RBG-46943, Supplement to License Amendment Request (LAR) Changes to Technical Specification 5.6.5, Core Operating Limits Report (Colr).

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Supplement to License Amendment Request (LAR) Changes to Technical Specification 5.6.5, Core Operating Limits Report (Colr).
ML092300229
Person / Time
Site: River Bend Entergy icon.png
Issue date: 08/12/2009
From: Roberts J
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
RBG-46943
Download: ML092300229 (20)
v • d • e


Text

Entergy Operations, Inc.

River Bend Station 5485 U.S. Highway 61 N St. Francisville, LA 70775 SEntergy Tel 225-381-4149 Jerry C. Roberts Director, Nuclear Safety Assurance RBG-46943 August 12, 2009 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Supplement to License Amendment Request (LAR) Changes to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)"

River Bend Station, Unit 1 Docket No. 50-458 License No. NPF-47 REFERENCE 1. Letter RBG-46863 from Entergy to USNRC, License Amendment Request (LAR) Changes to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)" dated November 20, 2008

2. Letter GNRO-2008/00053 from Entergy to USNRC, "Supplement to Amendment Request Changes to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)" dated July 21, 2008.

Dear Sir or Madam:

On November 20, 2008, Entergy Operations, Inc. (Entergy) requested a license amendment for River Bend Station, Unit 1 (RBS), Reference 1. The change was to Technical Specification (TS) 5.6.5, "Core Operating Limits Report (COLR)",to add a reference to an analytical method that will be used to determine core operating limits. In that submittal a commitment was made to submit information regarding the impact of the fuel transition on the plant's safety analyses in a supplemental letter. is the additional information on the impact of the fuel transition on the plant's safety analyses. This information is similar to the corresponding request for Grand Gulf Nuclear Station (GGNS), Reference 2. This information completes the commitment made in the initial submittal. This response does not include new commitments.

AOoi

RBG-46943 Page 2 of 2 If you' have any questions or require additional information, please contact David Lorfing at (225)-381-4157.

Sincerely, r ect Assessment River Bend Station - Unit 1 JCR/DNL/bmb Attachments:

1. Additional Information On The Impact Of The Fuel Transition On The Plant's Safety Analyses cc: Regional Administrator U. S. Nuclear Regulatory Commission Region IV 612 E. Lamar Blvd., Suite 400 Arlington, TX 76011-4125 NRC Senior Resident Inspector' P. O. Box 1050 St. Francisville, LA 70775 U. S. Nuclear Regulatory Commission Attn: Mr. Alan B. Wang MS 0-7 D1l Washington, DC 20555-0001 Mr. Jeffrey P. Meyers Louisiana Department of Environmental Quality Office of Environmental Compliance Attn. OEC - ERSD P. O. Box 4312 Baton Rouge, LA 70821-4312

i Attachment 1 RBG-46943 Additional Information On The Impact Of The Fuel Transition On The Plant's Safety Analyses

Attachment to RBG-46943 Page 1 of 17 Commitment Entergy will provide information regarding the impact of the fuel transition on the plant's safety analyses in a supplemental letter. Note the scope of this commitment was based upon a previous question for Grand Gulf Nuclear Station (GGNS) in Reference 2.

Previous NRC Question:

The application describes a transition from the current core with a full loading of ATRIUM-10 to a full loading of GE14 fuel. This transition will start with the upcoming refueling outage and continue over several refueling outages. Once the NRC staff approves the inclusion of the GEXL97 correlation into the TS, the licensee calculates the core operating limits listed in TS.

5.6.5a and performs the plant accident analyses without further review by the NRC staff.

Therefore, the staff requests that the licensee provide additional information to allow the staff to review the impact of the transition on the safety analyses.

Specifically, for loss-of-coolant accident (LOCA), anticipated transient without scam (ATWS),

abnormal operation occurrence (AOO), American Society of Mechanical Engineers (ASME)

Code overpressure, and stability analyses performed for the initial transition core:

a) state the approved methodology and/or the computer codes used and the reference (e.g. topical report) documenting the methodology/computer codes, b) state if the analysis is in compliance with all applicable restrictions in the staff safety evaluation(s) approving the methodology, and c) provide the quantitative results of the figure(s) of merit (FOM) compared against the acceptance criteria.

