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GNF. Gloal Nuclear Fuel

A .Joint Venture of GE. Toshiba. & Hitachi

DOCKET NUMBER • .... -PROD. & UTIL FAG, 5O- 61--4

0000-0035-6443-SRLR Revision 0

Class I December 2005

Supplemental Reload Licensing Report

for

Vermont Yankee Nuclear Power Station

Reload 24 Cycle 25

(with Extended Power Uprate)

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OFFICE OF SECRETARYRULEMAKINGS ANDADJUDICATIONS STAFF

VERMONT YANKEE 0000-0035-6443-SRLR Reload 24 Revision 0

Important Notice Regarding Contents of This Report

Please Read Carefully

This report was prepared by Global Nuclear Fuel - Americas, LLC (GNF-A) solely for use by Entergy Nuclear Vermont Yankee ("Recipient") in support of the operating license for Vermont Yankee (the "Nuclear Plant"). The information contained in this report (the "Information") is believed by GNF-A to be an accurate and true representation of the facts known by, obtained by or provided to GNF-A at the time this report was prepared.

The only undertakings of GNF-A respecting the Information are contained in the contract between Recipient and GNF-A for nuclear fuel and related services for the Nuclear Plant (the "Fuel Contract") and nothing contained in this document shall be construed as amending or modifying the Fuel Contract. The use of the Information for any purpose other than that for which it was intended under the Fuel Contract, is not authorized by GNF-A. In the event of any such unauthorized use, GNF-A neither (a) makes any representation or warranty (either expressed or implied) as to the completeness, accuracy or usefulness of the Information or that such unauthorized use may not infringe privately owned rights, nor (b) assumes any responsibility for liability or damage of any kind which may result from such use of such information.

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VERMONT YANKEE Reload 24


Acknowledgement

The engineering and rejoad licensing analyses, which form the technical basis of this Supplemental Reload Licensing Report, were performed by GNF - Fuel Engineering Services and GE Energy - NSA personnel. The Supplemental Reload Licensing Report was prepared by L. A. Leatherwood. This document has been verified by G. N. Marrotte.

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The basis for this report is General Electric Standard Application for Reactor Fuel, NEDE-2401 I-P-A

14, June 2000; and the U.S. Supplement, NEDE-2401 I-P-A-14-US, June 2000.

1. Plant-unique Items

Appendix A: Appendix B: Appendix C: Appendix D: Appendix E: Appendix F:

Analysis Conditions Decrease in Core Coolant Temperature Events ARTS Power and Flow Operating Limits Adjustments Implementation of Extended Power Uprate (EPU) Stability Solution Option 1-D Exclusion and Buffer Regions at EPU Condition List of Acronyms

2. Reload Fuel Bundles

FuelypeCycle Number Fuel Type Loaded

Irradiated:

GE 14-P 10DNAB394-7G5.0/6G4.0- 1 OOT- 1 50-T6-2566 (GE I 4C) 23 92

GE 14-P 1 ODNAB394-8G5.0/6G4.0- I OOT- I 50-T6-2595 (GE I 4C) 23 16

GEl 4-P 1ODNAB394-12G5.0-1OOT- 150-T6-2596 (GE14C) 23 20

GEl 4-P 1 ODNAB426-16G6.0- I OOT- I 50-T6-2682 (GEl 4C) 24 32

GEl 4-P 1 ODNAB390-14GZ- I OOT- I 50-T6-2683 (GE 14C) 24 44

GE 14-P 1 ODNAB3 88-17GZ- I O1T- I 50-T6-2684 (GE 14C) 24 40

New:

GE 14-P 1ODNAB383-1 7G6.0- I OOT- 150-T6-2865 (GE14C) 25 16

GE14-P 1ODNAB383- 13G6.0-1OOT- 150-T6-2863 (GE14C) 25 28

GE 14-P 1ODNAB383-14G6.0-100T- 150-T6-2864 (GE 14C) 25 32

GE 14-P 1 ODNAB422-16GZ- I OOT- I 50-T6-2862 (GE 14C) 25 48

Total: 368

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3. Reference Core Loading Pattern

Core Average Cycle Average Exposure Exposure

27305 MWd/MT 12153 MWd/MT

Nominal previous end-of-cycle exposure: (24771 MWd/ST) (11025 MWd/ST)

Minimum previous end-of-cycle exposure (for cold 26976 MWd/MT 11823 MWd/MT shutdown considerations): (24472 MWd/ST) (10726 MWd/ST)

14829 MWd/MT 0 MWd/MT

Assumed reload beginning-of-cycle exposure: (13453 MWd/ST) (0 MWd/ST)

Assumed reload end-of-cycle exposure (rated 29093 MWd/MT 14264 MWd/MT conditions): (26393 MWd/ST) (12940 MWd/ST)

Reference core loading pattern: Figure 1

4. Calculated Core Effective Multiplication and Control System Worth - No Voids, 200 C

Beginning of Cycle, keffective

Uncontrolled 1.118

Fully controlled 0.953

Strongest control rod out 0.988

R, Maximum increase in strongest rod out reactivity during the cycle (Ak) 0.0001

14264 MWd/MT

Cycle average exposure at which R occurs (12940 MWd/ST)

5. Standby Liquid Control System Shutdown Capability

Boron (ppm) Shutdown Margin (Ak)

(at 20(C) (at 160°C, Xenon Free) Analytical Requirement Achieved

800 _>0.010 0.037

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6. Reload Unique GETAB Anticipated Operational Occurrences (AOO) Analysis

Initial Condition Parameters

Operating domain: ICF (HBB) Exposure range : BOC to MOC (Application Condition: 1)

Peaking Factors

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow Design___ _(MWt) (1000 lb/hr) MCPR GE14C 1.45 1.43 1.29 1.040 7.244 110.6 1.35

Operating domain: ICF (HBB) Exposure range : MOC to EOC (Application Condition: 1)

Peaking Factors

Fuel Bundle Bundle Initial Local Radial Axial R-Factor Power Flow Design (MWt) (1000 lb/hr) MCPR

GEI4C 1.45 1.42 1.31 1.040 7.159 111.9 1.34

Operating domain: MELLLA (HBB) Exposure range : BOC to MOC (Application Condition: 1)

Peaking Factors

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow Design__ __(MWt) (1000 lb/hr) MCPR

GE14C 1.45 1.41 1.28 1.040 7.125 102.0 1.33

Operating domain: MELLLA (HBB) Exposure range : MOC to EOC (Application Condition: 1 )

Peaking Factors

Fuel Bundle Bundle Initial Design Local Radial Axial R-Factor Power Flow Design__ __(MWt) (1000 lb/hr) MCPR GE14C 1.45 1.40 1.30 1.040 7.054 103.0 1.33

' Exposure range designation is defined in Table 7-1. Application condition number is defined in Section 11.

