ML20206N179

Rev 0 to 2B21-0554, Brunswick Unit 2,Cycle 14 Colr

ML20206N179
Person / Time
Site:	Brunswick
Issue date:	04/23/1999
From:	Money M, Galen Smith CAROLINA POWER & LIGHT CO.
To:
Shared Package
ML20137N757	List: BSEP-99-0062, Forwards Latest Revs of COLR & Suppl Reload Licensing Rept, for Bsep,Unit 2.Proprietary Rev 6,Suppl 2 to NEDC- 31624P & EMF-2168P,included.Proprietary Encls Withheld, Per 10CFR2.790 & 10CFR9.17 ML20206N179 ML20206N183
References
2B21-0554, 2B21-0554-R00, 2B21-554, 2B21-554-R, NUDOCS 9905170160
Download: ML20206N179 (32)

Text

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 O

B2C14 Core Operating Limits Report Page 1, Revision 0 BRUNSWICK UNIT 2, CYCLE 14 CORE OPERATING LIMITS REPORT April 1999 O

i hi(ht/fo, Date:

2Ob3 Prepared By:

/1 uk

()ry V. Money

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Approved By:

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Date:

06orge E. Smith Superintendent BWR Fuel Engineering O

9905170160 990511 PDR ADOCK 05000324 PDR P

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i CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc, No. 2B21-0554 B2C14 Core Operating L:mits Report Page 2, Revision 0 l

LIST OF EFFECTIVE PAGES Page(s)

Revision I

1 32 0

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t'^s CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc No. 2B21-0554 V

B2C14 Core Operating Limits Report Page 3, Revision 0 1

i TABLE OF CONTENTS l

l Subject Paae Cover....

............1 List of Effective Pages........

.2 Table of Contents...

.3 List of Tables.,

....4 List of Figures........

.4 Introduction and Summary......

.5 Single Loop Operation...

.6 Inoperable Main Turbine Bypass System.................

... 6 APLHGR Limits.......

...6 MCPR Limits...

.7 RBM Rod Block instrumentation Setpoints..

.7 THI E1 A Stability Solution...........

..... 7 References..............

.....8

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'(d B2014 Core Operating Limits Report CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 Page 4, Revision 0 CAUTION References to COLR Figures or Tables should be made using titles only; figure and table numbers may change from cycle to cycle.

1 LIST OF TABLES l

Table Title

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Table 1:

MCPR Limits....

.9 Table 2:

RBM System Setpoints.........

.10 Table 3:

Aligned Drive Flow......

.... 11 LIST OF FIGURES l

l Fig Lre Title or Description Page.

Figure 1:

APLHGR Limit Versus Average Planar Exposure.

.. 12 Figure 2:

APLHGR Limit Versus Average Planar Exposure.

13 b

i i

V Figure 3:

APLHGR Limit Versus Average Planar Exposure...........

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Figure 4:

APLHGR Limit Versus Average Planar Exposure...

...... 15 Figure 5:

APLHGR Limit Versus Average Planar Exposure...

16 Figure 6:

APLHGR Limit Versus Average Planar Exposure....

....... 17 Figure 7:

APLHGR Limit Versus Average Planar Exposure.........

.18 Figure 8:

Not Used.

.19 Figure 9:

Flow-Dependent MAPLHGR Limit, MAPLHGR(F)...

...... 20 Figure 10: Power-Dependent MAPLHGR Limit, MAPLHGR(P).

... 21 j

Figure 11: GE13 Flow-Dependent MCPR Limit, MCPR(F).......

.. 22 Figure 11a: A10 Flow-Dependent MCPR Limit, MCPR(F).....

...... 23 Figure 12: Power-Dependent MCPR Limit, MCPR(P)....

.24 Figure 13: Power / Flow Map Stability Regions: Normal T rw, Non-Setup.,

.. 25 i

Figure 14: Power / Flow Map Stability Regions: Normal T rw, Setup..