Below is an example of a format that would provide the set of information to the staff.

Analysis Methodology / Staff Comply with all FOM Result Valve Code(s) Approval applicable vs.

Used restrictions and Acceptance conditions Criterion ATWS Ref. xx Y/N Peak pressure PCT Peak pool temp Peak containment pres LOCA PCT Local MWR Core wide MWR ASME Peak pressure Overpres sure Stability Stability re gions 1 AOO OLMCPR2

Attachment to RBG-46943 Page 2 of 17 For the LOCA analysis, also provide a narrative showing that:

a) the limiting break location, break size, and single failure(s) were identified and used.

b) the limiting power/flow conditions and axial power shape were used.

c) the legacy fuel analysis (MAPLHGR) will continue to be applicable through the transition cycles.

Note 1: Stability regions = A figure showing the stability regions Note 2: OLMCPR = operating limit minimum critical power ratio

Response

The response for each of the subject areas doesn't conveniently fit into a single table. The response is organized with the methods subjects first, followed by the River Bend results. The major subjects are organized in the same order as the table: ATWS, LOCA, ASME Overpressure, Stability, and AOOs (Including Off-Rated Limits). With respect to the requested COLR figures, the information supporting the COLR is included here. The requested figures will be provided with the normal COLR submittal.

ATWS Methodology Comply with all Methodelog Stapplicable Code(s)

U Staff Approval restrictions and Used conditions ISCOR09 Note 1 Yes PANAC1 1 NEDE-30130-P-A, Steady-State Nuclear Methods,- Yes April 1985. Note 2 ODYN09 NEDC-24154P-A, Qualification of the One- Yes Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4),

February 2000.

STEMP04 Note 3 N/A TASC03 NEDC-32084P-A, TASC-03A, A Computer Program Yes for Transient Analysis of a Single Channel, Revision 2, July 2002.

(1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods (2) The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE),

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods", (TAC NO. MA6481), November 10, 1999.

Attachment to RBG-46943 Page 3 of 17 (3) The STEMP code uses fundamental mass and energy conservation laws to calculate the suppression pool heatup. The use of STEMP was noted in NEDE-24222, "Assessment of BWR Mitigation of ATWS, Volume I & II (NUREG-0460 Alternate No. 3)

December 1, 1979." The code has been used in ATWS applications since that time.

There is no formal NRC review and approval of STEMP.

Parameter Result Acceptance Criteria Peak Vessel Pressure (psig) 1497 <1 500 Peak Suppression Pool 182.4 <185.0 Temperature (OF)

Peak Containment Pressure (psig) 9.2 *<15.0 Peak Cladding Temperature (OF) 1405 *2200 Peak Local Cladding Oxidation (%) insignificant(1' *<17 (1) The calculated peak cladding temperature is less than 1600OF and therefore cladding oxidation is insignificant compared to the acceptance criteria and is not explicitly calculated.

Attachment to RBG-46943 Page 4 of 17 LOCA Methodology Comply with all Staff Approval applicable MCode(s)

Uede Srestrictions and Used conditions ISCOR09 Note 1 Yes LAMB08 NEDE-20566P-A, General Electric Analytical Model for Yes Loss-of-Coolant Analysis in Accordance with 10CFR50 Appendix K, September 1986.

SAFER04/ NEDE-23785-1 -PA Rev. 1, The GESTR-LOCA and Yes GESTR08 SAFER Models for the Evaluation of the Loss-of- Note 4 Coolant Accident, Vol. 1, GESTR-LOCA - A Model for the Prediction of Fuel Rod Thermal Performance, October 1984.

NEDE-23785-1-PA Rev. 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 2, SAFER - Long Term Inventory Model for BWR Loss-of-Coolant Analysis, October 1984.

NEDE-23785-1-PA Rev. 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3, SAFER/GESTR Application Methodology, October 1984.