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0000-0035-6443-SRLR

Operating domain: ICF (UB) Exposure range : MOC to EOC (Application Condition: 1 )

Peaking Factors

Fuel Bundle Bundle Initial Local Radial Axial R-Factor Power Flow

Design (MWt) (1000 lb/hr) MCPR

GE14C 1.45 1.49 1.31 1.040 7.556 107.5 1.34

Operating domain: MELLLA (UB) Exposure range : MOC to EOC (Application Condition: 1)

Peaking Factors

Fuel Bundle Bundle Initial

Design Local Radial Axial R-Factor Power Flow

Design (MWt) (1000 lb/hr) MCPR GE14C 1.45 1.49 1.42 1.040 7.518 98.3 1.31

7. Selected Margin Improvement Options 2

Recirculation pump trip:

Rod withdrawal limiter:

Thermal power monitor:

Improved scram time:

Measured scram time:

Exposure dependent limits:

Exposure points analyzed:

No

No

No

Yes (ODYN Option B)

No

Yes

2

Table 7-1 Cycle Exposure Range Designation

Name Exposure Range 3

BOC to MOC BOC25 to EOR25-1213 MWd/MT (I 100 MWd/ST)

MOC to EOC EOR25-1213 MWd!MT (1100 MWd/ST) to EOC25

BOC to EOC BOC25 to EOC25

2 Refer to GESTAR for those margin improvement options that are referenced and supported within GESTAR. 3 End of Rated (EOR) is defined as the cycle average exposure corresponding to all rods out, 100% power/100% flow, and normal feedwater temperature.

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8. Operating Flexibility Options 4

The following information presents the operational domains and flexibility options which are supported

by the reload licensing analysis. Inclusion of these results in this report is not meant to imply that these domains and options have been fully licensed and approved for operation.

ExtendedOperating Domain (EOD): Yes

EOD type: Maximum Extended Load Line Limit (MELLLA)

Minimum core flow at rated power: 99.0 %

Increased Core Flow: Yes

Flow point analyzed throughout cycle: 107.0 %

Feedwater Temperature Reduction: No

ARTS Program: Yes

Single Loop Operation: Yes

Equipment Out of Service:

Safety/relief valves Out of Service: Yes (credit taken for 3 of 4 relief valves (I RV OOS))

9. Core-wide AOO Analysis Results 5

Methods used: GEMINI, GEXL-PLUS

Operating domain: ICF (HBB) Exposure range : BOC to MOC (Application Condition: I)

Uncorrected ACPR

Event Flux Q/A GEI4C Fig.

(%rated) (%rated)

FW Controller Failure 354 121 0.26 2

Load Rejection w/o Bypass 382 119 0.28 3

Turbine Trip w/o Bypass 372 118 0.27 4

Inadvertent HPCI /L8 347 123 0.27 5

4 Refer to GESTAR for those operating flexibility options that are referenced and supported within GESTAR. 5 Exposure range designation is defined in Table 7-1. Application condition number is defined in Section 11.

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Operating domain: ICF (HBB)

Exposure range : MOC to EOC (Application Condition: I)

Uncorrected ACPR

Event Flux Q/A GEI4C Fig. (%rated) (%rated)





Operating domain: MELLLA (HBB)

Exposure range : BOC to MOC (Application Condition: 1)

Uncorrected ACPR

Event Flux Q/A GEI4C Fig. (%rated) (%rated)




Inadvertent HPCI/L8 306 121 0.25 13

Operating domain: MELLLA (HBB) Exposure range : MOC to EOC (Application Condition: I )

Uncorrected ACPR

Event Flux Q/A GE14C Fig.

(%rated) (%rated)





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Operating domain: ICF (UB)


Uncorrected ACPR

Event Flux Q/A GE14C Fig. (%rated) (%rated)




Inadvertent HPCI/L8 247 118 0.26 21

Operating domain: MELLLA (UB)

Exposure range : MOC to EOC (Application Condition: 1)

Uncorrected ACPR

Event Flux Q/A GE14C Fig. (%rated) (%rated)





10. Local Rod Withdrawal Error (With Limiting Instrument Failure) AOO Summary

Rod withdrawal error (RWE) limits with ARTS are reported in Vermont Yankee Nuclear Power Station APRM/RBMiTechnical Specifications / Maximum Extended Load Line Limit Analysis (ARTS/MELLLA), NEDC-33089P, March 2003. A statistically based RWE limit of 1.40 is established in the Statistically Based Rod Withdrawal Error Analysis for Vermont Yankee Nuclear Power Station, GE-NE-0000-00 16345 1-RO, July 2003.

A cycle specific analysis was performed for Vermont Yankee Cycle 25 to determine the MCPR corresponding to full withdrawal. (RBM was not credited in this analysis.) For the exposure range from BOC25 to EOC25, it is concluded that the statistically based RWE analysis value of 1.40 bounds the Cycle 25 specific analysis value. Therefore, it is the statistically based value that is reported in Section I I of the SRLR.

The RBM operability requirements specified in Section 3.4 of ARTS Report NEDC-33089P have been evaluated and shown to be sufficient to ensure that the Safety Limit MCPR and cladding 1% plastic strain criteria will not be exceeded in the event of an unblocked RWE event.

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11. Cycle MCPR Values 678

Two ioop operation safety limit:

Single loop operation safety limit:

Stability MCPR Design Basis:

ECCS MCPR Design Basis:

Non-pressurization events:

1.07

1.09

See Section 15

See Section 16 (Initial MCPR)

Exposure range: BOC to EOCGEI4C

Loss of Feedwater Heating (See Appendix B) 1.20

Rod Withdrawal Error (full withdrawal) 1.40

Fuel Loading Error (misoriented) 1.28

Limiting Pressurization Events OLMCPR Summary Table:'

Appl. Exposure Range Option A Option B Cond. I

GEI4C GEI4C

I Normal Operation (w/ equipment in-service) BOC to MOC 1.47 1.36

-IMOC to EOC 1.57 1.40

6 The two loop and single loop Safety Limit values include a 0.02 additional bundle uncertainty at the EPU

condition as required by NRC and documented in VY Cycle 25 Extended Power Uprate (EPU) Safetiy Limit Minimum Critical Power Ratio (SLMCPR), Letter, B. Vita (VY) to C. Collins (GNF), NEA-05-067, November 30, 2005. This restriction will remain in place until such time that NRC removes it.