....................................26 1

Figure 15: Power / Flow Map Stability Regions: Reduced T rw, Non-Setup............

.27 Figure 16: Power / Flow Map Stability Regions: Reduced T rw, Setup...

....... 28 Figure 17: E1 A Setpoint Allowable Values versus Aligned Drive Flow: Normal T rw, Non-Setup..... 29 l

Figure 18: E1 A Setpoint Allowable Values versus Aligned Drive Flow: Normal T rw, Setup.

.30 Figure 19: E1 A Setpoint Allowable Values versus Aligned Drive Flow: Reduced T rw, Non-Setup........... 31 Figure 20: EI A Setcoint Allowable Values versus Aligned Drive Flow: Reduced T rw, Setup.

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O CP&L Nuclear Fuels Mgmt. & Safety Analysis J

Design Calc. No. 2B210554 B2C14 Core Operating Limits Report Page 5, Revision 0 Introduction and Summary This report provides the values of the power distribution limits and control rod withdrawal block instrumentation setpoints for Brunswick Unit 2, Cycle 14 as required by TS 5.6.5.

OPERATING LIMIT REQUIREMENT Average Planar Linear Heat Generation Rate (APLHGR) limits TS 5.6.5.a.1 (with associated core flow and core power adjustment factors)

Minimum Critical Power Ratio (MCPR) limits TS 5.6.5.a.2 (with associated core flow and core power adjustment factors)

Allowable Values for Function 2.b of TS 3.3.1.1, APRM Flow Biased TS 5.6.5.a.3 Simulated Thermal Power High Allowable Values and power range setpoints for Rod Block Monitor Upscale TS 5.6.5.a.4 Functions of TS 3.3.2.1 Per TS 5.6.5.b and 5.6.5.c, these values have been determined using NRC approved methodology and are established such that all applicable limits of the plant safety analysis are met.

The limits specified in this report support single loop operation (SLO) as required by TS LCO 3.4.1 and inoperable Main Turbine Bypass System as required by TS 3.7.6.

In order to support the Thermal Hydraulic Instability (THI) E1 A Stability Solution, the following is also included in this report:

OPERATING LIMIT REQUIREMENT Thermal Hydraulic Instability (THI) E1 A Stability Solution TS 3.2.3 and 3.3.1.3, Monitored Region and Restricted Region and TRMS 3.2 Thermal Hydraulic Instability (THI) E1 A Stability Solution Implicit Exclusion Region

" Setup" and *Non-Setup" scram values of the APRM Flow Biased Simulated TS 3.2.3 and 3.3.1.1 Thermal Power-High Allowable Value (" Flow Biased Scram")

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" Setup" and "Non-Setup" control rod block values of the APRM Flow Biased -

TRMS 3.3 Upscale Allowable Value (* Flow Eiased Rod Block")

Four Siemens ATRIUM-10 (A10) Lead Qualification Assemblies will be loaded in the B2C14 core.

j Reference 4 concludes the A10 is bounded by the GE13 operating limits and licensing analyses, provided additional operating and design constraints are imposed on the GE13 fuel type used to monitor the A10. The i

additional operating requirements have been incorporated herein as applicable.

Preparation of this report was performed in accordance with Quality Assurance requirements as specified in l

Reference 1.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 B2C14 Core Operating Limits Report Page 6, Revision 0 Sinole Loop Operation Brunswick Unit 2, Cycle 14 may operate over the entire MEOD range with Single recirculation Loop Operation (SLO) as permitted by TS 3.4.1 with applicable limits specified in the COLR for TS LCO's 3.2.1,3.2.2 and 3.3.1.1:

LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR) Limits: per Reference 1 and Figures 9 and 10, the APLHGR Limits include a SLO limitation of 0.8 on the MAPLHGR(F) and MAPLHGR(P) multipliers.

