ML040770079

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to Section 1 of Core Operating Limits Report for LaSalle, Unit 1, Cycle 11.
ML040770079
Person / Time
Site: LaSalle Constellation icon.png
Issue date: 01/20/2004
From:
Exelon Generation Co, Exelon Nuclear
To:
Office of Nuclear Reactor Regulation
References
-RFPFR, L1C11
Download: ML040770079 (29)
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Technical Requirements Manual - Appendix I L1 C1 1 Core Operating Limits Report

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Section 1 Core Operating Limits Report For LaSalle Unit 1 Cycle 11 Revision 0

Technical Requirements Manual - Appendix I Ll C11 Core Operating Limits Report

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Issuance of Changes Summary Affected Affected Pages Summary of Changes Revision Date Sections All All Original Issue LaSalle Unit 1 Cycle 11 0 1/2004

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LaSalle Unit 1 Cycle 11 Hi Revision 0

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Table of Contents References......................................................................................................................... iv

1. Average Planar Linear Heat Generation Rate (3.2.1) .................. ........................ 1-1 1.1 Technical Specification Reference ... .......................................................... 1-1 1.2 Description ................................................................ 1-1
2. Minimum Critical Power Ratio (3.2.2) ................................................................ 2-1 2.1 Technical Specification Reference ............................................................. 2-1 2.2 Description ................................................................ 2-1
3. Linear Heat Generation Rate (3.2.3) ................................. ,,,.,.,,... 3-1 3.1 Technical Specification Reference .3-1 3.2 Description .,,, , 3-1
4. Control Rod Withdrawal Block Instrumentation (3.3.2.1) . ..................... 4-1 4.1 Technical Specification Reference ... 4-1 4.2 Description .4-1
5. Traversing In-Core Probe System (3.2.1, 3.2.2, 3.2.3) ........................................ 5-1 5.1 Technical Specification Reference .5-1 5.2 Description .5-1 5.3 Bases ......... , 5-1
6. Allowed Modes of Operation (B 3.2.2, B 3.2.3) ...................................... ,,,,,,. 6-1
7. Methodology (5.6.5) ................................................................ 7-1

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References

1. Exelon Generation Company, LLC Docket No. 50-373 LaSalle County Station, Unit 1, License No. NPF-11.
2. Letter from D. M. Crutchfield to All Power Reactor Licensees and Applicants, Generic Letter 88-16; Concerning the Removal of Cycle-Specific Parameter Limits from Tech Specs, October 3, 1988.
3. EMF-2690 Revision 0, "LaSalle Unit 1 Cycle 10 Reload Analysis," Framatome ANP, Inc.,

January 2002.

4. EMF-2563 (P) Revision 1, "Fuel Mechanical Design Report Exposure Extension for ATRIUM-9B Fuel Assemblies at Dresden, Quad Cities, and LaSalle Units," August 2001.
5. J11-03692-LHGR Revision 1, "ComEd GE9/GE10 LHGR Improvement Program," [NDIT NFMO00067 Sequence 00], February 2000.
6. Letter from A. Giancatarino to J. Nugent, "LaSalle Unit 1 and Unit 2 Rod Block Monitor COLR Setpoint Change," NFM:MW:01-0106, April 3, 2001.
7. Letter from R. G. Grummer to N. J. Carr, "Plant Startup Testing with POWERPLEX-Ill",

RGG:04:001, January 8, 2004.

8. Deleted
9. Deleted
10. NEDC-31531 P and Supplement 1, "ARTS Improvement Program Analysis for LaSalle Units 1 and 2," December 1993 and June 1998, respectively.
11. 24A5180AA Revision 0, "Lattice-Dependent MAPLHGR Report for LaSalle County Station Unit 1 Reload 7 Cycle 8," December 1995.
12. NFM Calculation No. BSA-L-99-07, "LaSalle GE9 MAPFACf Thermal Limit Multiplier for 105%

Maximum Core Flow," October 1999.

13. GE-NE-187-13-0792 Revision 2, "Evaluation of a Postulated Slow Turbine Control Valve Closure Event For LaSalle County Station Units 1 and 2," NDIT NFM-98-00146 Sequence 00, July 1998.
14. GNF 0000-0017-8285-SRLR,"Supplemental Reload Licensing Report for LaSalle County Nuclear Station Unit 1 Reload 10 Cycle 11," December 2003.
15. GE-NE-0000-0013-2777-00, "GE14 Fuel Design Cycle-Independent Analyses for LaSalle Unit 1,"

Revision 1, January 2004.