Exposure range designation is defined in Table 7-1. 8 For single loop operation, the MCPR operating limit is 0.02 greater than the two loop value. 9 Each application condition (Appl. Cond.) covers the entire range of licensed flow and feedwater temperature unless specified otherwise. The OLMCPR values presented apply to rated power operation based on the two loop operation safety limit MCPR.

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Pressurization events: 10

Operating domain: ICF (HBB) Exposure range : BOC to MOC (Application Condition: 1)

Option A Option B

GE14C GEI4C

FW Controller Failure 1.46 1.35

Load Rejection w/o Bypass 1.47 1.36

Turbine Trip w/o Bypass 1.46 1.35

Inadvertent HPCI /L8 1.47 1.36

Operating domain: ICF (HBB) Exposure range : MOC to EOC (Application Condition: 1)

Option A Option B

GE14C GEI4C





Operating domain: MELLLA (HBB) Exposure range : BOC to MOC (Application Condition: 1)

Option A Option B

GEI4C GEI4C





'0 Application condition numbers shown for each of the following pressurization events represent the application

conditions for which this event contributed in the determination of the limiting OLMCPR value.

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Operating domain: MELLLA (HBB) Exposure range : MOC to EOC (Application Condition: I)

Option A Option B

GEI4C GEI4C





Operating domain: ICF (UB)


Option A Option B

GEI4C GEI4C





Operating domain: MELLLA (UB)


Option A Option B

GE14C GEI4C





12. Overpressurization Analysis Summary

Event Psi Pdome Pv Plant

(psig) (psig) (psig) Response

MSIV Closure (Flux Scram) - ICF (HBB) 1302 1303 1328 Figure 26

MSIV Closure (Flux Scram) - MELLLA 1299 1300 1324 Figure 27 (HBB) I f

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13. Loading Error Results

Variable water gap misoriented bundle analysis: Yes"

Misoriented Fuel Bundle ACPR

GE 14-P 1 ODNAB426-16G6.0-1 OOT- I 50-T6-2682 (GE I 4C) 0.07

GE 14-P 1 ODNAB390-14GZ- IOOT- 150-T6-2683 (GE 14C) 0.20

GEl 4-P 1ODNAB422-16GZ-1OOT-150-T6-2862 (GEl4C) 0.07

GE14-PI ODNAB383-14G6.0-1OOT-150-T6-2864 (GE14C) 0.09

GE 14-P 1 ODNAB3 88-17GZ- I OOT- 1 50-T6-2684 (GE I 4C) 0.21

GE14-PI ODNAB383-13G6.0-1OOT-150-T6-2863 (GEl 4C) 0.08

GE 14-P 1 ODNAB383-17G6.0-1OOT- 150-T6-2865 (GE 14C) 0.08

14. Control Rod Drop Analysis Results

This is a banked position withdrawal sequence plant, therefore, the control rod drop accident analysis is not required. NRC approval is documented in NEDE-2401 I-P-A-US.

15. Stability Analysis Results

Vermont Yankee has implemented the Option l-D solution documented in Reference 15.1. The NRC approved ODYSY methodology (Reference 15.2) has been applied to this reload. ODYSY applications offer the benefit of more accurate simulations of BWR stability events and conditions. Option I-D has (1) "prevention" elements (Exclusion and Buffer Regions) and (2) a "detect & suppress" element (MCPR safety limit (MCPRSL) protection provided by the flow-biased APRM flux trip for the dominant corewide mode of coupled thermal-hydraulic/neutronic reactor instability). Core and hot channel decay ratio

calculations (Reference 15.3) to determine the Exclusion Region and additional bases demonstrate that core-wide is the predominant reactor instability mode for Vermont Yankee. The detect and suppress calculation (Reference 15.4) consists of (A) calculation of a 95% probability/95% confidence level statistically-based hot channel oscillation magnitude for anticipated core-wide mode reactor instability and (B) calculation of the stability-based Operating Limit MCPR (OLMCPR) which provides 95/95 MCPRSL protection.

The detect and suppress calculation requires the use of the Delta CPR over Initial MCPR Versus the

Oscillation Magnitude (DIVOM) curve. Recent TRACG evaluations by GE have shown that the generic core-wide DIVOM curve specified in Reference 15.5 may not be conservative for current plant operating conditions for plants that have implemented Stability Option l-D. Specifically, a non-conservative deficiency has been identified for high power-to-flow ratios in the generic core-wide mode DIVOM

11 Includes a 0.02 penalty due to variable water gap R-factor uncertainty.

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curve. The deficiency results in a non-conservative slope of the associated core-wide DIVOM curve so that the APRM flux trip setpoint is too high. GE had made a Part 21 Notification on this issue. For Option I-D plants, the applicability of the core-wide mode DIVOM curve may be determined by comparing the core average power-to-flow ratio following a simulated flow runback on the rated rod line to approximately 30% of rated core flow to a value of 66 (MWt/Mlbm/hr) (Reference 15.6). If the core average power-to-flow ratio exceeds this value, then the generic core-wide mode DIVOM curve is not

applicable and appropriate corrective actions should be taken. For Vermont Yankee, the calculated core average power-to-flow ratio is 71.7 (MWt/Mlbm/hr) for Cycle 25 EPU conditions, which confirms that the core-wide mode generic DIVOM slope value of 0.175 is not applicable. Therefore, a conservative DIVOM slope of 0.35 (twice the generic DIVOM slope of 0.175) was selected for Cycle 25 EPU operation.

(1) The Exclusion Region and Buffer Region were calculated for Cycle 25 operation. The regions are

based on Reference 15.3 and are provided in Appendix E for EPU conditions.

(2A) The hot channel oscillation magnitudes at natural circulation, 45% rated core flow, and 55% rated

core flow were determined for Cycle 25 EPU conditions. Revised values for 45% rated core flow and 55% rated core flow are used in the calculation of the stability-based OLMCPR.

(2B) The stability-based OLMCPR was calculated for Cycle 25 EPU conditions. The calculation

demonstrated that reactor stability does not produce the limiting OLMCPR for Cycle 25 EPU operation. It is shown that the rated OLMCPR, OLMCPR(1 00%P/ I 00%F), is greater than the OLMCPR for a twopump trip scenario, OLMCPR(2PT), the OLMCPR at 100% rod line and 45% rated core flow,

OLMCPR(100%RL/45%F), is greater than the OLMCPR at steady state, OLMCPR(SS,45%F), and that the OLMCPR at 100% rod line and 55% rated core flow, OLMCPR(100%RL/55%F), is greater than the OLMCPR at steady state, OLMCPR(SS,55%F). For this analysis, the rated OLMCPR, OLMCPR(100%RL/45%F), and OLMCPR(100%RL/55%F) values are used for comparison (Reference 15.4). For all scenarios considered, the criteria are met at EPU conditions.