LCO 3.2.2, Minimum Critical Power Ratio (MCPR) Limits: per Reference 1, Table 1 and Figures 11,11a and 12, the MCPR limits presented 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 and the THi E1 A STABILITY SOLUTION, these limits apply to SLO without modification.

Inoperable Main Turbine Bypass System Brunswick Unit 2, Cycle 14 may operate with an inoperable Main Turbine Bypass System in accordance with TS 3.7.6 with applicable limits specified in the COLR for TS 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 entire MEOD range and up to 110*F rated g) equivalent feedwater temperature reduction. The system response tirne assumed by the safety analyses from iV event initiation to start of bypass valve opening is 0.10 seconds, with 80% bypass flow achieved in 0.30 seconds. The applicable limits are as follows:

LCO 3.2.1, Average Planar Linear Heat Generation Rate (APLHGR) Limits: in accordance with Reference 1 as shown in Figure 10, TBPOOS requires a reduction in the MAPLGHR(P) limits between 25% and 30% power.

LCO 3.2.2, Minimum Critical Power Ratio (MCPR) Limits: in accordance with Reference 1, TBPOOS requires an increase in the MCPR(P) multiplier between 25% and 30% power, as shown in Figure 12.

TBPOOS also requires increased MCPR limits, included in Table 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 1 through 7. These values were determined with the SAFER /GESTR LOCA methodology described in GESTAR-il (Reference 2). Figures 1 through 7 are to be used only when hand calculations are required as specified in the bases for TS 3.2.1. Hand calculated results may not match a POWERPLEX calculation since normal monitoring of the APLHGR limits with POWERPLEX uses the complete set of lattices for each fuel type provided in Reference 3.

The core flow and core power adjustment factors for use in TS 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.

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O CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2821-0554 B2C14 Core Operating Limits Report Page 7. Revision 0 MCPR Limits The ODYN OPTION A ODYN OPTION B and non-pressurization transient MCPR limits for use in TS 3.2.2 for each fuel type as a function of cycle average exposure are given in Table 1. These values were determined i

i with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR ll (Reference 2) and are consistent with a Safety Limit MCPR of 1.09 specified by TS 2.1.1.2.

The core flow and core power adjustment factors for use in TS 3.2.2 are presented in Figures 11,11a and 12.

For any given powerMiow state, the maximum of MCPR(F) determined from Figure 11 or 11a and MCPR(P) determined from Figure 12 is used to determine the governing limit.

All MCPR limits presented in Table 1 Figure 11, Figure 11a and Figure 12 were determined without EOC-RPT operable and apply to two recirculation pump operation and SLO without modification.

RBM Rod Block instrumentation Setooints The nominal trip setpoints and allowable values of the control rod withdrawal block instrumentation for use in TS 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 linots with the GEMINI methodology and GEXL-PLUS critical power correlation described in GESTAR-il (Reference 2).

O THl E1A Stability Solution The Enhanced Option 1 A methodology was used to develop the THi E1 A Stability Solution, which involves exclusion from certain areas of the powerMiow map and specific restrictions for operating in other areas.

The COLR provides the Stability Regions on the power /(core) flow map in Figures 13-16. These Figures define the Monitored and Restricted Regions for compliance with TS 3.2.3, TS 3.3.1.3 and TRMS 3.2 (and indirectly TS 3.3.1.1 and TRMS 3.3), and include the Exclusion Region (for which definition in the COLR is not a TS requirement). Core flow nominal trip setpoint values on Figures 13-16 correspond to the nominal trip i

setpoint values translated into drive flow and installed in the Flow Control Trip Reference (FCTR) cards.

Automatic features of the TH1 E1 A Stability Solution implementation use digital FCTR cards that incorporate Trip Reference setpoints which are equivalent or more restrictive than the pre-Stability Solution APRM flow-biased and clamped limits. The FCTR cards support TS 3.3.1.1 (automatic APRM Flow-biased Scram) and TRMS 3.2 (Restricted Region Entry Alarm, which uses the TRMS 3.3 Flow-biased Rod Block setpoint).