NEDE-23785P-A Rev. 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3 Supplement 1, Additional Information for Upper Bound PCT Calculation, 'March 2002.

Notes 2 and 3 TASC03 NEDC-32084P-A, TASC-03A, A Computer Program for Yes Transient Analysis of a Single Channel, Revision 2, July 2002.

(1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R. Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods.

(2) Letter, J.F. Klapproth (GE) to USNRC, "Transmittal of GE Proprietary Report NEDC-32950P

'Compilation of Improvements to GENE's SAFER ECCS-LOCA Evaluation Model," dated January 2000 by letter dated January 27, 2000.

(3) Letter, S.A. Richards (NRC) to J.F. Klapproth (GE), "General Electric Nuclear Energy Topical Reports NEDC-32950P and NEDC-32084P Acceptability Review," May 24, 2000.

Attachment to RBG-46943 Page 5 of 17 (4) The ECCS-LOCA evaluations are in compliance with the one SER restriction upon the SAFER/GESTR-LOCA methodology. This SER restriction is described within NEDE-23785-1-PA Rev. 1, The GESTR-LOCA and SAFER Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 3, SAFER/GESTR Application Methodology. October 1984, as

"... NRC will require a demonstration that the PCT value calculated by the licensing method always exceeds the upper-bound value". All ECCS-LOCA evaluations explicitly calculate both the licensing PCT and the upper bound PCT, and all evaluations report a licensing PCT which exceeds the upper bound PCT.

Parameter ATRIUM-10 Result GE14 Result Acceptance Criteria Licensing Basis < 1810 OF < 1660 OF < 2200 OF PCT Maximum Local < 3% < 1% < 17%

Oxidation Core Wide Metal- < 0.1%. < 0.1% < 1.0%

Water Reaction Coolable Items 1 AND 2 Items 1 AND 2 Satisfied by:

Geometry PCT < 2200 OF AND Maximum Local Oxidation < 17%

Core Long Term Satisfied by: Satisfied by: Core temperature Cooling EITHER EITHER acceptably low Core reflooded above Core reflooded above AND TAF TAF Long-term decay heat OR OR removed Core reflooded to the Core reflooded to the elevation of jet pump elevation of jet pump suction and 1 core suction and 1 core spray system in spray system in operation operation LOCA Specific NRC Question (a)

For the LOCA analysis, also provide a narrative showing that the limiting break location, break size, and single failure(s) were identified and used.

Response

The ECCS-LOCA evaluations evaluate all potentially limiting single failures including:

Attachment to RBG-46943 Page 6 of 17 Single ADS valve out-of-service The ECCS-LOCA evaluations identify the HPCS emergency diesel generator as the limiting single failure.

The limiting break location is not a function of fuel type, and the ECCS-LOCA evaluations utilize prior SAFER/GESTR-LOCA evaluations to identify the limiting break location as the recirculation suction line break.

The ECCS-LOCA evaluations examine break sizes ranging from a double-ended guillotine break of the recirculation suction line to small breaks for which no core heatup is predicted. The break sizes are selected to satisfy the Appendix K requirement for the use of the Moody Slip Flow Model with a discharge coefficient of 1.0, 0.8, and 0.6. Additional break sizes are evaluated to fully characterize the PCT versus break size response and to ensure that the limiting break size is identified.

The GE14 limiting large break is identified as the double-ended guillotine break of the recirculation suction line and the limiting small break is identified as a 0.05 ft 2 break of the recirculation suction line. The 0.05 ft2 break is identified as the overall most limiting break.

The ATRIUM-10 limiting large break is identified as the double-ended guillotine break of the recirculation suction line and the limiting small break is identified as a 0.05 ft 2 break of the recirculation suction line. The large break is identified as the overall most limiting break.

LOCA Specific NRC Question (b)

For the LOCA analysis, also provide a narrative showing that the limiting power/flow conditions and axial power shape were used.

Response

The potentially limiting power and flow conditions are evaluated, including the rated power/flow point and the point on the MELLLA rod line with highest power and lowest flow. Calculations are performed at these power/flow points using both nominal and Appendix K conditions. The rated power/flow point is identified as the limiting power and flow condition for GE14 and the MELLLA power/flow point is identified as the limiting power and flow condition for ATRIUM-1 0.

