GNRO-2008/00061, Supplement 2 to Amendment Request, Changes to Technical Specification 5.6.5, Core Operating Limits Report (Colr).

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Supplement 2 to Amendment Request, Changes to Technical Specification 5.6.5, Core Operating Limits Report (Colr).
ML082470361
Person / Time
Site: Grand Gulf Entergy icon.png
Issue date: 08/28/2008
From: Krupa M
Entergy Operations
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
GNRO-2008/00061
Download: ML082470361 (19)
v • d • e


Text

GNRO-2008/00061 August 28, 2008 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

SUBJECT:

Supplement 2 to Amendment Request Changes to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)"

Grand Gulf Nuclear Station, Unit 1 Docket No. 50-416 License No. NPF-29

REFERENCES:

1. Letter GNRO-2007/00071 from Entergy to USNRC, "License Amendment Request (LAR) Changes to Technical Specification 5.6.5, "Core Operating Limits Report (COLR)" dated December 5, 2007 (ADAMS Accession No. ML073440113)
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 (ADAMS Accession No. ML082070087).

Dear Sir or Madam:

By Reference 1 above, Entergy Operations, Inc. (Entergy) requested changes to Grand Gulf Nuclear Station (GGNS) Technical Specification (TS) 5.6.5, "Core Operating Limits Report (COLR)." Specifically, Entergy requested that (1) document NEDC-33383P, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel," Global Nuclear Fuel (GNF), be added to the list of NRC approved analytical methods that are used for determining core operating limits and (2) an exception given in document 24, NEDE 24011-P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-II)" be deleted.

Entergy provided supplemental information and a revision to NEDC-33383P by Reference 2 in response to an NRC staff request for additional information. The response addressed six of the staff's seven questions as the information required to address one of the questions was not yet available. The response to the remaining question is provided as Attachment 1.

The original no significant hazards consideration is not affected by any information contained in the supplemental letter. There are no new commitments contained in this letter.

GNRO-2008/00061 Page 2 If you have any questions or require additional information, please contact Ron Byrd at 601-368-5792.

I declare under penalty of perjury that the foregoing is true and correct. Executed on August 28, 2008.

Sincerely, MAKlRWB/amm

Attachment:

Response to Request for Additional Information cc: Mr. Elmo E. Collins Regional Administrator, Region IV U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 Arlington, TX 76011-4005 U.S. Nuclear Regulatory Commission ATTN: Mr. Jack N. Donohew, Jr.,NRR/ADRO/DORL (w/2)

ATTN: ADDRESSEE ONLY ATTN: U.S. Postal Delivery Address Only Mail Stop OWFN/8 G14 Washington, D.C. 20555-0001 Dr. Ed Thompson, MD, MPH Mississippi Department of Health P. O. Box 1700 Jackson, MS 39215-1700 NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150

Attachment 1 to GNRO-2008/00061 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION to GNRO-2008/00061 Page 1 of 16 RESPONSE TO REQUEST FOR ADDITIONAL INFORMATION REGARDING LICENSE AMENDMENT REQUEST:

CHANGES TO TECHNICAL SPECIFICATIONS 5.6.5, CORE OPERATING LIMITS REPORT (COLR)

This supplementprovides a response to a NRC Question concerning Entergy's request for changes to Grand Gulf Nuclear Station (GGNS) Technical Specifications Section 5.6.5, "Core Operating Limits Report (COLR)" (Reference 1). The changes are needed to allow a fuel transition beginning with operating Cycle 17. Entergy previously responded to six additional questions by Reference 2 of this letter.

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 Scram (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.

to GNRO-2008/00061 Page 2 of 16 Question (continued):

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

Analysis Methodology I Staff Comply with all FOM Result Value Code(s) Used Approval applicable vs.

restrictions Acceptance and conditions Criterion ATWS Ref. xx YIN Peak pressure PCT Peak pool temp Peak containment pres LOCA PCT Local MWR Core wide MWR ASME Peak pressure Overpres sure Stability Stability regions*

L AOO OLMCPR 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, like Figures 4-4 and 4-5 of the Core Operating Limits Report (COLR) submitted by letter dated March 23, 2004 (GNRO-2004/00022).