Figures 17-20, E1 A Setpoint Allowable Values Versus Aligned Drive Flow, are based on drive flow and not core flow to support the flow signal used for the FCTR cards. A!so, Figures 17-20 allow quantification of Technical Specification compliance once the drive flow input is aligned in accordance with Table 3.

"Non-Setup" setpoints (Figures 13,15,17,19) enforce the normal Exclusion and Restricted Regions described above. Setup setpoints (Figures 14,16,18,20) are to be used only when 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 powerMiow 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 one justifies 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.

s 1

For BNP the two loop operation (TLO) Flow-biased Scram and Rod Block setpoints, and TLO Stability 1

Regions, are equivalent to the SLO counterparts over all applicable portions of the operating domain.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 D

B2C14 Core Operating Limits Report Page 8, Revision 0 The E1 A Stability Solution provides for distinct Flow-biased Scram and Rod Block setpoints for normal and reduced feedwater temperature conditions (" normal" and "altemate" setpoints) because the core is more susceptible to instabilities with decreasing 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.

Altemate setpoints (Figures 15,1619,20) are to be used above 30% power or when feedwater is reduced by more than 50 F rated equivalent (50*F * (% power /100)") in accordance with 2OP-32.

References

1) BNP Design Calculation 2B21-0554;" Preparation of the B2C14 Core Operating Limits Report,"

Revision 0, April 1999.

2) NEDE-24011-P-A; " General Electric Standard Application for Reactor Fuel," (latest approved version).

i

3) NEDC-31624P," Loss-of-Coolant Accident Analysis Report for Brunswick Steam Electric Plant Unit 2 Reload 13 Cycle 14," Supplement 2, Revision 6, February 1999.

4) EMF-2168(P)," Brunswick ATRIUM-10 Lead Out.fification Assemblies Safety Analysis," Revision 0, March 1999.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 k

B2C14 Core Operating Limits Report Page 9, Revision 0 Table 1 MCPR Limits (EOC-RPT Not Required)

Steady State, Non-pressurization Transient MCPR Limits j

FuelType Exposure Range: BOC-EOC GE13 1.29 A10 1.43 Pressurization Transient MCPR Limits, OLMCPR (100%P): Turbine Bypass System Operable Normal and Reduced Feedwater Temperature Exposure Range:

Exposure Range:

MCPR Option Fuel Type BOC to EOFPC-2205 mwd /MT EOFPC-2205 mwd /MT to EOC A

GE13 1.39 1.46 A10 1.55 1.62 B

GE13 1.34 1.38 A10 1.49 1.53 Pressurization Transient MCPR Limits, OLMCPR (100%P): Turbine Bypass System Inoperable Normal and Reduced Feedwater Temperature MCPR Option Fuel Type BOC to EOC A

GE13 1.48 i

A10 1.65 B

GE13 1.40 A10 1.56 This Table is referred to by Technical Specifications 3.2.2,3.4.1 and 3.7.6.

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CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 82C14 Core Operating Limits Report Page 10, Revision 0 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 s 64.0 High Power Setpoint (HPSP*)

82.0 s 84.0 6

Low Trip Setpoint (LTSP )

5 115.1 s 115.5 b

Intermediate Trip Setpoint (ITSP )

sm2 sWJ l

High Trip Setpoint (HTSPD s 105.5 s 105.9 tar s 2.0 seconds s 2.0 seconds Setpoints in percent of Rated Thermal Power.

b Setpoints 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 byTechnical Specification 3.3.2.1 (Table 3.3.2.1-1).