Single loop operation is investigated and shown to be bounded by two loop operation. Likewise, feedwater temperature reduction is also examined and shown to be bounded by LOCA at rated feedwater temperature. Increased core flow is also dispositioned as bounded by LOCA at rated core flow.

Both mid-peaked and top-peaked axial power shapes are evaluated by the ECCS-LOCA evaluations. The limiting axial power shape is identified as mid-peaked for large breaks and the limiting axial power shape for small breaks is identified as top-peaked.

LOCA Specific NRC Question (c)

For the LOCA analysis, also provide a narrative showing that the legacy fuel analysis (MAPLHGR) will continue to be applicable through the transition cycles.

Attachment to RBG-46943 Page 7 of 17 Response C.

The legacy fuel MAPLHGRs are not applied to the transition cycles because a new SAFER/GESTR-LOCA evaluation is performed for the legacy fuel and new MAPLHGRs are generated. This legacy fuel SAFER/GESTR-LOCA evaluation satisfies all 10 CFR 50.46 licensing requirements, the one SER restriction, and generates MAPLHGR limits which are applicable to the transition cycles.

ASME Overpressure Methodology Comply with all MCode(s) Staff Approval applicable Ueds Srestrictions and Used conditions PANAC1 1 NEDE-30130-P-A, Steady-State Nuclear Methods, Yes April 1985, Note 1 ODYN09 NEDC-24154P-A, Qualification of the One- Yes Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4),

Revision 1, February 2000.

(1) The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE),

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods", (TAC NO. MA6481), November 10,

-1999.

Steam TS Dome ASME Line Pressure Vessel Event Pressure Pressure Pressure Limit Limit (2)

(psig) (psig)

(psig) ( (psig) (psig)

MSIV Closure (Flux Scram) - ICF 1306 1313 1348 1325 1375 (HBB)

MSIV Closure (Flux Scram) - MEOD 1301' 1308 1333 1325 1375 (HBB)

(1) Technical Specification 2.1.2 Reactor Coolant System Pressure SL (2) GESTAR II, Revision 16, Section S.3 Vessel Pressure ASME Code Compliance Model

Attachment to RBG-46943 Page 8 of 17 Stability Methodology Comply with all Staff Approval applicable MCode(s)

Ueds Srestrictions and Used conditions ISCOR09 Note 1 Yes PANACI 1 NEDE-30130-P-A, Steady-State Nuclear Methods, Yes April 1985.

ODYSY05 NEDE-33213P-A, ODYSY Application for Stability Yes Licensing Calculations Including Option I-D and II Long Term Solutions, April 2009.

NEDO-32339-A, Reactor Stability Long-Term Solution: Enhanced Option I-A, Revision 1, April 1998.

NEDO-32339-A Supplement 3, Reactor Stability Long-Term Solution: Enhanced Option I-A, Flow Mapping Methodology, Revision 1, April 1998.

NEDO-32339-A Supplement 4, Reactor Stability Long-Term Solution: Enhanced Option I-A, Generic Technical Specifications, Revision 1, April 1998.

NEDC-32339P-A Supplement 1, Reactor Stability Long-Term Solution: Enhanced Option I-A, ODYSY Application to ElA, December 1996.

NEDC-32339P-A Supplement 2, Reactor Stability Long-Term Solution: Enhanced Option I-A, Solution Design, Revision 1, April 1998.

(1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-24011-P Rev. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R.

Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATVVS, Stability, and LOCA applications is consistent with the approved models and methods.

Attachment to RBG-46943 Page 9 of 17 Decay Ratios Resulting from ODYSY Analysis for River Bend Cycle 16 Cycle 16 Case State Core Decay Ratio Channel Decay Ratio RVM1 B 0.350 0.009 RVM2 A' 0.559 0.202 RVM3 B' 0.654 0.069 RVM4 FRE H1 0.517 0.113 RVM5 FRE HO 0.467 0.014 RVM6 A' LOFH 0.789 0.291 RVM7 B' LOFH 0.774 0.319 W"

See F-igure 7 Tor comparison OT the result values to me Acceptance Criteria.