16. TGO-EXN-HA1 03-011, Letter from T. Orr to R. Chin, "Transmittal of Peak Pellet LHGR Limits for LaSalle 1 Cycle 11 GE14 Bundles with Gad Suppression," July 3, 2003.
17. "OPL3 Parameters for LaSalle Unit 1 Cycle 11 (GE14 Transition) Transient Analysis", TODI NF0300053 Sequence 01, September 25, 2003.

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18. "Final FRED for LaSalle Unit 1 Cycle 11 ", TODI NF 0300055 Seq. 0, June 23, 2003.
19. "LaSalle 1 Cycle 11 Final Core Operating Limits Report Markup", CJP-EXN-HA1 106, December 15, 2003.
20. "Exelon LaSalle Unit 1 SAFER/GESTR Loss-of-Coolant-Accident Analysis for GE14 Fuel", GE-NE-0000-0022-8684-RO, December 2003.
21. "Transient Analysis Evaluation for LaSalle 3 TCV Operation at Power Uprate and MELLLA Conditions", NFM:BSA:00-025, R. W. Tsai to D. Bost, April 13, 2000.
22. "LaSalle Units 1 and 2 Operation with One MSIV Closed", NF-MW:03-015, C. de la Hoz to K. W.

Peterman, January 15, 2003.

23. "LaSalle Units 1 and 2 Operation with One TSV OOS", EC Number 344312, August 18, 2003.
24. "LaSalle County Station Power Uprate Project, Task 201: Reactor Power/Flow Map", GE-NE-A1300384-07-01, Rev. 1, September 1999.
25. "Operation with a Pressure Regulator Out of Service at LaSalle", NF-MW:03-0063, Carlos de la Hoz to Kirk Peterman, February 7, 2003.
26. "Scram Times Versus Notch Position", DRF Al 2-00038-3, Vol. 4, G. A. Watford, May 22, 1992.

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1. Average Planar Linear Heat Generation Rate (3.2.1) 1.1 Technical Specification

Reference:

Section 3.2.1.

1.2

Description:

Tables 1-1 through 1-3 are used to determine the maximum average planar linear heat generation rate (MAPLHGR) limit for each fuel type. Limits given in Tables 1-1 through 1-3 are for dual reactor recirculation loop operation.

For single reactor recirculation loop operation (SLO), the MAPLHGR limits given in Tables 1-1 through 1-3 must be multiplied by a SLO MAPLHGR multiplier from Table 1-4.

Table 1-1 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for ATRIUM-10 Fuel ATRM1 0-P1 OCAZB404-15GZ-10OU-9WR-149-T6-2651 ATRM10-Pl OCAZB404-16GZ-10OU-9WR-149-T6-2652 (Reference 3)

Planar Average Exposure I MAPLHGR (GWd/MT) (kWIft__

l0.0 12.5 15.0 12.5 l 55.0 9.1 64.0 7.6 Table 1-2 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for ATRIUM-9B Fuel ATRM9-P9CATB393-16GZ-SPC100M-9WR-149-T6-3921 ATRM9-P9CATB396-12GZ-SPC100M-9WR-149-T6-3922 ATRM9-P9CATB384-1 1GZ-SPC80M-9WR-149-T6-3920 ATRM9-P9CATB396-12GZ-SPCIOOM-9VWR-149-T6-3923 (Reference 3)

Planar Average Exposure MAPLHGR (GWd/MT) (kWIft) 0.0 135 20.0 13.5 64.3 9.07 ErI]

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Table 1-3 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) for GE14C Fuel GE14-P1 OCNAB421 -18GZ-120T-150-T6-2673 GE1 4-PI OCNAB422-19GZ-1 20T-1 50-T6-2677 (Reference 14)

Planar Average MAPLHGR Exposure Limits (GWd/MT) (kWlft) l 0.00 13.4 16.00 13.4 63.50 8.0 70.00 5.0 Table 1-4 Maximum Average Planar Linear Heat Generation Rate (MAPLHGR)

SLO Multipliers for All Fuel Types (References 3 and 14)

Fuel Product Line SLOMuplier MAPLHGRl

[ __ __ __

ATRIUM-9B3 ATRIUM-10

__ __ __ _I M ultiplier 0.90 0.90 GE14C 0.78

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2. Minimum Critical Power Ratio (3.2.2) 2.1 Technical Specification

Reference:

Section 3.2.2.