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l CP&L Nuclear Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 B2014 Core Operating Limits Report Page 11, Revision 0 Table 3 Aligned Drive Flow The Scram and Rod Block trip setpoints are provided by Flow Control Trip Reference (FCTR) cards. The FCTR cards have their drive flows calibrated each cycle by OPT-50.10,"APRM FCTR Card Drive Flow l

Alignment". The calibration " aligns" the current cycle drive flow to the drive flow used when the E1 A flow mapping solution was developed for BNP. The COLR presents the Scram and Rod Block trip setpoints as a function of aligned drive flow. This table provides an equation for deriving the aligned drive flow from the FCTR card input drive flow signal:

100.005 A" - 30.294 8" + 69.711 W3 W"

• 69.711-(d* - A")

where:

W is the a a and o'ggned drive flow to be used for Figures 17 through 20 g

are the current values for the FCTR card alignment W is the input drive flow signal 3

This Table supports Technical Specifications 3.2.3 and 3.3.1.1 and Technical Requirements Manual Specifications 3.2 and 3.3.

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CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 B2C14 Core Operating Limits Report Page 12, Revision 0 Figure 1 Fuel Type GE13-P9DTB363-11GZ-100T-146-T (GE13) l Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure

]

14.0 This Figure is Referred To By j

13.0 Technical Specification 3.2.1 D

12.0 /

11.0 Exposure Lrut (GWd/MT) (kW/ft)

I h

]

jj]

Permissible g

6 10.0 Region of 1.10 11.62 I:

2.20 11.71 Operation E

3.31 11.Bo 4.41 11.89 h

5.51 11.99 z 9.0 6.61 12.10

[

7.72 12.21 4

8.82 12.33 9.92 12A6 11.02 12.59 8.0 13.78 12.60 16.53 12.35 19.29 12.08 22.05 11.82 27.56 11.14 33.07 10.43 7.0 38.58 9.74 44.09 9.07 49.60 841 55.12 7.73 60.63 7.04 6.0 65.26 s.44 t']

5.0 C

0 5

10 15 20 2-5 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 q

B2C14 Core Operating Limits Report Page 13, Revision 0 V

Figure 2 Fuel Type GE13-P9DTB363-11GZ1-100T-146-T (GE13)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 This Figure is Referred To By 13.0 Technical Specification 3.2.1 D \\

12.0 11.0 O

Exposure Limit m

E (GWd/MT) (kW/ft)

Permissible k 10.0 -

0 j4 Region of

\\

2 6

b 1.10 11.62 Operation 2

2.20 11.70 3

3.31 11.79 h

4.41 11.89 5

8-S'.1 66 12.09 7.72 12.20 8.82 12.32 9.92 12.44 11.02 12.55 8.0 13.78 12.57 N

16.53 12.34 1

19.29 12.07 22.05 11.81 l

27.56 11.14 7.0 33.07 10.43 h

38.58 9.73 44.09 9.06 49.60 8.41 y

55.12 7.73 60.63 7.04 6.0 65.29 6.44 i

O 50 0

5 10 15 20 25 30 35 40 45 50 55 60 65 70 AVERAGE PLANAR EXPOSURE (GWd/MT)

CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 B2C14 Core Operating Limits Report Page 14, Revision 0 V

Figure 3 Fuel Type GE13-P9DTB393-4G6.0/9G5.0-100T-146-T (GE13)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 This Figure is Referred To By 13.0 Technical Specification 3.2.1 12.0 11.0 '

h Exposure Limit O

_C (GWd/MT) (kW#t)

E 0.00 11.03 Permissible

!!g 10.0 0.22 11.09 Region of b

1.10 11.19 Operation E

2.20 11.28 J

3.31 11.37

@ 9.0 4.41 11.47 I

5.51 11.57 k

g 6.61 11.67 7.72 11.78 8.82 11.89 9.92 12.00 11.02 12.11 g

8.0 13.78 12.12 16.53 12.00 19.29 11.76 22.05 11.50 27.56 10.96 33.07 10.28 7.0 38.58 9.54 44.09 8.85 49.60 8.19 f