Figure 1. River Bend Station Region Boundaries (Normal Trip Reference)

Core Flow (Mlb/hr) 0 10 20 30 40 50 60 70 80 90 100 120 3600 110 100 I 3300 3000 90 2700 80 2400 70 2100 s-PL 60 1800 50 1500 2 40 1200 30 900 20 600 10 300 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow (%)

Attachment to RBG-46943 Page 10 of 17 Figure 2. River Bend Decay Ratios vs. ODYSY Stability Criteria 1

0.9 -- ODYSY Stability Criteria 0.8

  • RVM1 0.7 A RVM2 0

n

, 0.6 RVM3 00.5 0 RV M4 T

0 0.4

  • R\/MK 0.3

)KRVW 0.2

  • RVM7 0.1 0

0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 1 Channel Decay Ratio

Attachment to RBG-46943 Page 11 of 17 AQOs Including Off-Rated Limits Comply with all Methodology applicable

/ Code(s) Staff Approval restrictions and Used conditions ISCOR09 Note 1 Yes PANAC1 1 NEDE-30130-P-A, Steady-State Nuclear Methods, Yes April 1985, Note 2 ODYN09 NEDC-24154P-A, Qualification of the One- Yes Dimensional Core Transient Model (ODYN) for Boiling Water Reactors (Supplement 1 - Volume 4),

Revision 1, February 2000.

TASC03 NEDC-32084P-A, TASC-03A, A Computer Program Yes for Transient Analysis of a Single Channel, Revision 2, July 2002.

(1) The ISCOR code is not approved by name. However, the SER supporting approval of NEDE-2401 1-P Rev. 0 by the May 12, 1978 letter from D. G. Eisenhut (NRC) to R.

Gridley (GE) finds the models and methods acceptable, and mentions the use of a digital computer code. The referenced digital computer code is ISCOR. The use of ISCOR to provide core thermal-hydraulic information in reactor internal pressure differences, Transient, ATWS, Stability, and LOCA applications is consistent with the approved models and methods (2) The physics code PANACEA provides inputs to the transient code ODYN. The use of PANAC Version 11 in this application was initiated following approval of Amendment 26 of GESTAR II by letter from S.A. Richards (NRC) to G.A. Watford (GE),

Subject:

"Amendment 26 to GE Licensing Topical Report NEDE-2401 1-P-A, GESTAR II Implementing Improved GE Steady-State Methods", (TAC NO. MA6481), November 10, 1999.

Limiting Pressurization Events OLMCPR Summary Table Application Condition / Exposure Option A GE14 Option A ATRIUM-Range* 10 Equipment in Service BOC to MOC 1.27 1.28 MOC to EOC 1.31 1.31 RPTOOS BOC to MOC 1.33 1.34 MOC to EOC 1.36 1.35 PROOS BOC to MOC 1.27 1.28 MOC to EOC 1.31 1.31

Attachment to RBG-46943 Page 12 of 17 TBVOOS BOC to MOC 1.30 1.31 MOC to EOC 1.34 1.34 TBVOOS & RPTOOS BOC to MOC 1.36 1.37 MOC to EOC 1.38 1.39 PROOS & TBVOOS BOC to MOC 1.30 1.31 MOC to EOC 1.34 1.34 PROOS & TBVOOS & RPTOOS BOC to MOC 1.36 1.37 MOC to EOC 1.38 1.39

  • Each application condition covers the entire range of licensed flow and feedwater temperature.