2.2

Description:

The various MCPR limits are described below.

2.2.1 Manual Flow Control MCPR Limits The Operating Limit MCPR (OLMCPR) is determined from either section 2.2.1.1 or 2.2.1.2, whichever is greater at any given power and flow condition.

2.2.1.1 Power-Dependent MCPR For operation at greater than 25.0% core thermal power, the OLMCPR as a function of core thermal power is determined by multiplying the applicable rated condition OLMCPR limit shown in Table 2-1 or 2-2 by the applicable MCPR multiplier K(P) given in Table 2-3. Tables 2-1, 2-2, and 2-3 are applicable to all fuel types.

2.2.1.2 Flow-Dependent MCPR Table 2-4 gives the dual loop MCPRF limit as a function of the flow for most modes of operation. If 1 TCV or 1 TSV is stuck closed AND the combined steam flow capacity of the operating turbine control valves plus the bypass capacity (as limited by the Maximum Combined Flow Limiter) is less than 115% steam flow, the limits in Table 2-5 are applied. The limits for single loop operation are listed in Tables 2-6 and 2-7 and the same criteria as stated above for dual loop determine which table is applicable. The MCPRF limit determined from these tables is the flow dependent OLMCPR based on a maximum core flow of 105%.

2.2.2 Automatic Flow Control MCPR Limits Automatic Flow Control is not supported for LI 011.

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[H]I 2.2.3 Option A and Option B Option A and Option B refer to scram speeds.

Option A scram speed is the Improved Technical Specification scram speed.

Reload analyses performed by GNF for Cycle 11 Option A MCPR limits utilized a 20% core average insertion time of 0.900 seconds. Compliance with the surveillance requirements of Technical Specifications 3.1.3 and 3.1.4 assures that this insertion time is met.

To utilize the MCPR limits for the Option B scram speed, the core average scram insertion time for 20% insertion must be less than or equal to 0.694 seconds (Reference 17) (0.672 seconds at notch position 39, Reference 26). If the core average scram insertion time does not meet the Option B criteria, but meets Technical Specification 3.1.3 and 3.1.4 criteria, the appropriate MCPR value may be determined from a linear interpolation between the Option A and B limits with standard mathematical rounding to two decimal places. When performing a linear interpolation to determine MCPR limits, ensure that the time used for Option A is 0.900 seconds.

2.2.4 Recirculation Flow Control Valve Settings Cycle 11 was analyzed with a maximum core flow runout of 105%; therefore the recirculation flow control valve must be set to maintain core flow less than 105% (113.9 Mlb/hr) for all runout events (Reference 18). This value is consistent with the analyses of Reference 15.

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Table 2-1 MCPR Option A Based Operating Limits For All Fuel Types (References 14 and 15)

CycleE xposure EOOS Combination Fuel Type < EOR1 - 1984 > EOR - 1984 (MWdIMT) (MWd/MT)

ATRIUM 1.53 1.57 BASE GE14C 1.57 1.61 ATRIUM 1.54 1.58 BASE SLO GE14C 1.58 1.62 ATRIUM 1.53 1.68 RPTOOS GE14C 1.57 1.69 ATRIUM 1.54 1.69 RPTOOS SLO GE14C 1.58 1.70 ATRIUM 1.53 1.57 TBPOOS GE14C 1.57 1.61 ATRIUM 1.54 1.58 TBPOOS SLO GE14C 1.58 1.62 ATRIUM 1.53 1.68 TCV SLOW CLOSURE + RPTOOS GE14C 1.57 1.69 ATRIUM 1.54 1.69 TCV SLOW CLOSURE + RPTOOS SLO GE14C 1.58 1.70 ATRIUM 1.53 1.57 1TCV or 1TSV STUCK CLOSED GE14C 1.57 1.61 ATRIUM 1.54 1.58 1TCV or 1TSV STUCK CLOSED SLO GE14C 1.58 1.62 ATRIUM 1.53 1.61 PROOS GE14C 1.57 1.64 ATRIUM 1.54 1.62 PROOSSLO GE14C 1.58 1.65

1. EOR refers to the end of rated power (i.e. 100% power/1 00% flow operation with all rods out)
2. Note thatATRIUM refers to both Framatome-ANP ATRIUM-9B and ATRIUM-10 fuel types