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 information first, followed by the GGNS results. The major subjects are organized in the same order as the table: ATWS, LOCA, ASME Overpressure, Stability, and AOOs (Including Off-Rated Limits).

to GNRO-2008/00061 Page 3 of 16 ATWS Methods/Codes Comply with all Methodology I applicable Staff Approval Code(s) Used restrictions and conditions ISCOR09 Note 1 Yes PANAC11 NEDE-30130-P-A, Steady-State Nuclear Methods, April Yes 1985. Note 2 ODYN09 NEDC-24154P-A, Qualification of the One-Dimensional Yes Core Transient Model (ODYN) for Boiling Water Reactors, Supplement 4, Volume 1, January 1998.

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

NOTES (1) The ISCOR code is not approved by name. However, the NRC Safety Evaluation (SE) 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-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods",

(TAC NO. MA6481), November 10,1999.

(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.

Results vs. Acceptance Criteria (ATWS)

Parameter Result Acceptance Criteria Peak Vessel Pressure (psig) 1299  ::::;1500 Peak Suppression Pool Temperature (OF) 170  ::::;185 Peak Containment Pressure (psig) 8.2  ::::;15.0 Peak Cladding Temperature (OF) 1509  ::::;2200 Peak Local Cladding Oxidation (%) Insignificant  ::::;17 to GNRO-2008/00061 Page 4 of 16 LOCA Methods/Codes Comply with all Methodology I applicable Staff Approval Code(s) Used restrictions and conditions ISCOR09 Note 1 Yes LAMB08 NEDE-20566P-A, General Electric Analytical Model for Loss-of- Yes Coolant Analysis in Accordance with 10CFR50 Appendix K, September 1986.

SAFER041 NEDE-23785-1-PA Rev. 1, The GESTR-LOCA and SAFER Yes GESTR08 Models for the Evaluation of the Loss-of-Coolant Accident, Vol. 1, Note 4 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 Transient Yes Analysis of a Single Channel, Revision 2, July 2002.

to GNRO-2008/00061 Page 5 of 16 NOTES (1) The ISCOR code is not approved by name. However, the SE 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.

(4) The ECCS-LOCA evaluations are in compliance with the one SE restriction upon the SAFER/GESTR-LOCA methodology. This SE 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.

Results vs. Acceptance Criteria (LOCA)

Parameter ATRIUM-10 Result GE14 Result Acceptance Criteria Licensing Basis  : : ; 1880 of  ::::; 1630 of  ::::; 2200 OF PCT Maximum Local  ::::;3%  : : ; 1%  ::::;17%

Oxidation Core Wide Metal-  : : ; 0.1%  ::::; 0.1%  ::::; 1.0%

Water Reaction Coolable See Licensing Basis See Licensing Basis Satisfied by:

Geometry PCT and Maximum PCT and Maximum PCT::::; 2200 OF Local Oxidation results Local Oxidation results above above AND Maximum Local Oxidation ::::; 17%

Core Long Term Satisfied by: Satisfied by: Core temperature Cooling acceptably low EITHER EITHER AND Core reflooded above Core reflooded above TAF TAF Long-term decay heat removed OR OR 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 to GNRO-2008/00061 Page 6 of 16 Additional LOCA Analysis Information a) Use of Limiting Break and Single Failures The ECCS-LOCA evaluations evaluate all potentially limiting single failures including:

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. 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 requirement for the use of the Moody Slip Flow Model with a discharge coefficient of 1.0, 0.8, and 0.6; the break sizes are selected to fully characterize the PCT versus break size response and are selected to ensure that the limiting break size is identified. The 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 0.08 fe break of the recirculation suction line. The double-ended guillotine break is identified as the overall most limiting break.

b) Use of Limiting Power/Flow Conditions and Axial Power Shape 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 point on the MELLLA rod line with the highest power and lowest flow is identified as the limiting power and flow condition. 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.

c) Applicability of the Legacy Fuel Analysis (MAPLHGR) Through Transition Cycles 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 10CFR50.46 licensing requirements, the one SE restriction, and generates MAPLHGR limits which are applicable to the transition cycles.

to GNRO-2008/00061 Page 7 of 16 ASME Overpressure Methods/Codes Comply with all Methodology I applicable Staff Approval Code(s) Used restrictions and conditions ISCOR09 Note 1 Yes PANAC11 NEDE-30130-P-A, Steady-State Nuclear Methods, April Yes 1985, Note 2 ODYN09 NEDC-24154P-A, Qualification of the One-Dimensional Yes Core Transient Model (ODYN) for Boiling Water Reactors, August 1986.

NOTES:

(1) The ISCOR code is not approved by name. However, the SE 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-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods",

(TAC NO. MA6481), November 10,1999.