/

Attachment to RBG-46943 Page 13 of 17 MCPRp Limits for GE14 and Atrium-10 Below Ph... Above Phns 23.8% P 23.8% P App. 240%P 40% 10 Cond Exp. Fuel F< F> F< F>P 40%P

, 50%P

, 70%P

, 70%P

, 8510 85%P 85%P 90%P p 50.0% 50.0% 50.0% 50.0%

EOC G1446 2.13 2.44 1.90 2.04 1.63 1.63 1.50 - 1.35 1.31 ATR10 2.08 2.34 1.86 1.97 1.60 1.60 1.48 - 1,35 1.31 MOC G1446 2.11 2.44 1.88 2.04 1.56 1.55 1.44 1.41 - 1.30 1.27 ATR10 2.05 2.34 1.83 1.97 1.56 1.53 1.44 1.40 - 1.31 1.28 EOC G1446 2.13 2.44 1.90 2.04 1.63 1.63 - 1.50 - 1.39 1.36 ATR10 2.08 2.34 1.86 1.97 1.60 1.60 - 1.48 - 1.39 1.35 G1446 2.11 2.44 1.88 2.04 1.56 1.55 1.44 1.42 - 1.36 1.33 MOC ATR10 2.05 2.34 1.83 1.97 1.56 1.53 1.44 1.43 - 1.37 1.34 EOC G1446 2.13 2.44 1.90 2.04 2.04 1.95 1.75 1.66 1.39 1.35 1.31 ATR10 2.08 2.34 1.86 1.97 1.97 1.88 - 1.70 1.61 1.38 1.35 1.31 G1446 2.11 2.44 1.88 2.04 2.04 1.93 - 1.74 1.65 1.33 1.30 1.27 MOO ATR10 2.05 2.34 1.83 1.97 1.97 1.87 - 1.69 1.60 1.33 1.31 1.28 G1446 2.13 2.44 1.90 2.04 1.63 1.63 - 1.52 - - 1.38 1.34 ATR10 2.08 2.34 1.86 1.97 1.60 1.60 - 1.50 - - 1.38 1.34 G1446 2.11 2.44 1.88 2.04 1.56 1.55 - 1.44 - - 1.33 1.30 MOC ATR10 2.05 2.34 1.83 1.97 1.56 1.53 - 1.44 1.35

- - 1.31 G1446 2.13 2.44 1.90 2.04 1.63 1.63 - 1.52 - - 1.42 1.38 EOC _1 0_1 3.

ATR10 2.08 2.34 1.86 1.97 1.60 1.60 1.50 1.43 1.39 G1446 2.11 2.44 1.88 2.04 1.58 1.55 1.46 - - 1.40 1.36 ATR10 2.05 2.34 1.83 1.97 1.59 1.54 1.47 - - 1.41 1.37

Attachment to RBG-46943 Page 14 of 17 Below Pbvoass Above Pbvpass 23.8% P 23.8% P App. , 40%P*F 40%P* 40%P 50%P 70%P 70%P 100%%

Cond Exp. Fuel F< F> F< F> , , , 85%P 85%P 90%P p 500.5.%

50.0% 50.0% 50.0% 50.0%

EOC G1446 2.13 2.44 1.90 2.04 2.04 1.95 - 1.75 1.66 1.42 1.38 1.34 ATR10 2.08 2.34 1.86 1.97 1.97 1.88 - 1.70 1.61 1.41 1.38 1.34 6 G1446 2.11 2.44 1.88 2.04 2.04 1.93 - 1.74 1.65 1.36 1.33 1.30 MOC ATR10 2.05 2.34 1.83 1.97 1.97 1.87 - 1.69 1.60 1.37 1.35 1.31 EOC G1446 2.13 2.44 1.90 2.04 2.04 1.95 - 1.75 1.66 1.45 1.42 1.38 ATR10 2.08 2.34 1.86 1.97 1.97 1.88 - 1.70 1.61 1.45 1.43 1.39 7 G1446 2.11 2.44 1.88 2.04 2.04 1.93 - 1.74 1.65 1.42 .1.40 1.36 MOC ATR10 2.05 2.34 1.83 1.97 1.97 1.87 - 1.69 1.60 1.43 1.41 1.37 Application Conditions:

1. Equipment in Service
2. RPTOOS
3. PROOS
4. TBVOOS
5. TBVOOS & RPTOOS
6. PROOS & TBVOOS
7. PROOS & TBVOOS & RPTOOS
  • While operating in SLO, a 0.02 adder must be applied to all MCPRp limits. In addition, the SLO operating limit MCPR must remain above 1.44 for GE14 and 1.40 for Atrium-1 0.