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Table 2-2 MCPR Option B Based Operating Limits For All Fuel Types (References 14 and 15)

Cycle Exposure EOOS Combination Fuel Type2 < EOR1 -1984 > EOR1 -1984 MWd/MT MWd/MT ATRIUM 1.39 1.46 BASE GE14C 1.43 1.47 ATRIUM 1.40 1.47 BASE SLO GE14C 1.44 1.48 ATRIUM 1.42 1.51 RPTOOS GE14C 1.44 1.52 ATRIUM 1.43 1.52 RPTOOS SLO GE14C 1.45 1.53 ATRIUM 1.44 1.52 TBPOOS GE14C 1.47 1.54 ATRIUM 1.45 1.53 TBPOOS SLO GE14C 1.48 1.55 ATRIUM 1.42 1.51 TCV SLOW CLOSURE + RPTOOS GE14C 1.46 1.52 ATRIUM 1.43 1.52 TCV SLOW CLOSURE + RPTOOS SLO GE14C 1.47 1.53 ATRIUM 1.39 1.46 1TCV or 1TSV STUCK CLOSED GE14C 1.43 1.47 ATRIUM 1.40 1.47 1TCV or 1TSV STUCK CLOSED SLO GE14C 1.44 1.48 ATRIUM 1.39 1.46 PROOS GE14C 1.43 1.47 ATRIUM 1.40 1.47 PROOS SLO GE14C 1.44 1.48

1. EOR refers to the end of rated power (i.e. 100% power/100% flow operation with all rods out)
2. Note that ATRIUM refers to both Framatome-ANP ATRIUM-9B and ATRIUM-10 fuel types

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Table 2-3 MCPRp for All Fuel Types (Reference 15)

Core Thermal Power (% of Rated)

I 0 1 25 1 40 1 60 1 80 80 1 100 EOOS Combinations OLMCPR Multiplier, Kp Base 1.55 1.55 1.347 1.15 1.10 1.0 RPTOOS 1.55 1.55 1.347 1.15 1.10 1.0 TBPOOS 1.55 1.55 1.347 1.15 1.10 1.0 1TCV or ITSV Stuck 1.55 1.55 1.347 1.15 1.10 1.0 Closed RTPC T OO S 1.80 1.80 1.64 1.21 1.15 1.0 PROOS 1.80 1.80 1.64 1.21 1.15 1.0 Notes for Table 2-3

  • Values are to be linearly interpolated between relevant power levels
  • For thermal limit monitoring at greater than 100% core thermal power, the 100% core thermal power multiplier, Kp, should be applied.
  • Allowable EOOS conditions are listed in Section 6
  • MCPRp limits are independent of scram speed

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Table 2-4 MCPRF limits for All Fuel Types and All Operating Conditions, DLO, Except 1TCV or 1TSV Stuck Closed with less than 115% Steam Flow Capacity (Reference 15)

Flow(% of rated) l MCPRF 105.0 1.20 89.5 1.20 30.0 1.5567 0.0 1.5567

  • Values are to be linearly interpolated between relevant flow values.
  • The MCPRF limit is independent of scram speed.
  • This table is not applicable to 1 TCV or 1 TSV stuck closed with less than 115%

steam flow capacity operating conditions.

Table 2-5 MCPRF limits for All Fuel Types, DLO, with ITCV or 1TSV Stuck Closed with less than 115% Steam Flow Capacity (Reference 15)

Flow (% of rated) MCPRF 105.0 1.217 30.0 1.6667 0.0 1.6667

  • Values are to be linearly interpolated between relevant flow values.
  • The MCPRF limit is independent of scram speed.
  • This table is only applicable to 1 TCV or 1 TSV stuck closed with less than 115%

steam flow capacity operating conditions.

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Table 2-6 MCPRF limits for All Fuel Types and All Operating Conditions, SLO, Except 1TCV or 1TSV Stuck Closed with less than 115% Steam Flow Capacity (Reference 15)

Flow (% of rated) MCPRF 105.0 1.21 89.62 1.21 30.0 1.5667 0.0 1.5667

  • Values are to be linearly interpolated between relevant flow values.
  • The MCPRF limit is independent of scram speed.
  • This table is not applicable to 1 TCV or 1 TSV stuck closed with less than 115%

steam flow capacity operating conditions.