55.12 7.55 60.63 6.92 j

6.0 64.14 6.48

^

5.0

\\

0 5

10 15 20 25 30 35 40 45 50 55 60 65 i

AVERAGE PLANAR EXPOSURE (GWd/MT) l I

CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 B2C14 Core Operating Limits Report Page 15, Revision 0 Fuel Type GE13-P9DTB395-12G5.0-100T-146-T (GE13)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 I

J This Figure is Referred To By 13.0 Technical Specification 3.2.1 12.0 Y

/

%N 11.0 N

Exposure Limit I{ 10.0 1 4 (GWd/MT) (kW#t)

Permissible

\\

1.10 11.38 Region of h

2 2.20 11.50 Operation 3

3.31 11.57 gr 4.41 11.66 0

5.51 11.74 3

6.61 11.83 A

7.72 11.92 8.82 12.02 9.92 12.11

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11.02 12.21 8.0 13.78 12.21 16.53 12.04 19.29 11.78 22.05 11.51 27.56 10.98 33.07 10.44

.0 38.58 9.73 44.09 9.03 49.60 8.33 55.12 7.64 60.63 6.95 6.0 64.28 6.49 5.0

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0 5

10 15 20 25 30 35 40 45 50 55 60 65 AVERAGE PLANAR EXPOSURE (GWd/MT)

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CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 B2C14 Core Operating Limits Report Page 16, Revision 0 Fuel Type GE13-P9DTB403-5G6.0/7G5.0-100T-146-T (GE13)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 This Figure is Referred To By 13.0 Technical Specification 3.2.1 12.0 N

11'0 [

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Exposure Limit 10.0 (GWd/MT) (kW/ft)

\\

{

Permissible E

1.10 10.85 Region of J

2.20 11.00 Operation 3.31 11.12 I

9.0 4.41 11.25 p

5.51 11.38 6.61 11.52 7.72 11.66 8.82 11.81 9.92 11.95

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8.0 11.02 12.05 13.78 12.04 16.53 11.97 19.29 11.79 22.05 11.54 27.56 11.02 7.0 -

33.07 10.44 38.58 9.69 i

44.09 8.98

~

49.60 8.30 55.12 7.64 6.0 60.63 7.00 64.48 6.54 5.0 0

5 10 15 20 25 30 35 40 45 50 55 60 65 AVERAGE PLANAR EXPOSURE (GWd/MT)

CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 B2C14 Core Operating Limits Report Page 17, Revision 0 Fuel Type GE13-P9DTB403-7G6.0/7G5.0-100T-146-T (GE13)

Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 This Figure is Referred To By 13.0 Technical Specification 3.2.1 12.0 r

x 11.0 0

\\

10.0 Exposure Limit

\\

{

(GWd/MT) (kW/ft)

Permissible 0.00 10.44 g

Region of 0;

Operation 3

o 2.20 10.74 5

8-33' 10 88 n.

4.41 11.02 4

5.51 11.17 6.61 11.32 7.72 11.48 8.0 8.82 11.62 9.92 11.73 K

11.02 11.85 13.78 11.86 16.53 11.86 19.29 11.76 22.05 11.54 7.0 27.E6 11.02 33.07 10.49 38.58 9.85 44.09 9.13 49.60 8.43 6.0 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)

L CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 B2C14 Core Operating Limits Report Page 18, Revision 0 Fuel Type Atrium-10 Average Planar Linear Heat Generation Rate (APLHGR) Limit Versus Average Planar Exposure 14.0 This Figure is Referred To By 13.0 Technical Specification 3.2.1 12.0

\\

11 0 O

gg Exposure Limit

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5, 10.0 (GWd/MT) (kW/ft) b 0.00 10.65 Permissible 1

3 0.22 10.72 O' " O'

~~.i 1.10 10.85 et 2.20 11.00 Operation O

3.31 11.12 l

5 8o 4.4i 11.25 g

5.51 11.38 6.61 11.52 7.72 11.66 8.82 11.81 8.0 9.92 11.95 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 49.60 8.30 55.12 7.64 6.0 60.63 7.00 64.48 6.54 5.0 0