OLMCPR(I00%P/100%F) > OLMCPR(2PT)

OLMCPR(100%RL/45%F) > OLMCPR(SS,45%F)

OLMCPR(100%RL/55%F) > OLMCPR(SS,55%F)

References:

15.1 Application of the "Regional Exclusion with Flow-Biased APRM Neutron Flux Scram" Stability Solution (Option I-D) to the Vermont Yankee Nuclear Power Plant, Licensing Topical Report, GENE-637-018-0793, July 1993.

15.2 ODYSY Application for Stability Licensing Calculations, Licensing Topical Report, NEDC32992P-A, July 2001.

15.3 Vermont Yankee Cycle 25 Option I-D Stability Exclusion Region Analysis (EPU), GE-NE-0000-0043-577 I-RO, December 2005.

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15.4 Vermont Yankee Cycle 25 Option I-D Stability Detect and Suppress Analysis (EPU), GE-NE-0000-0044-5878-RO, December 2005.

15.5 Reactor Stability Detect and Suppress Solutions Licensing Basis Methodology for Reload

Applications, Licensing Topical Report, NEDO-32465-A, August 1996.

15.6 Determination of Figure of Merit for Stability DIVOM OGO1-0228-001, July 16, 2001.

Curve Applicability,

16. Loss-of-Coolant Accident Results

16.1 10CFR50.46 Licensing Results

The ECCS-LOCA analysis is based on the SAFER/GESTR-LOCA methodology. The licensing results applicable to all fuel types in the new cycle are summarized in the following table:

Table 16.1-1 Licensing Results

Core-Wide Licensing Local Ce-Wide

Fuel Type Basis PCT Oxidation Reatio (OF) (%)Reaction (°F) (%) (%)

GE14C 1960 < 3.00 < 0.10

The SAFER/GESTR-LOCA analysis results for GE14C fuel type are documented in Section 16.4, Reference I.

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16.2 10CFR50.46 Error Evaluation

The IOCFR50.46 errors applicable to the Licensing Basis PCT are shown in the tables below.

Table 16.2-1 Impact on Licensing Basis Peak Cladding Temperature for GEI4C

I 0CFR50.46 Error Notifications

PCT Impact (OF)

2003-05 Impact of postulated hydrogen-oxygen recombination 0 on PCT.

STotal PCT Adder (*F) 0

The GE14C Licensing Basis PCT remains below the IOCFR50.46 limit of 2200'F.

16.3 ECCS-LOCA Operating Limits

The ECCS MAPLHGR operating limits for the fuel bundles in this cycle are shown in the tables below.

Table 16.3-1 MAPLHGR Limits

Bundle Type: GE 14-P 1 ODNAB426-16G6.0- 100T- 150-T6-2682 (GEl 4C)

Average Planar Exposure MAPLHGR Limit

GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00


Bundle Type: GE 14-Pi ODNAB390-14GZ- IOOT- 150-T6-2683 (GE 14C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00

Page 17




Bundle Type: GE 14-P 1ODNAB422-16GZ- 1OOT- 150-T6-2862 (GE 14C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00

Table 16.34 MAPLHGR Limits

Bundle Type: GE14-PiODNAB383-14G6.0-IOOT-150-T6-2864 (GEI4C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00


Bundle Type: GE1 4-PI ODNAB388-17GZ-1 OOT-150-T6-2684 (GEl 4C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00


Bundle Type: GE 14-Pi ODNAB383-13 G6.0-1 OOT- 150-T6-2863 (GE 14C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00

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Bundle Type: GE 14-P 1ODNAB383-17G6.0-1OOT- 150-T6-2865 (GE 14C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00


Bundle Type: GE 14-P 1ODNAB394-7G5.0/6G4.0-1OOT- 150-T6-2566 (GE 14C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00


Bundle Type: GEl 4-P 1ODNAB394-8G5.0/6G4.0-1OOT- 150-T6-2595 (GE 14C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00

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Bundle Type: GE14-PIODNAB394-12G5.0-1OOT-150-T6-2596 (GEI4C)


GWd/MT GWd/ST kW/ft

0.00 0.00 12.82

21.08 19.12 12.82

63.50 57.61 8.00

70.00 63.50 5.00

The single-loop operation multiplier on PLHGR and MAPLHGR and ECCS analytical Initial MCPR values applicable to each fuel type in the new cycle core are shown in the table below.

Table 16.3-5 Initial MCPR and Single Loop Operation PLHGR and MAPLHGR Multiplier

Single Loop Operation Fuel Type Initial MCPR PLHGR and MAPLHGR

Multiplier

GEI4C 1.275 0.82

16.4 References

The SAFER/GESTR-LOCA analysis report applicable to the new cycle core is listed below for each fuel type. The report is based on a power level of 1912 MWt. ECCS-LOCA results at this power level bound any lower power levels.

References for GEI4C

1. Entergy Nuclear Operations Incorporated Vermont Yankee Nuclear Power Station Extended Power Uprate ECCS-LOCA SAFER/GESTR, GE-NE-0000-0015-5477-01, July

2003.

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44

42

40

38

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34

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1 3 5 7 9 E] 1 15 17 19 21 23 25 27 29 31 33 35 37 39 41 43

Fuel Type

A=GE14-PI ODNAB426-16G6.0-100T- 150-T6-2682 (Cycle 24) F=GE] 4-PI ODNAB383-13G6.0- OOT- 150-T6-2863 (Cycle 25) B=GE]4-PIODNAB390-14GZ-IOOT-150-T6-2683 (Cycle 24) G=GE14-PIODNAB383-17G6.0-1OOT-150-T6-2865 (Cycle 25) C=GE]4-PIODNAB422-16GZ-IOOT-150-T6-2862 (Cycle 25) H=GE14-PIODNAB394-7G5.0/6G4.0-1OOT-150-T6-2566 (Cycle 23) D=GE14-PIODNAB383-14G6.0-1 OOT- I 50-T6-2864 (Cycle 25) I=GE I4-PI ODNAB394-8G5.0/6G4.0-10OT-1 50-T6-2595 (Cycle 23) E=GE14-PIODNAB388-17GZ-1OOT-150-T6-2684 (Cycle 24) J=GE 14-PIODNAB394-12G5.0-1OOT-150-T6-2596 (Cycle 23)

Figure 1 Reference Core Loading Pattern

Page 21



Iw U

0.0 10.0 20.0 30.0

Tine (sec)0.0 10.0 20.0 30.0

Tam (sec)

150.0

0 U

0.0 10.0 20.0 Tmfe (sec)