Table 2-7 MCPRF limits for All Fuel Types, SLO, with 1TCV or 1TSV Stuck Closed with less than 115% Steam Flow Capacity (Reference 15)

Flow (% of rated) l MCPRF 105.0 1.227 30.0 1.6767 0.0 1.6767

  • Values are to be linearly interpolated between relevant flow values.
  • The MCPRF limit is independent of scram speed.
  • This table is only applicable to 1 TCV or 1 TSV stuck closed with less than 115% steam flow capacity operating conditions.

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3. Linear Heat Generation Rate (3.2.3) 3.1 Technical Specification

Reference:

Section 3.2.3.

3.2

Description:

The LHGR Limit is the product of the LHGR Limit from Tables 3-1, 3-2, 3-3, 3-4, 3-5, 3-6, 3-7, or 3-8 and the minimum of either the power dependent LHGR Factor, LHGRFACp, the flow dependent LHGR Factor, LHGRFACF or the single loop operation (SLO) multiplication factor where applicable. The applicable power dependent LHGR Factor (LHGRFACp) is determined from Table 3-9, applicable to all fuel types. The applicable flow dependent LHGR Factor (LHGRFACF) is determined from either Table 3-10 or Table 3-11 applicable to all fuel types. The SLO multiplication factor can be found in Table 3-12 Table 3-1 Steady-State LHGR Limits for all ATRIUM-10 Fuel ATRM1 0-Pl OCAZB404-15GZ- 10OU-9WR-149-T6-2651 ATRM 10-Pl OCAZB404-16GZ-10OU-9WR-149-T6-2652 (Reference 3)

Planar Average Exposure (GWd/MT) 1 l

LHGR Limit (kW/ft) 0.0 13.4 15.0 13.4 55.0 9.1 64.0 7.3 Table 3-2 Steady-State LHGR Limits for ATRIUM-9B Fuel ATRM9-P9CATB393-16GZ-S PC 100M-9WR-149-T6-3921 ATRM9-P9CATB396-12GZ-SPC 100M-9WR-149-T6-3922 ATRM9-P9CATB384-1 1 GZ-SPC80M-9WR-1 49-T6-3920 ATRM9-P9CATB396-12GZ-SPC100M-9WR-149-T6-3923 (Reference 3)

Planar Average LHGR Limit Exposure (kW/ft)

(GWdI T)__ _ _ _ _ _ _

0.0 14.4 15.0 14.4 64.3 7.9

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Table 3-3 Steady-State LHGR Limits for GEI 4-P1 OCNAB421-18GZ-1 20T-1 50-T6-2673 Lattices 6095, 6096, 6097 and 6100 (Reference 16)

U02 Pellet LHGR Burnup Limit (GWd/MT) (kW/ft) 0.0000 13.4000

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Table 3-4 Steady-State LHGR Limits for GE14-Pl OCNAB421-1 8GZ-1 20T-1 50-T6-2673 Lattice 6098 (Reference 16)

U02 Pellet l LHGR Burnup Limit (GWd/MT) (kW/ft) 0.0000 13.4000

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Table 3-5 Steady-State LHGR Limits for GEI 4-PI OCNAB421-18GZ-1 20T-1 50-T6-2673 Lattice 6099 (Reference 16)

U02 Pellet LHGR Burnup Limit (GWd/MT) (kWlft) 0.0000 13.4000

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Table 3-6 Steady-State LHGR Limits for GEI 4-PI OCNAB422-19GZ-1 20T-1 50-T6-2677 Lattices 6095, 6118, 6119 and 6122 (Reference 16)

U02 Pellet LHGR Burnup Limit (GWd/MT) . (kW/ft) 0.0000 13.4000

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Table 3-7 Steady-State LHGR Limits for GEI 4-PI OCNAB422-19GZ-1 20T-1 50-T6-2677 Lattice 6120 (Reference 16)

U02 Pellet LHGR Burnup Limit (GWd/MT) (kW/ft) 0.0000 13.4000

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Table 3-8 Steady-State LHGR Limits for GE14-PlOCNAB422-19GZ-120T-150-T6-2677 Lattice 6121 (Reference 16)

U02 Pellet LHGR Burnup Limit (GWd/MT) (kW/ft) 0.0000 13.4000

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(( ]1 Table 3-9 LHGRFACp for All Fuel Types (Reference 15)

Core Thermal Power (% of Rated) 0 l 25 l 100 EOOS Combinations LHGR Multiplier, LHGRFAC(P)

Base 0.608 0.608 1.0 RPTOOS 0.608 0.608 1.0 TBPOOS 0.608 0.608 1.0 ITCV or 1TSV Stuck 068068 .