5 10 15 20 25 30 35 40 45 50 55 60 65 AVERAGE PLANAR EXPOSURE (GWdMT)

CP&L Nuclear Fuel Mgmt. & Safety Analysis Design Calc. 2B21-0554 p

82C14 Core Operating Limits Report Page 19, Revision 0 O

Figure 8 l

Not Used O

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CP&L Nuclear Fue:s Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 B2C14 Core Operating Limits Report Page 20, Revit.,

• O O-Figure 9 Flow-Dependent MAPLHGR Limit, MAPLHGR(F) 1.10 This Figure is Referred To By Two Loop Operation Limit l Technical Specifications 1.05 3.2.1 and 3.4.1 1.00

,////

Max Floie = 102.5%

\\

A 0.95 s u nn m 7

I 0.90

]

sSingle Loop Operation LimitL

n. 0.85 i

/ f l N' l

5

/ /=/ /-

15 0.80 s

l

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- =

=

=

=

=

A A

A z o 75

/ / / /

,/,/,/,/

O.70 i

} 0.65///

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MAPLHGR(F) = MAPFACr

• MAPLHGRsm MAPLHGR p = Standard MAPLHGR Limits 37 E

MAPFACr(F) = Minimum (1.0, ArWc/100+Br) 0.60 w = % Rated Core Flow g

c Ar And Br Are Fuel Type Dependent Constants Given Below:

0.55 Max Core Flow

(% Rated)

Ap By 102.5 0.6784 0.4861 0.50 107.0 0.6758 0.4574 112.0 0.6807 0.4214 117.0 0.6886 0.3828 0.45 0.40 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 Core Flow (% Rated)

CP&L Nuclur Furls Mgmt. & Saf:ty Analysis Design Calc. No. 2B21-0554 B2C14 Core Operating Limits Report Page 21, Revision 0 Figure 10 Power-Dependent MAPLHGR Limit, MAPLHGR (P) 1.0--

l This Figure is Referrad To By 0.95 Technical Speci' cations L

3.2.1, 3.4.1 ar.d 3.7.6 i

/

0 f,0 r

i t

1.35 lTwo Loop Operation Lirnitl h

l l

g 0.80

=

=

• =

- - - = ~ -

k

$ 0.75 O j

/

'E"I*E I8"'"i "3*1

5 0.70 II 0.65

/

l Core Flow s 50%l MAPLHGR(P) = MAPFACp

• MAPLHGRsTo MAPLHGRsTo = Standard MAPLHGR Limits h 0'60 Turbine Bypass f

Operable For P < 25% :

o n-No Thermal Limits f.lonitoring Required

~ ~ Ine Bypass"s For 25% s P < 30%:

~

Turb For Core Flow s 50% & Turbine Bypass Operable, e

Inoperable 0.55 MAPFACp = 0.585 + 0.005224 (P-30%)

For Core Flow s 50% & Turbine Bypass inoperable,

,4 MAPFACp = 0.567 + 0.0128 (P-30%)

For Core Flow > 50%,

r 0.50 MAPF ACp = 0.433 + 0.005224 (P-30%)

For P t 30%:

Core Flow > 50%

MAPFACp= 1.0 + 0.005224 (P-100%)

._d 0.45 Turbine Bypass

)(

Operable or inoperable 0.40 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 Power (% Rated)

CP&L Nuclear Fu:Is Mgmt. & Safsty Analysis Design Calc. No. 2B21-0554 82C14 Core Operating Limits Report Page 22, Revision 0 Figure 11 GE13 Flow-Dependent MCPR Limit, MCPR(F) 1.80 I

For W (% Rated Core Flow) < 40%,

1.75 e

MCPR(F) = (AgWd100+Br)*[1+0.0032 (40-Wc)]

For W (% Rated Core Flow) 2 40%,

1,70 7

e MCPR(F) = Max (1,20, ArWc/100+Br) 1.65 Max Core Flow

(% Rated)