Results vs. Acceptance Criteria (ASME Overpressure)

Steam TS Dome ASME Dome Vessel Line Pressure Vessel Event Pressure Pressure Limit (1) Limit (2)

Pressure (psig) (psig)

(psig) (psig) (psig)

MSIV Closure (Flux Scram) - ICF (HBB) 1274 1276 1308 1325 1375 MSIV Closure (Flux Scram) - MEOD 1269 1271 1293 1325 1375 (HBB)

NOTES:

(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 to GNRO-200B/00061 Page B of 16 Stability Methods/Codes Comply with all Methodology I Code(s) Used Staff Approval applicable restrictions and I

conditions ISCOR09 Note 1 Yes PANAC11 NEDE-30130-P-A, Steady-State Nuclear Methods, April Yes 1985.

ODYSY05 NEDC-32992P-A, ODYSY Application for Stability Yes Licensing Calculations, July 2001.

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 E1A, December 1996.

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

NOTES:

(1) The ISCOR code is not approved by name. However, the SE 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 to GNRO-2008/00061 Page 9 of 16 Results vs. Acceptance Criteria (Stability)

Below are the decay ratios resulting from ODYSY analysis for Grand Gulf Cycle 17. Figure 1 shows the stability regions. Figure 2 provides a comparison of the result values to the Acceptance Criteria.

Decay Ratios Cycle 17 Case State Core Decay Ratio Channel Decay Ratio RVM1 B 0.427 0.172 RVM2 A' 0.520 0.280 RVM3 B' 0.587 0.124 RVM4 FRE H1 0.615 0.276 RVM5 FRE HO 0.422 0.021 RVM6 A'LOFH 0.754 0.297 RVM7 B'LOFH 0.709 0.307 to GNRO-2008/00061 Page 10 of 16 Figure 1. GGNS C17 EtA Stability Regions 110 3898 MWth 100 MELLLA Boundary 90 80 "C

Q)

C'I:S 0:::

70 60 100.0% RL

~

0 i o.

Q)

~ 50 0 Monitored 0.. Region Q) io. 40 0

0 30 Natural 20 Circulation FCY-Cavitation 10 Approx Minimum Protection Pump Speed 0

0 10 20 30 40 50 60 70 80 90 100 110 Core Flow (% Rated) to GNRO-2008/00061 Page 11 of 16 Figure 2. GGNS C17 Decay Ratios vs. ODYSY Stability Criteria I I I I I I I I I I I I I I I I I I I I I I I 0.9 ---r---r---r- r - - r --,- r- ,-- I--

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0 0 0.1 0.2 0.3 0.4 0.5 0.6 0.7 0.8 0.9 Channel Decay Ratio to GNRO-2008/00061 Page 12 of 16 AOOs Including Off-Rated Limits Methods/Codes Comply with all Methodology I applicable Staff Approval Code(s) Used restrictions and conditions ISCOR09 Note 1 Yes PANAC11 NEDE-30130-P-A, Steady-State Nuclear Methods, April Yes 1985, Note 2 ODYN09 NEDC-24154P-A, Qualification of the One-Dimensional Yes Core Transient Model (ODYN) for Boiling Water Reactors, August 1986.

NOTES:

(1) The ISCOR code is not approved by name. However, the SE 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-24011-P-A, GESTAR II Implementing Improved GE Steady-State Methods",

(TAC NO. MA6481), November 10,1999.

The following tables provide operating limit information for Cycle 17. Although not in graphic form as normally expressed in the COLR, the information is comparable to the current Cycle 16 power-dependent and flow-dependent operating limits. The COLR figures for Cycle 17 will be provided upon issuance with the normal COLR submittal as required by Technical Specification S.6.S.d.

to GNRO-2008/00061 Page 13 of 16 Limiting Pressurization Events OLMCPR(1) Summary Table Option A Application Condition I Exposure Range GE14C ATR10 Equipment in Service BOC to MOC 1.18 1.19 MOC to EOC 1.31 1.33 FWH 005 BOC to MOC 1.18 1.19 MaC to EOC 1.32 1.33 EOC-RPT 005 BOC to MaC 1.23 1.23 MOC to EOC 1.37 1.39 EOC-RPT + FWH 005 BOC to MOC 1.23 1.23 MOC to EOC 1.39 1.40 NOTES:

(1) The OLMCPR values presented apply to rated power operation based on the two loop operation safety limit MCPR.