30.0 0.0 10.0 20.0 Time (sec)

30.0

Figure 2 Plant Response to FW Controller Failure (MOC ICF (HBB))

Page 22



150.0

l100.0 .-I

200.0

NU

-a- Dome Press Rise (psi) -- Safety Va Row -.-Relief Vave Flow

Bypass Valve FRow

100.01

0.0 I0.0 6.0 0.03.0

Time (sec)3.0

Thn=e (sec)&0

100.0

8

a C

2 0

0n 3.0 6.0 0.0 3.0 6.0 Tim- 5Cc) Tfim (sec)

Figure 3 Plant Response to Load Rejection w/o Bypass (MOC ICF (HBB))

Page 23




3C

12100.0

U•2

--- Dorne Press Rise (psi) a fety Valfe Flow 0- Relief Va Flow

-. Bypass Valve Row

00

0.0

0.0 - -I* -

I1

0.0 3.0

Tum= (sec)6.0 0.0 3.0

Time (sec)6.0

E

a 9U

-100.0 40.0 3.0 6.0 0.0 3.0

Time (sec) Time (sec)6.0

Figure 4 Plant Response to Turbine Trip w/o Bypass (MOC ICF (HBB))

Page 24

VERMONT YANKEE R•Inad 94A

0000-0035-6443-SRLR Reviqion 0

Reload 24

2MO.0

150.0

S10D.0

50.0

0.0

150.0

100.0

50.0

0.0

-e W~uton FILtz' 10 -- A~e Surfac Float RwLD .~Core Let Flow -- Core 'Let Subcooling

0.0 10.0 20.0 30.0 40.0 50.0

Time (sec)0.0 10.0 20.0 30.0 40.0 50.0

Time (sec)

U

5

C

U

0.0 10.0 20.0 30.0 40.0 50.0

Tine (sec)0.0 10.0 20.0 30.0 40.0 50.0

Tfm• (sec)

Figure 5 Plant Response to Inadvertent HPCI L8 (MOC ICF (HBB))

Page 25



U

0.04 I

0.0 10.0 20.0 30.0

Tune (sec)0.0 10.0 2D.0 30.0

Time (Sec)

150.0

V. 2

0

50.0

0.00.0 10.0 20.0

Tune (see)30.0 0.0 10.0 20.0 30.0

Time (sec)

Figure 6 Plant Response to FW Controller Failure (EOC ICF (HBB))

Page 26



0.0 3.0

Time (sec)

100.0

- 0.0 6.0 0.0

1.0

0 .0

-2.0

6.0 0.0

&( Tuner (Sec)

6.0

U

0.0 3.0 Tune (sec)

3.0 Time (sec)

6.0

Figure 7 Plant Response to Load Rejection w/o Bypass (EOC ICF (HBB))

Page 27

VERMONT YANKEE 0000-0035-6443-SRLRw

3C

lf3102

-e-- Dome Press Rise (psi) Sa- .fety Vah Flow o.o.- Relief VaFlow

-- Bypass WV-e Row

0.0

0.0

0.0 3.0

Tune (see)6.0 0.0 &e

Tinm (sec)6.0

I U

6

I

0.0 3.0 Tune (sec)

6.0 0.0 3.0 Time (sec)

6.0

Figure 8 Plant Response to Turbine Trip w/o Bypass

(EOC ICF (HBB))

Page 28



V

0.0 10.0 20.0 30.0 40.0 50.0

Time (sec)0.0 10.0 20.0 30.0 40.0 50.0

Time (sec)

150.0-

10500.0

50.0-

6

0.0 10.0 20.0 30.0 40.0 50.0 Time (see)

0.0 10.0 20.0 30.0 40.0 50.0 Time (sec)

Figure 9 Plant Response to Inadvertent HPCI !L8 (EOC ICF (HBB))

Page 29



10

0.0 10.0 20.0 30.0

Time (see)0.0 10.0 20.0 30.0

Tum (see)

150.01

100.0

50.0

E3 Lev(inch-REF-SEP-SKRI" -*- Vessel Steam Flow

* Turbine Stearn Flow Feedwater Flow

U

U.U i I I - - A -'-

0.0 10.0 20.0

T-ue (see)30.0 0.0 10.0 20.0

Time (sec)30.0

Figure 10 Plant Response to FW Controller Failure (MOC MELLLA (HBB))

Page 30

VERMONT YANKEE P•l•nA )A

0000-0035-6443-SRLR Rpviv~rnt A

100.0

.02W.0

--- Dome Press Rise (psi) --- Safe VMale F'kN

-Reid Wve FAm --- Bypass \vae Row

100.0 t

U.U

0.0 3.0 Time (sec)

6.0 0.0 3.0

Tume (sec)6.0

VJ

0

0.0 30 6.0 0.0 3.0 Time (sec) Time (sec)

6.0

Figure I I Plant Response to Load Rejection w/o Bypass (MOC MELLLA (HBB))

Page 31



150.0

I100.0

-6- Nmutcg F1lux 10 x~ Am Surface Weda RLDx

-& Core Irilet Flowv 300.0 i

200~.0

-6- Doe Press Rise (psi) S Vaele Flow ---Relief Vale Flo

-%- Bypass Vale Flow

(- - - 75-

100.0

0.0 3.0 Tune (sec)

6.0 0.0 3.0

Tine (sec)6.0

E

0.0 3.0 Tmne (sec)

6.0 0.0 3.0 Time (sec)

6.0

Figure 12 Plant Response to Turbine Trip w/o Bypass (MOC MELLLA (HBB))

Page 32



100.0

50.0

0.0 0

150.0

,01000

50.0

4J

.0 10.0 20.0

Time (sec)0.0 10.0 20.0

T-me (sec)

C C

S C

L)

I

0.0 10.0 20.0 0.0 10.0 20.0 Time (sec) Tune (sec)

Figure 13 Plant Response to Inadvertent HPCI 1L8 (MOC MELLLA (HBB))

Page 33



200.0

150.0

10D.0

50.0

0.0

0.0 10.0 20.0 30.0

Tnre (sec)0.0 10.0 20.0 30.0

Time (sec)

150.0

100.0

50.0

r 0

0

0.00.0 10.0 20.0

Ti'm (sec)30.0 0.0 10.0 20.0

Tine (sec)30.0

Figure 14 Plant Response to FW Controller Failure (EOC MELLLA (HBB))

Page 34



100.0

0.02-

6.0 0.00.0 3.0 Tinm (sec)

3.0 T-m~ (Sec)

6.0

I-i

G,

0.0 3.0 6.0 0.0 3.0 6.0 Time (sec) Time (see)

Figure 15 Plant Response to Load Rejection w/o Bypass (EOC MELLLA (HBB))