Closed 0.608 0.608 1.0 TCV Slow Closure + 0.40 0.40 1.0 RPTOOS PROOS 0.40 0.40 1.0 Notes for Table 3-9:

  • Values are to be linearly interpolated between relevant power levels.
  • For thermal limit monitoring at greater than 100% core thermal power, the 100% core thermal power LHGRFACp multiplier should be applied.
  • Allowable EOOS conditions are listed in Section 6.
  • The LHGRFACp multiplier is independent of scram speed.
  • The LHGR multiplier for any core power/flow condition is limiting of the LHGRFACp, LHGRFACF, and SLO Multiplier (if applicable).

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((11 Table 3-10 LHGRFACF Multipliers for All Fuel Types, For All Modes of Operation, Except 1 TCV or 1 TSV Stuck Closed with less than 115% Steam Flow Capacity (Reference 15)

Core Flow LHGRFACF

(% of rated) I Multiplier 105 1.00 75 1.00 30 0.55 0 0.55 Table 3-11 LHGRFACF Multipliers for All Fuel Types, With I TCV or 1 TSV Stuck Closed with less than 115% Steam Flow Capacity (Reference 15)

Core Flow LHGRFACF

(% of rated) Multiplier 105 1.00 89 1.00 30 0.41 0 0.41 Notes for Tables 3-10 and 3-1 1:

  • Values are to be linearly interpolated between relevant flow values.
  • LHGRFACF multipliers are independent of scram speed
  • For thermal limit monitoring above 105% rated core flow, utilize the 105% rated core flow LHGRFACF multiplier.
  • The LHGR multiplier for any core power/flow condition is the limiting of the LHGRFACp, LHGRFACF and SLO Multiplier (if applicable).

Table 3-12 LHGR SLO Multipliers For All Fuel Types (References 3 and 14)

Fuel Type l SLO LHGR I__ _ Multi her ATRIUM-9B 1.00 ATRIUM-10 1.00 GE14C 0.78 Notes for Table 3-12:

The LHGR multiplier for any core power/flow condition is the limiting of the LHGRFACp, LHGRFACF and SLO Multiplier (if applicable).

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4. Control Rod Withdrawal Block Instrumentation (3.3.2.1) 4.1 Technical Specification

Reference:

Table 3.3.2.1-1 4.2

Description:

The Rod Block Monitor Upscale Instrumentation Setpoints are determined from the relationships shown below (Reference 6):

ROD BLOCK MONITOR1 O V UPSCALE TRIP FUNCTION ALLOWABLE VALUE Two Recirculation Loop 0.66 Wd + 54%

Operation Single Recirculation Loop 0.66 Wd + 48.7%

Operation__ _ _ _ _ _ _ _ _ _

The setpoint may be lower/higher and will still comply with the Rod Withdrawal Error (RWE) Analysis because RWE is analyzed unblocked. The allowable value is clamped, with a maximum value not to exceed the allowable value for a recirculation loop flow (Wd) of 100%.

Wd - percent of recirculation loop flow required to produce a rated core flow of 108.5 Mlb/hr.

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5. Traversing In-Core Probe System (3.2.1, 3.2.2, 3.2.3) 5.1 Technical Specification

Reference:

Technical Specification Sections 3.2.1, 3.2.2, 3.2.3 for thermal limits require the TIP system for recalibration of the LPRM detectors and monitoring thermal limits.

5.2

Description:

When the traversing in-core probe (TIP) system (for the required measurement locations) is used for recalibration of the LPRM detectors and monitoring thermal limits, the TIP system shall be operable with the following:

1. movable detectors, drives and readout equipment to map the core in the required measurement locations, and
2. indexing equipment to allow all required detectors to be calibrated in a common location.

The following applies for use of the SUBTIP methodology:

With one or more TIP measurement locations inoperable, the TIP data for an inoperable measurement location may be replaced by data obtained from a 3-dimensional BWR core monitoring software system adjusted using the previously calculated uncertainties, provided the following conditions are met:

1. All TIP traces have previously been obtained at least once in the current operating cycle when the reactor core was operating above 20% power, (Reference 7) and
2. The total number of simulated channels (measurement locations) does not exceed 42% (18 channels).

