 3,jo . _ _ _ . _ 1.00 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 PsypAss Power (% Rated)
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