Page 35



, 2W0.0

100.0

0.01 6.0 0.1

-6-- Dcme Press Rise (psi) -- SafetyVa Flow -•-Relief VaKoe Raw -.- Bypass Vave Flow

0.0 3.0

Tine (sec)

0 x.3.0

Time (Sec)6.0

ov

o=

U 0J

2J

0.0 3.0 6.0 0.0 3.0

TI'Me (sec) Time (sec)6.0

Figure 16 Plant Response to Turbine Trip w/o Bypass (EOC MELLLA (HBB))

Page 36

VERMONT YANKEE D I ~A IA

0000-0035-6443-SRLR IDn- ; ;-l A

ý w" I ý

Sb

0.0 10.0 20.0

Tinr (sec)0.0 10.0 20.0

Tune (sec)

C 0

U Sb

0.0 10.0 20.0 0.0 10.0 20.0

Time (sec) Ture (see)

Figure 17 Plant Response to Inadvertent HPCI /L8

(EOC MELLLA (HBB))

Page 37



4J

0.0 10.0 20.0 30.0

-Tme (sec)0.0 10.0 20.0 30.0

Time (sec)

150.0

100.0

U

0.0 10.0 20.0 Thmi (sec)

30.0 0.0 10.0 20.0 T-ue (see)

30.0

Figure 18 Plant Response to FW Controller Failure (EOC ICF (UB))

Page 38



200.0

150.01

4100.0

-R NautronnFux/I10 x*~ Aw.e Surface t-oat Flux

--- Core kInet Raw~~

11

50.0 t-

0.0 3.0

Tfnf (sec)6.0 0.0 3.0

T'mx (sec)6.0

E 0

0.0 3.0 6.0 0o 3.0 Tine (sec) Time (sec)

6.0

Figure 19 Plant Response to Load Rejection w/o Bypass (EOC ICF (UB))

Page 39



300.0

200.0

- Dome Press Rise (psi) - - Safety VaKe Floa

*---Refief Vve Row -.- Bypass Va\ke FRc

f-

100.01

0.0 3.0

Tine (sc)6.0 0.0 3.0

Time (sec)6.0

200.0

100.0O

0.0

-1000.

0.0

V

3.0 6.0 0.0 3.0 Tune (Sec) Tine (see)

6.0

Figure 20 Plant Response to Turbine Trip w/o Bypass (EOC ICF (UB))

Page 40



100ID..i

25.0

-- -- - - -25-0 40.0 50.0 0.0 10.0 20.0 30.0

-rne (Sec)0.0 10.0 20.0 30.0

Tinr (sec)40.0 50.0

150.0

10 100.0

a I a 0. 2 0 U

S U

0.0 10.0 20.0 30.0 40.0 50.0 0.0 10.0 20.0 30.0

Tinm (sec) Time (Sec)40.0 50.0

Figure 21 Plant Response to Inadvertent HPCI /L8 (EOC ICF (UB))

Page 41

VERMONT YANKEE Relmnd 24


200.0

aB- NeUtr Flux / 10 --x- Ave SurfacHe -at Flux -A Core InA Flow

-- Core Inrd Subcooing150.0.

3 S S R E3 E3 .3

U

50.0t

Oa n "

0.0 10.0 20.0 T-ru (Sec)

30.0 0.0 10.0 20.0 30.0 Time (sec)

I

0

U

U

0.0 10.0 20.0 Tine (sec)

0.0 10.0 20.0 Time (sec)

Figure 22 Plant Response to FW Controller Failure (EOC MELLLA (UB))

Page 42

VERMONT YANKEE !0.1I.A 9A

0000-0035-6443-SRLR

3W0.0

200.0

100.0

-s- Dane Press Rise (psi) SafetyVae Flo

-Relief Wh~eFlow, -- Bypass \kd~e Row

----77Z ý-.0 0.0 6.0 0

00.0 3.0

Tin= (sec)3.0

Time (Sec)6.0

100.0

a UC C

C I0.0 3.0 6.0 0.0 3.0 6.0

TMne (see) Time (sec)

Figure 23 Plant Response to Load Rejection w/o Bypass

(EOC MELLLA (UB))

Page 43


0000-0035-6443-SRLR Revision. 0

300.0

200.0

100.0

-- 0.oe 6.0 0.00.0 3.0

Time (Sec)&o

Ti'm (see)6.0

200.0

,,0 100.0

0.0

-100.0

Q0

0.0 3.0 6.0 0.0 3.0 Time (sec) Time (sec)

60

Figure 24 Plant Response to Turbine Trip w/o Bypass (EOC MELLLA (UB))

Page 44



C2. 100.0

50.0

0.0

150.0

"0100.0

50.0

.5 U

0.0 10.0 20.0

Tine (see)0.0 10.0 20.0

Thef (Se)

1.0 . -- Scran Reacd - Total ReactiVI

o, 0.0

-1.0

0-010.0

T¶==m (sec)0.0 10.0

Tine (see)20.0

Figure 25 Plant Response to Inadvertent HPCI /L8 (EOC MELLLA (UB))

Page 45



0.0 4.0 8.0 0.0 4.0 8.0 Tmne (sec) Time (sec)

100.0

'4

5 C

U

0.0 4.0 8.0 0.0 4.0

Time (sec) Time (see)8.0

Figure 26 Plant Response to MSIV Closure (Flux Scram) - ICF (HBB)

Page 46



0.0 4.0 8.0 0.0 4.0

Time (see) Time (sec)8.0

200.0

100.0

0.0

T

-100.0

0.0 4.0 8.0 0.0 4.0

Tmue (sec) Tine (sec)8.0

Figure 27 Plant Response to MSIV Closure (Flux Scram) - MELLLA (HBB)

Page 47



Appendix A

Analysis Conditions

The reactor operating conditions and the pressure relief and safety valve configuration used in the reload licensing analysis for this plant and cycle are presented in Tables A-I and A-2 below.

Table A-1 Reactor Operating Conditions

Analysis ValueParameter ICF LCF

NFWT NFWT

Thermal power, MWt 1912.0 1912.0

Core flow, Mlb/hr 51.4 47.5

Reactor pressure (core mid-plane), psia 1041.4 1039.8

Inlet enthalpy, Btu/lb 520.9 518.6

Non-fuel power fraction 0.036 0.036

Steam flow, Mlb/hr 7.92 7.91

Dome pressure, psig 1010.0 1010.0

Turbine pressure, psig 946.4 946.5

Table A-2 Pressure Relief and Safety Valve Configuration

Number of Lowest Setpoint Valve Type Valves (psig)

Safety/Relief Valve 4 1113.0

Spring Safety Valve 3 1277.0

Page 48

VERMONT YANKEE 0000-0035-6443-SRLR Reload 24 Revision .0

Appendix B

Decrease in Core Coolant Temperature Events

The Loss of Feedwater Heating (LFWH) event and Inadvertent High Pressure Coolant Injection (HPCI) Startup event are the only cold water injection AOOs checked on a cycle-by-cycle basis.