Otherwise, with the TIP system inoperable, suspend use of the system for the above applicable monitoring or calibration functions.

5.3 Bases

The operability of the TIP system with the above specified minimum complement of equipment ensures that the measurements obtained from use of this equipment accurately represent the spatial neutron flux distribution of the reactor core. The normalization of the required detectors is performed internal to the core monitoring software system.

Substitute TIP data, if needed, is 3-dimensional BWR core monitoring software calculated data which is adjusted based on axial and radial factors calculated from previous TIP sets. Since the simulation and adjustment process could introduce uncertainty, a maximum of 18 channels may be simulated to ensure that the uncertainties assumed in the substitution process methodology remain valid.

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6. Allowed Modes of Operation (B 3.2.2, B 3.2.3)

The allowed modes of operation with combinations of equipment out-of-service are as described below:

Equipment Out of Service ELLLA MELLLA ICF Coastdown 3 O ptions1' , 2 _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

Base Case Yes Yes Yes Yes Base Case SLO Yes No N/A Yes RPTOOS Yes Yes Yes Yes RPTOOS SLO Yes No N/A Yes TBPOOS Yes Yes Yes Yes TBPOOS SLOI Yes No N/A Yes TCV Slow Closure + RPTOOS Yes Yes Yes Yes TCV Slow Closure + RPTOOS SLO Yes No N/A Yes 1 TCV or 1 TSV Stuck Closed Yes Yes Yes Yes 1 TCV or 1 TSV Stuck Closed SLO Yes No N/A Yes 4

PROOS Yes Yes Yes Yes PROOS SLo 4 Yes No N/A Yes ELLLA- Extended Load Line Limit Analysis MELLLA - Maximum Extended load Line Limit Analysis ICF - Increased Core Flow

' Each OOS option is applicable for Option A or B and may be combined with up to 2 TIP machines OOS (or less than TIP channels equivalent to 42% of the total number of channels with 100%

available at startup), feedwater temperature reduction of 100'F, and up to 50% of the LPRMs OOS with an LPRM calibration frequency of 1250 effective full power hours (EFPH) (1000 EFPH +25%).

2 A single MSIV may be taken OOS (shut) under any and all OOS Options, so long as core thermal power is maintained <75% of 3489 MWt (Reference 22).

3 Coastdown operation is defined as any cycle exposure beyond full power/flow, all rods out condition with plant power slowly lowering while core flow is held constant.

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4 For operation with a pressure regulator OOS (PROOS), Reference 25 should be reviewed. PROOS and TCV slow closure is NOT an analyzed OOS combination.

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7. Methodology (5.6.5)

The analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC, specifically those described in the following documents:

1. XN-NF-81-58 (P)(A), Revision 2 and Supplements 1 and 2, "RODEX2 Fuel Rod Thermal-Mechanical Response Evaluation Model," March 1984.
2. Letter from Ashok C. Thadini (NRC) to R.A. Copeland (SPC), "Acceptance for Referencing of ULTRAFLOWrM Spacer on 9x9-IX/X BWR Fuel Design," July 28,1993.
3. ANF-524 (P)(A) Revision 2 and Supplements 1 and 2, "ANF Critical Power Methodology for Boiling Water Reactors," November 1990.
4. XN-NF-80-19 (P)(A) Volume 1 Supplement 3, Supplement 3 Appendix F, and Supplement 4, "Advanced Nuclear Fuels Methodology for Boiling Water Reactors: Benchmark Results for CASMO-3G/MICROBURN-B Calculation Methodology," November 1990.
5. XN-NF-85-67 (P)(A) Revision 1, "Generic Mechanical Design for Exxon Nuclear Jet Pump BWR Reload Fuel," September 1986.
6. ANF-913 (P)(A) Volume 1 Revision 1, and Volume 1 Supplements 2, 3, 4, "COTRANSA2: A Computer Program for Boiling Water Reactor Transient Analyses," August 1990.
7. XN-NF-84-105 (P)(A), Volume 1 and Volume 1 Supplements 1 and 2; Volume 1 Supplement 4, "XCOBRA-T: A Computer Code for BWR Transient Thermal-Hydraulic Core Analysis," February 1987 and June 1988, respectively.
8. ANF-89-014 (P)(A) Revision 1 and Supplements 1 & 2, "Generic Mechanical Design for Advanced Nuclear Fuels Corporation 9X9 - IX and 9x9 - 9X BWR Reload Fuel," October 1991.
9. EMF-2209 (P)(A), Revision 1, "SPCB Critical Power Correlation," July 2000.
10. ANF-89-98 (P)(A), Revision 1 and Revision 1 Supplement 1, "Generic Mechanical Design Criteria for BWR Fuel Designs," May 1995.
11. ANF-91-048 (P)(A), "Advanced Nuclear Fuels Corporation Methodology for Boiling Water Reactors EXEM BWR ECCS Evaluation Model," January 1993.
12. Commonwealth Edison Company Topical Report NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods," Revision 0 and Supplements on Neutronics Licensing Analysis (Supplement 1) and La Salle County Unit 2 benchmarking (Supplement 2), December 1991, March 1992, and May 1992, respectively.
13. EMF-85-74 (P)(A) Revision 0 and Supplement 1(P)(A) and Supplement 2(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model," February 1998.
14. NEDE-24011-P-A-14, "General Electric Standard Application for Reactor Fuel (GESTAR)," June 2000.
15. EMF-CC-074 (P) Volume 4 Revision 0, "BWR Stability Analysis: Assessment of STAIF with Input from MICROBURN-B2," August 2000.