Attachment to RBG-46943 Page 15 of 17 LHGRFACP Limits for GE14 and Atrium-10 Below PbvPass Above Pbypass App. 23.8% P 23.8% P 40% P 40% P 100%

Cond Exp. Fuel F:5 F> F:5 F> 40% P 50% P- 70% P 85% P 85% P p 50.0% 50.0% 50.0% 50.0%

G1446 0.749 0.647 0.854 0.742 1.000 - - - - 1.000 EO ATRi 0.781 0.628 0.890 0.733 1.000 - - - - 1.000 1,2, 0 4, 5 G1446 0.861 0.651 0.959 0.785 1.000 - - - - 1.000 MO C ATRI 0.810 0.707 0.900 0.777 1.000 - - - - 1.000 0

G1446 0.749 0.647 0.854 0.742 0.742 0.797 0.875 0.924 1.000 1.000 EO ATRI 0.781 0.628 0.890 0.733 0.733 0.789 0.876 0.930 1.000 1.000 3,6, 0 7 G1446 0.861 0.651 0.959 0.785 0.785 0.838 0.914 0.960 1.000 1.000 MO C ATRI 0.810 0.707 0.900 0.777 0.777 0.808 0.879 0.930 1.000 1.000 0

Application Conditions:

1. Equipment in Service
2. RPTOOS
3. PROOS
4. TBVOOS
5. TBVOOS & RPTOOS
6. PROOS & TBVOOS
7. PROOS & TBVOOS & RPTOOS

Attachment to RBG-46943 Page 16 of 17 Loop Manual TBV In-Service MCPRf Limits for GE14-and*Atrium-10 (Application Conditions 1, 2 and 3) 20% F* 30% F* 50% F* 110% F*

1.26 1.26 1.19 1.19 Loop Manual TBVOOS MCPRf Limits for GE14 and Atrium-10 (Application Conditions 4, 5, 6 and 7) 20% F* 30% F* 50% F* 110% F*

1.38 1.38 1.31 1.31 Application Conditions:

1. Equipment in Service
2. RPTOOS
3. PROOS
4. TBVOOS
5. TBVOOS & RPTOOS
6. PROOS & TBVOOS
7. PROOS & TBVOOS & RPTOOS
  • While operating in SLO, a 0.02 adder must be applied to all MCPRf limits. In addition, the SLO operating limit MCPR must remain above 1.44 for GE14 and 1.40 for Atrium-1 0.

Loop Automatic TBV In-Service MCPRf Limits for GE14 and Atrium-'O0 (Application Conditions 1, 2 and 3) 20% F* 30% F* 75%F* 110% F*

1.34 1.34 1.19 1.19 Loop Automatic TBVOOS MCPRf Limits for GE14 and Atrium-10 (Application Conditions 4, 5, 6 and 7) 20% F* 30% F* 75% F* 110% F*

1.46 1.46 1.31 1.31 Application Conditions:

1. Equipment in Service
2. RPTOOS
3. PROOS
4. TBVOOS
5. TBVOOS & RPTOOS
6. PROOS & TBVOOS
7. PROOS & TBVOOS & RPTOOS
  • While operating in SLO, a 0.02 adder must be applied to all MCPRf limits. In addition, the SLO operating limit MCPR must remain above 1.44 for GE14 and 1.40 for Atrium-1 0.

Attachment to RBG-46943 Page 17 of 17 Loop Manual LHGRFACf Limits for GE14 and Atrium-10 (All Application Conditions) 20% F 47.3% F 76.6% F 110%

1.000 F 0.700 1.000 0.700 Loop Automatic LHGRFACf Limits for GE14 and Atrium-10 (All Application Conditions) 20%F 39.2% F 83.1% F 110%F

- 0.550 0.550 1.000 1.000 Application Conditions:

1. Equipment in Service
2. RPTOOS
3. PROOS
4. TBVOOS
5. TBVOOS & RPTOOS
6. PROOS & TBVOOS
7. PROOS & TBVOOS & RPTOOS
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