The Loss-of-Feedwater Heating event was analyzed at the EPU power level (1912 MWt) using the BWR Simulator Code. The use of this code is permitted in GESTAR II. The transient plots, neutron flux and heat flux values normally reported in Section 9 are not an output of the BWR Simulator Code; therefore, those items are not included in this document. The OLMCPR result is shown in Section 11.

An ODYN analysis has been performed for the inadvertent High Pressure Coolant Injection (HPCI) Startup event as part of the reload licensing analysis. Therefore, inadvertent HPCI calculations assuming a L8 trip were performed. The OLMCPR results are shown in Section 1I.

Page 49


Appendix C ARTS Power and Flow Operating Limits Adjustments

The Vermont Yankee plant uses the PANAC II methodology, so it does not have thermal-mechanical MAPLHGR limits. This plant requires the use of LHGRFAC multipliers (power and flow dependent multipliers on LHGR) to ensure that off-rated AOO thermal-mechanical criteria are met. The LHGRFAC multipliers also provide adequate protection for the off-rated LOCA conditions since a constant local peaking factor is used in the LOCA evaluation. Off-rated MAPLHGR multipliers are not required.

The ARTS power and flow dependent operating limits for all operating flexibility options are provided in References C-1, C-2 and C-3. Due to Cycle 25 having no plant specific changes, including no safety limit change from that assumed in Cycle 24, no adjustments were made to the MCPR(p) and MCPR(f) limits curves. Figures C-1 through C4 provide the power and flow dependent limits for VY Cycle 25 at EPU conditions.

References

C-1 Vermont Yankee Nuclear Power Station APRM/RBMiTechnical Specifications / Maximum Extended Load line Limit Analysis (ARTS/MELLLA), NEDC-33089P, March 2003.

C-2 Entergy Nuclear Operations Incorporated Vermont Yankee Nuclear Power Station Extended Power Uprate, GE-NE-0000-001 1-7129-01, Rev. 0, Project Task Report, Task T0900, July 2003.

C-3 Statistically Based Rod Withdrawal Error Analysis for Vermont Yankee Nuclear Power Station, GE-NE-0000-00l16-345 1-RO, July 2003.

Page 50

VERMONT YANKEE ID.I, -4'A

0000-0035-6443-SRLR SRevision 0

2.80

2.60

2.40

2.20

2.00

E. 1.80 a

' 1.60

1.40

1.20

1.00

> 60% Flow

< 60% Flow

Operating Limit MCPR (P) = K(P) x Operating Limit MCPR (100)

For P < 23%: No Thermal Limits Required

For 23% < P< 25%, >60% Flow: OLMCPR(P) = 2.48 - 0.034 x (P - 25%)

For 23%

VERMONT YANKEE Vop*1r~A )

0000-0035-6443-SRLR

Revision 0

1.10

1.00

0.90

0.80

0.70

0.60

5 0.50

00,,

60% 1 Flow I

4-

_N I > 60% Flow

LHGR(P) = LHGRFAC(P) x LHGRstd Where LHGRstd = Rated LHGR limits

For P< 23%: No Thermal Limits Required

For 23% < P 60% Flow: LHGRFAC(P) = 0.586 + 0.0084 x (P - 25%)

For 23%



1.10

1.00

0.90

, 0.80

U.

w,

0.70

U.

0.60

0.50

0.40

0.30

Max Flow= 109.5%

LHGR(F) = LHGRFAC(F) X LHGRstd LHGRstd = STANDARD LHGR LIMITS

For W (% Rated Core Flow) > 30% And Max Runout Flow < 109.5%

LHGRFAC(F) = The Minkium of EITHER 1.0 OR { 0.8737 x (Wi 100) + 0.2779)

W = % Rated Core Flow

20 30 40 50 60 70 80

Core Flow (% Rated)90 100 110 120

Figure C-3 LHGR Flow Factor LHGRFAC(F) for EPU

Page 53

VERMONT YANKEE P10.1A !)A

0000-0035-6443-SRLR

Revision 0

1.8

1.7

1.6

E

S 1.5

Iz 1.4

1.3

1.2

1.1

f - For W(C) (% Rated Core Flow) > 30% MCPR(F) = MAX(1.20, A(F) * W(C) 1100 + B(F))

Max Runout Flow < 109.5%

A(F) = - 0.602 B(F) = 1.747

-~ p -p - I - -.-

Maximum Flow Rate 109.5% 4

20 30 40 50 60 70 80

Core Flow (% Rated)

90 100 110 120

Figure C-4 Flow Dependent MCPR Operating Limit MCPR(F) for EPU

Page 54



Appendix D

Implementation of Extended Power Uprate (EPU)

To provide the Vermont Yankee Nuclear Power Station (VYNPS) with operating improvements, analyses were performed to increase the rated power from 1593 MWt to 1912 MWt. Reference D-1 provides the basis for operation of VYNPS at Extended Power Uprate (EPU), i. e., 1912 MWt conditions. The required OLMCPRs are provided in Section 11.

References

D- I Safety Analysis Report for Vermont Yankee Nuclear Power Station Constant Pressure Power Uprate, NEDC-33090P, Rev. 0, September 2003.

Page 55



Appendix E

Stability Solution Option 1-D Exclusion and Buffer Regions at EPU

Condition

The stability Option I-D Exclusion Region and Buffer Region were calculated for Vermont Yankee Cycle 25 operation at the EPU (Extended Power Uprate) conditions. The ehdpoints of the regions are defined in Table E-1. The region boundaries are defined using the Generic Shape Function (GSF), Equation E-1. The regions are shown on the Vermont Yankee Cycle 25 power/flow map in Figure E-I. These regions are valid up to an end of cycle exposure of 13,930 MWd/ST.

Table E-1: Exclusion and Buffer Region Endpoints'

Exclusion Region Power Flow (% rated) 2 (% rated) 3

A 66.26 52.25

B 35.69 31.27

Buffer Region Power Flow (% rated)

2 (% rated) 3

A 70.10 57.25

B 30.69 31.17

I. Point "A" is on the High Flow Control Line (HFCL) and point "B" is on the Natural Circulation Line (NCL). For Vermont Yankee Cycle 25, the HFCL is the Maximum Extended Load Line Limit Analysis (MELLLA) boundary.