Ap Br 102.5

- 0.571 1.655 1.60 107.0

- 0.586 1.697 t

112.0

- 0.602 1.747 l

117.0

- 0.632 1.809

\\j -~ \\

1.50 3

C b 1.45

\\

U 1.40 i

1.35 Max Flow = 117%

j 11 2%

\\

1.30 107%

T' 102 5%

1.25

\\d 1.20 -

l This Figure is Referred To By 1.15 Technical Specification 3.2.2,3.4.1 and 3.7.6 1.10 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 Core Flow (% Rated)

CP&L Nucitar Fuels Mgmt. & Safety Analysis Design Calc. No. 2B21-0554 B2C14 Core Operating Limits Report Page 23, Revision 0 O

rio r 44-A10 Flow-Dependent MCPR Limit, MCPR(F) l l

l l

l l

l l

l l

l For We (% Rated Core Flow) < 40%,

1J/5 MCPR(F) = 1.026*(ArWc/100+Br)*(1+0.0032 (40-Wc))

For We (% Rated Core Flow) 240%,

k MCPR(F) = Maximum of 1.23 or 1.026*(A,W /100+Br) 1.70 c

Max Core Flow 1.65 -

(% Rated)

Ap Br 102.5

- 0.571 1.655 107.0

- 0.586 1.697

'l 112.0

- 0.602 1.747 1.60 117.0

- 0.632 1.809 1.55

-, N\\

1.50 e

s s s

j4 s.e 1.35 isiax Fiow u iii%

112%

1.30 107*/.

' 02.5%

1'25

\\

\\\\ \\\\

1.20 This Figure is Referred To By 1.15 Technical Specification 3.2.2,3.4.1 and 3.7.6 O

1.10 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 110 115 120 l

Core Flow (% Rated)

CP&L Nucl ar Fu:Is Mgrnt. & Safsty Analysis Design Calc. No. 2B21-0064 B2C14 Core Operating Limits Report Page 24, Revision 0 O

Pioure,2 Power - Dependent MCPR Limit, MCPR (P) 3.30 OLMCPR 3.20 Rated MCPR Multiplier (Kp) 3.10 3.00 j l Core Flow > 50%l I

Turbine Bypass l Operating Limit MCPR(P) = Kp* Operating Limit MCPR(100) 2.90 J -

-l Inoperable

[ 2.80 For P < 25% :

No Thermal Limits Monitoring Required VI 2.70 No Limits Specified N

o 2.60 Core Flow > 50%

For 25% s P s PeypAss :

Where PerpAss = 30%

1 Turbine Bypass Kp = Maxirnum of 1.481 or Kpm h 2.50 Operable For Core Flow s 50% & Turbine Bypass Operable, g

Kpg = [1.90 + 0.02 (30% - P)] / OLMCPR(100)

$ 2.40

-[

For Core Flow > 50% & Turbine Bypass Operable I

3 Kpg = [2.20 + 0.02 (30% P)] / OLMCPR(100) 3 0 2.30 For Core Flow s50% & Turbine Bypass Inoperable, g

g Kpg = [1.96 + 0.072 (30% - P)] / OLMCPR(100)

I x 2.20 For Core Flow > 50% & Turbine Bypass inoperable I

m=[.

+ 0.05 (30Yo N / ONR(100) lEore7low~s 57/ol 2.10 Turbine Bypass l For 30% s P < 45% :

in p ble 2.00 K, = 1.28 + 0.0134 (45% - P)

For 45% s P < 60% :

L90 n

g Kp = 1.15 + 0.00867 (60% - P)

I 1.80 Core Flow s 50%

For P 2 60% :

A

'E 1.70 Turbine Bypass Kp = 1.00 + 0.00375 (100% - P)

Operable E

.E 1.60 1.50 E 1.40 -

This Figure is Referred To By o

Technical Specification 3

\\

3.2.2, 3.4.1, 3.7.6 3

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