(2) =

BOC Beginning of Cycle MOC = Middle of Cycle

=

EOC End of Cycle FWH = Feedwater Heater

=

EOC-RPT End-of-Cycle Recirculation Pump Trip

=

OOS Out of Service

Attachment 1 to GNRO-2008/00061 Page 14 of 16 MCPRp Limits for GE14 and AtriumlO Below Pbypass Above Pbypass App. 25% P 25% P 40%P 40%P Exp. Fuel 40% P 50% P 70% P 70% P 90% P 100% P Group ':::50% flow >50% flow ':::50% flow >50% flow GE14 1.93 2.03 1.73 1.93 1.56 1.52 1.44 1.43 1.23 1.21 MOe Atrium10 1.89 1.99 1.69 1.88 1.56 1.52 1.44 1.40 1.23 1.21 1

GE14 2.02 2.10 1.81 1.93 1.58 1.58 - 1.53 1.35 1.31 EOC Atrium10 1.98 2.06 1.78 1.88 1.57 1.57 - 1.51 1.37 1.33 GE14 1.93 2.03 1.73 1.93 1.56 1.52 1.44 1.43 1.23 1.21 Moe Atrium10 1.89 1.99 1.69 1.88 1.56 1.52 1.44 1.40 1.23 1.21 2

GE14 2.02 2.10 1.81 1.93 1.61 1.58 - 1.53 1.37 1.32 EOC Atrium10 1.98 2.06 1.78 1.88 1.61 1.57 - 1.51 1.39 1.33 GE14 1.93 2.03 1.73 1.93 1.56 1.52 1.44 1.43 1.23 1.23 Moe Atrium10 1.89 1.99 1.69 1.88 1.56 1.52 1.44 1.40 1.24 1.23 3

GE14 2.02 2.10 1.81 1.93 1.61 1.58 - 1.53 1.41 1.37 EOC Atrium10 1.98 2.06 1.78 1.88 1.62 1.58 - 1.52 1.43 1.39 GE14 1.93 2.03 1.73 1.93 1.56 1.52 1.44 1.43 1.24 1.23 Moe Atrium10 1.89 1.99 1.69 1.88 1.56 1.52 1.44 1.40 1.25 1.23 4

GE14 2.02 2.10 1.81 1.93 1.67 1.60 - 1.53 1.44 1.39 EOC Atrium10 1.98 2.06 1.78 1.88 1.67 1.60 - 1.52 1.45 1.40

  • NOTES:

Application Group 1: Equipment in Service Application Group 2: FWH OOS Application Group 3: EOC-RPT OOS Application Group 4: EOC-RPT + FWH OOS

Attachment 1 to GNRO-2008/00061 Page 15 of 16 LHGRFACp Limits for GE14 and AtriumlO Below Pbypass Above Pbypass App. 25% P 25% P 40%P 40%P Exp. Fuel 40%P 50% P 70% P 90% P 100% P Group =:,50% flow >50% flow =:,50% flow >50% flow GE14 0.822 0.767 0.904 0.780 0.911 1.0 1.0 1.0 1.0 MOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 1

GE14 0.822 0.767 0.885 0.780 0.911 1.0 1.0 1.0 1.0 EOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 GE14 0.822 0.767 0.904 0.780 0.911 1.0 1.0 1.0 1.0 MOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 2

GE14 0.822 0.767 0.885 0.780 0.911 1.0 1.0 1.0 1.0 EOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 GE14 0.822 0.767 0.904 0.780 0.911 1.0 1.0 1.0 1.0 MOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 3

GE14 0.822 0.767 0.885 0.780 0.911 1.0 1.0 1.0 1.0 EOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 GE14 0.822 0.767 0.904 0.780 0.911 1.0 1.0 1.0 1.0 MOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0 4

GE14 0.822 0.767 0.885 0.780 0.911 1.0 1.0 1.0 1.0 EOC Atrium10 0.854 0.761 0.911 0.782 0.911 1.0 1.0 1.0 1.0

  • NOTES:

Application Group 1: Equipment in Service Application Group 2: FWH OOS Application Group 3: EOC-RPT OOS Application Group 4: EOC-RPT + FWH OOS to GNRO-200B/00061 Page 16 of 16 MCPRf Limits for GE14 and AtriumlO 20% F 30% F 70% F 110% F 1.30 1.30 1.21 1.21 LHGRFACfLimits for GE14 and AtriumlO 20% F 30% F 60% F 110% F 0.748 0.748 1.000 1.000

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