2. Rated core power is 1912 MWt.

3. Rated core flow is 48 Mlb/hr.

Equation E-1: Generic Shape Function

I[ w-wo +( w-w

PB2

P = ,PB (L II W "-.

where, P = a core thermal power value on the region boundary (% of rated), W = the core flow rate corresponding to power, P, on the region boundary (% of rated),

Page 56



PA = core thermal power at point A (% of rated), PB = core thermal power at point B (% of rated), WA = core flow rate at point A (% of rated), and WB = core flow rate at point B (% of rated).

Core Flow (Mlb/hr) 0 5 10 13 20 25 30 35 40 45 50 55

120

1100

90o

so

•70 I

~60~ me

•0

~40

30

20

0o

k A N~am61 oo 8 : M ri-iPwnp Sp~d C. 153.3% Pow.. 360M 01 D158% P-1 1111O1 0 4113% P-11i 75.M E:B 100.0% Po-1t 1000 V. 03.3% Pow/ 100.0 r. '.000% Po1 , WA P: 813% Pow-ol070,

.0 le11 7% P-d107.01, K I 18.7% P-1/ lOOAl

1:157% Pool 327A -NOt. (11):V Wdo wto prot~on kw. Ovs Imis~

* Exaujslor

* Buffer Reg

*Minimum Pump Spi

* Natural Cimulatio

4 Flow 4 Flow 4Flow %Flow

4 Flow 4 Fl~ow

RFow 4 Flow (1) 4 , low~l) tFlow (i)

t- pump c0ot.on ft fwidu 90er IHo.

MELLLA Upper Boundary Line

n Region

iion(1

D E .91

S E' ___ 3 v

83.3% EPURodLne o 100% OLTP Rod Line)

renased Core Flow Region

Minimum Power Line (1)

100% E.Pu 1912 IVA 100A Co. * 10 VF

22D0

2000

law0

1600

1400]

1200~

600

400

A 8

J I I

so 90 100 110 1200 10 20 30 40 50 60 70

Core Flow (%)

Figure E-1: Exclusion and Buffer Regions on the EPU Power/Flow Map

Page 57



Appendix F

List of Acronyms

Acronym Description

ACPR Delta Critical Power Ratio

Ak Delta k-effective 2RPT Two Recirculation Pump Trip ADS Automatic Depressurization System ADSOOS Automatic Depressurization System Out of Service AOO Anticipated Operational Occurrence APRM Average Power Range Monitor ARTS APRM, Rod Block and Technical Specification Improvement Program BOC Beginning of Cycle BSP Backup Stability Protection Btu British thermal unit BWROG Boiling Water Reactor Owners Group COLR Core Operating Limits Report CPR Critical Power Ratio DIVOM Delta CPR over Initial MCPR vs. Oscillation Magnitude DR Decay Ratio DS/RV Dual Mode Safety/Relief Valve ECCS Emergency Core Cooling System EEOC Extended End of Cycle ELLLA Extended Load Line Limit Analysis EOC End of Cycle EOR End of Rated (All Rods Out I 00%Power / 100%Flow / NFWT) ER Exclusion Region FFWTR Final Feedwater Temperature Reduction FMCPR Final MCPR FOM Figure of Merit FWCF Feedwater Controller Failure FWTR Feedwater Temperature Reduction GDC General Design Criterion GESTAR General Electric Standard Application for Reactor Fuel GETAB General Electric Thermal Analysis Basis GSF General Shape Function HAL Haling Bum HBB Hard Bottom Bum HBOM Hot Bundle Oscillation Magnitude HCOM Hot Channel Oscillation Magnitude HFCL High Flow Control Line HPCI High Pressure Coolant Injection

Page 58



Acronym Description

ICA Interim Corrective Action ICF Increased Core Flow IMCPR Initial MCPR IVM Initial Validation Matrix L8 Turbine Trip on high water level (Level 8) LCF Low Core Flow LHGR Linear Heat Generation Rate LOCA Loss of Coolant Accident LPRM Local Power Range Monitor LRHBP Load Rejection with Half Bypass LRNBP Load Rejection without Bypass LTR Licensing Topical Report MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio MELLLA Maximum Extended Load Line Limit Analysis MELLLA+ MELLLA Plus MOC Middle of Cycle MRB Maximal Region Boundaries MSIV Main Steam Isolation Valve MSIVOOS Main Steam Isolation Valve Out of Service MTU Metric Ton Uranium MWd Megawatt day MWd/ST Megawatt days per Standard Ton MWd/MT Megawatt days per Metric Ton MWt Megawatt Thermal NBP No Bypass NCL Natural Circulation Line NFWT Normal Feedwater Temperature NOM Nominal Burn NTR Normal Trip Reference OLMCPR Operating Limit MCPR OOS Out of Service OPRM Oscillation Power Range Monitor Pdome Peak Dome Pressure PsI Peak Steam Line Pressure Pv Peak Vessel Pressure PCT Peak Clad Temperature PHE Peak Hot Excess PLHGR Peak Linear Heat Generation Rate PLUOOS Power Load Unbalance Out of Service PRFDS Pressure Regulator Failure Downscale PROOS Pressure Regulator Out of Service Q/A Heat Flux RBM Rod Block Monitor RC Reference Cycle RCF Rated Core Flow

Page 59



Acronym Description RFWT Reduced Feedwater Temperature RPS Reactor Protection System RPT Recirculation Pump Trip RPTOOS Recirculation Pump Trip Out of Service RV Relief Valve RVM Reload Validation Matrix RWE Rod Withdrawal Error SC Standard Cycle SL Safety Limit SLMCPR Safety Limit Minimum Critical Power Ratio SLO Single Loop Operation SRLR Supplemental Reload Licensing Report S/RV Safety/Relief Valve SRVOOS Safety/Relief Valve(s) Out of Service SS Steady State SSV Spring Safety Valve STU Short Tons (or Standard Tons) of Uranium TBV Turbine Bypass Valve TBVOOS Turbine Bypass Valves Out of Service TCV Turbine Control Valve TCVOOS Turbine Control Valve Out of Service TCVSC Turbine Control Valve Slow Closure TLO Two Loop Operation TRF Trip Reference Function TSIP Technical Specifications Improvement Program TTHBP Turbine Trip with Half Bypass TTNBP Turbine Trip without Bypass UB Under Burn

Page 60

P 9. GNF. - NRC: Home Page · 2012. 11. 21. · Inadvertent HPCI/L8 306 121 0.25 13 Operating...

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Transcript of P 9. GNF. - NRC: Home Page · 2012. 11. 21. · Inadvertent HPCI/L8 306 121 0.25 13 Operating...