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16. ANF-1125 (P)(A) and ANF-1125(P)(A) Supplements 1 and 2, "ANFB Critical Power Correlation,"

Advanced Nuclear Fuels Corporation, April 1990.

17. ANF-1125 (P)(A) Supplement I Appendix E, "ANFB Critical Power Correlation Determination of ATRIUM TM -9B Additive Constant Uncertainties," September 1998.
18. EMF-1125 (P)(A) Supplement 1 Appendix C, "ANFB Critical Power Correlation Application for Co-Resident Fuel," August 1997.
19. Commonwealth Edison Topical Report NFSR-0085 Revision 0, "Benchmark of BWR Nuclear Design Methods," November 1990.
20. Commonwealth Edison Topical Report NFSR-0085 Supplement 1 Revision 0, "Benchmark of BWR Nuclear Design Methods - Quad Cities Gamma Scan Comparisons," April 1991.
21. Commonwealth Edison Topical Report NFSR-0085 Supplement 2 Revision 0, "Benchmark of BWR Nuclear Design Methods - Neutronic Licensing Analyses," April 1991.
22. ANF-CC-33(P)(A) Supplement 1 Revision 1 and Supplement 2, "HUXY: A Generalized Multirod Heatup Code with 10CFR50, Appendix K Heatup Option," August 1986 and January 1991, respectively.
23. XN-NF-80-19 (P)(A) Volume 4 Revision 1, "Exxon Nuclear Methodology for Boiling Water Reactors: Application of the ENC Methodology to BWR Reloads," June 1986.
24. XN-NF-80-19 (P)(A) Volume 3 Revision 2, "Exxon Nuclear Methodology for Boiling Water Reactors, THERMEX: Thermal Limits Methodology Summary Description," January 1987.
25. ANF-91-048 (P)(A) Supplement I and Supplement 2, "BWR Jet Pump Model Revision for RELAX,"

October 1997.

26. XN-NF-80-19 (P)(A) Volumes 2, 2A, 2B, and 2C, "Exxon Nuclear Methodology for Boiling Water Reactors: EXEM BWR ECCS Evaluation Model," September 1982.
27. XN-NF-80-19 (P)(A) Volume 1 and Supplements 1 and 2, "Exxon Nuclear Methodology for Boiling Water Reactors - Neutronic Methods for Design and Analysis," March 1983.
28. NEDE-24011-P-A-14, June 2000 and the U.S. Supplement NEDE-24011-P-A-US, June 2000, "General Electric Standard Application for Reactor Fuel".
29. EMF-2158(P)(A), Revision 0, "Siemens Power Corporation Methodology for Boiling Water Reactors: Evaluation and Validation of CASMO-4/MICROBURN-B2", Siemens Power Corporation, October 1999.
30. NEDC-32981P(A), Revision 1, "GEXL96 Correlation for ATRIUM-9B Fuel", May 2002.
31. NEDC-33106P, Revision 1, "GEXL97 Correlation for ATRIUM-1 0 Fuel", June 2003.

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