ML083530599

Proposed Change to Pilgrim'S Technical Specification Concerning the Safety Limit Minimum Critical Power Ratio

ML083530599
Person / Time
Site:	Pilgrim
Issue date:	12/16/2008
From:	Bronson K H Entergy Nuclear Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML083530599 (38)

Text

{{#Wiki_filter:S' Entergy Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station 600 Rocky Hill Road Plymouth, MA 02360 Kevin H. Bronson Site Vice President December 16, 2008 U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555

SUBJECT:

Entergy Nuclear Operations, Inc Pilgrim Nuclear Power Station Docket No. 50-293 License No. DPR-35

REFERENCES:

Prooosed Change to Pilgrim's Technical Specification Concerning the Safety Limit Minimum Critical Power Ratio 1. GNF* Additional Information Regarding the Requested Changes to the Technical Specifications SLMCPR, Pilgrim Cycle 18, Proprietary Version, eDRF Section 000-0090-6001, dated October 16, 2008 2. GNF Additional Information Regarding the Requested Changes to the Technical Specifications SLMCPR, Pilgrim Cycle 18, Non-Proprietary Version, eDRF Section 000-0090-6001, dated October 16, 2008 LETTER NUMBER: 2.08.057

Dear Sir/Madam:

Pursuant to 10 CFR 50.90, Entergy proposes to amend Facility Operating License DPR-35 for the Pilgrim Nuclear Power Station by modifying Technical Specification Section 2.1.2.This application proposes to change the Safety Limit Minimum Critical Power Ratio (SLMCPR)in Technical Specification (TS) Section 2.1.2 from 1.06 to 1.08 for two recirculation loop operation and from 1.08 to 1.11 for single loop operation. In addition, References 5 and 6 from BASES of Section 2.0 SAFETY LIMITS are deleted since the TS Section 5.6.5.b.1 includes the enveloping Reference to NEDE-2401 1-P-A (GESTAR II), Amendment 22 based upon which fuel reload analysis was performed for Operating Cycle 18.Attachment 1 contains a description of the proposed changes, a safety evaluation of the changes, the Determination of No Significant Hazards Consideration, and an Environmental Assessment. Attachment 2 contains the current Technical Specification pages marked up with the proposed revisions. Marked up and Retyped BASES pages are included for information.

• Global Nuclear Fuel -Americas, LLC is a joint venture of General Electric, Toshiba, and Hitachi.AwlI Entergy Nuclear Operations, Inc. Letter Number: 2.08.057 Pilgrim Nuclear Power Station Page 2 Attachment 3 contains the proposed revised Technical Specification and BASES pages.Attachment 4 is a copy of Reference 1, which contains proprietary information, and is sent directly to the NRC Project Manager under separate cover, and includes an affidavit supporting the request that the information contained within the double brackets in the Attachment be considered Global Nuclear Fuel .(GNF) proprietary information as described in 10 CFR 2.390(b)(1)(i)(A).

Therefore, it is requested that Attachment 4 be withheld from public disclosure. Attachment 4 addresses the applicability of the SLMCPR methodology and uncertainties, and verifications. Attachment 5 is a copy of Reference 2, which contains non-proprietary version of Reference 1 and is being submitted to the Document Control Desk.Entergy requests NRC's timely review and approval of this proposed amendment to support startup from Refueling Outage (RFO)-17, which is scheduled to commence on or about April 17, 2009 and restart is expected on or about May 17, 2009. Therefore, in order to support Cycle 18 operation, it is requested the proposed changes be issued by April 17, 2009, to provide sufficient time to revise affected documents prior to startup from RFO-1 7.Following approval of the proposed amendment, the Core Operating Limits Report and applicable operating procedures will be revised prior to start-up from RFO-1 7.This submittal contains no new regulatory commitments. If you have questions regarding this subject, please feel free to contact Mr. Joseph Lynch at (508) 830-8403.I declare under the penalty of perjury that the foregoing information is true and correct.Executed on the day of EzP-_. oe-C" 2008 Sincerely, frKevin R. Bronson Site Vice President Attachments:

1. Description and Evaluation of Proposed Technical Specification Change to Minimum Critical Power Ratio Safety Limit (3 pages)2. Technical Specification Marked Up Pages (3 pages)3. Technical Specification Revised Pages.(3 pages)4. GNF Additional Information Regarding the Requested Changes to the Technical Specifications SLMCPR, Pilgrim Cycle 18, Proprietary Version, eDRF Section 000-0090-6001, dated October 16, 2008 (27 pages)5. GNF Additional Information Regarding the Requested Changes to the Technical Specifications SLMCPR, Pilgrim Cycle 18, Non-Proprietary Version, eDRF Section 000-0090-6001, dated October 16, 2008 (24 pages)

Entergy Nuclear Operations, Inc.Pilgrim Nuclear Power Station cc: Mr. James S. Kim, Project Manager Plant Licensing Branch I-1 Division of Operator Reactor Licensing Office of' Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission One White Flint North O-8C2 11555 Rockville Pike Rockville, MD 20852 Regional Administrator, Region 1 U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Letter Number: 2.08.057 Page 3 Mr. Robert Walker, Director Massachusetts Department of Public Health Schrafft Center Suite 1 M2A Radiation Control Program 529 Main Street Charlestown, MA 02129 Mr. John Giarrusso, Jr.Nuclear Preparedness Manager Mass. Emergency Management Agency 400 Worcester Road Framingham, MA 01702 Senior Resident Inspector Pilgrim Nuclear Power Station ATTACHMENT 1 Description and Evaluation of Proposed Technical Specifications Change to Minimum Critical Power Ratio Safety Limit BACKGROUND Entergy is proposing to amend the Pilgrim Technical Speciation (TS) Section 2.1.2 to revise the Safety Limit Minimum Critical Power Ratio (SLMCPR) for the Operating Cycle (OC) 18. OC 18 commences from the restart of Refueling Outage (RFO) 17. RFO-17 is scheduled to start on or about April 17, 2009 and restart from the outage is expected on or about May 17, 2009.-General Electric (GE) has performed the Pilgrim core configuration analysis using GE reload fuel Type GNF2 (10x10). As a result, the TS Section 2.1.2 is required to be revised for OC 18 to include revised SLMCPR.By a letter dated August 29, 2008, Global Nuclear Fuel (GNF) submitted to the NRC a Report titled "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II), NEDC-333270P, Revision 1". The report provides generic information relative to the GNF2 fuel design and analysis of GE BWRs for which GNF provides fuel. The scope of the assessment is in accordance with the fuel licensing acceptance criteria specified in NEDE-2401 1-P-A (GESTAR II), Amendment 22 process, as specified in TS 5.6.5.b.1. Pilgrim reload analysis for OC 18 follows NEDE-24011-P-A (GESTAR II), Amendment 22 process.The References 5 and 6 included in the BASES of Section 2.0 SAFETY LIMITS are deleted because they are enveloped by the Reference included in TS 5.6.5.b.1 PROPOSED CHANGES Based upon GE calculations for Pilgrim core reload analysis for Operating Cycle 18, the calculated SLMCPR changes from 1.06 to 1.08 for two recirculation loop operation and SLMCPR changes from 1.08 to 1.11 for single recirculation loop operation. Accordingly, TS Section 2.1.2 is revised to read as follows:* "MINIMUM CRITICAL POWER RATIO shall be > 1.08 for two recirculation loop operation and >1.11 for single recirculation loop operation."* In addition, References 5 and 6 from BASES of Section 2.0 SAFETY LIMITS are deleted since the TS Section 5.6.5.b.1 includes>the enveloping Reference to NEDE-24011-P-A (GESTAR II), Amendment 22 based upon which fuel reload analysis was performed for Operating Cycle 18. The BASES changes are provided for information only.REASONS FOR THE PROPOSED CHANGES The current required Safety Limit MCPR (SLMCPR) for Pilgrim Station is 1.06. Calculations performed by Global Nuclear Fuel for Pilgrim Station resulted in a minimum calculated for Operating Cycle 18 SLMCPR value of 1.08.Entergy is proposing to operate with a SLMCPR of 1.08 to account for SLMCPR variances in fuel cycle(s) during Cycle 18 for two loop recirculation loop operation and 1.11 SLMCPR for single recirculation Loop Operation. Deletion of Reference 5 and 6 are administrative, since the Reference included in TS 5.6.5.b.1 envelops these references. Page 1 of 3 ATTACHMENT 1 (cont.)SAFETY EVALUATION The 'Fuel Cladding Integrity Safety Limit is set such that no mechanistic fuel damage is calculated to occur if the limit is not violated. Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not necessarily result in damage to BWR fuel rods, the critical power at which boiling transition is calculated to occur has been adopted as a convenient limit. The uncertainties, however, in monitoring the core operating state and in the procedures used to calculate the critical power result in an uncertainty in the value of the critical power. Therefore, the Fuel Cladding Integrity Safety Limit is defined as the minimum critical power ratio (MCPR) in the limiting fuel assembly for which more the 99.9% of the fuel rods in the core are expected to avoid boiling transition considering the power distribution within the core and all uncertainties Global Nuclear Fuel's calculation of the revised plant-specific SLMCPR value for Pilgrim's Cycle 18 was performed as part of the Reload Licensing Analysis for Pilgrim Cycle 18 and is based upon NRC approved methods specified in the Reference section of the TS BASES for Section 2.0 SAFETY LIMITS. The new Pilgrim Station SLMCPR is 1.08 for two recirculation loop operation and 1.11 for single recirculation loop operation. Based on the above, it is concluded that the proposed SLMCPR values are appropriate for the Pilgrim Cycle 18 core. Entergy will submit Cycle 18 Power/Flow map as part of the Core Operating Limits Report submittal for Cycle 18.The submittal "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II), NEDC-333270P, Revision 1, dated August 29, 2008", provides generic information relative to the GNF2 fuel design and analysis of GE BWRs for which GNF provides fuel. The scope of the assessment is in accordance with the fuel licensing acceptance criteria specified in NEDE-24011-P-A (GESTAR II), Amendment 22 process, as specified in TS 5.6.5.b.1. Pilgrim reload analysis for OC 18 follows NEDE-2401 1-P-A (GESTAR II), Amendment 22 process.Concurrent with this change, the References 5 and 6 included in the BASES of Section 2.0 SAFETY LIMITS are deleted because they are enveloped by the Reference included in TS 5.6.5.b.1. DETERMINATION OF NO SIGNIFICANT HAZARDS CONSIDERATIONS The Code of Federal Regulations, 10 CFR 50.91, requires licensees requesting an amendment to provide an analysis, using the standards in 10 CFR 50.92, which determines whether a significant hazards consideration exists. The following analysis is provided in accordance with 10 CFR 50.91 and 10 CFR 50.92 for the proposed amendment.

1. The proposed changes to Technical Specification do not involve a significant

• increase in the probability of an accident previously evaluated.

The proposed Safety Limit MCPR (SLMCPR), and its use to determine the Operating Cycle 18 thermal limits, have been derived using NRC approved methods specified in the Reference section of the Technical Specification Bases Section for 2.0 SAFETY LIMITS. These methods do not change the method of operating the plant and have no effect on the probability of an accident initiating event or transient. Page 2 of 3 ATTACHMENT 1 (cont.)The basis of the SLMCPR is to ensure no mechanistic fuel damage is calculated to occur if the limit is not violated. The new SLMCPR preserves the margin to transition boiling, and the probability of fuel damage is not increased. Therefore, the proposed changes to Technical Specifications do not involve an increase in the probability or consequences of an accident previously evaluated.

2. The proposed changes to Technical Specifications do not create the possibility of a new or different kind of accident from any accident previously evaluated.

The proposed changes result only from revised methods of analysis for the Cycle 18 core reload. These methods have been reviewed and approved by the NRC, do not involve any new or unapproved method for operating the facility, and do not involve any facility modifications. No new initiating events or transients result from these changes.Therefore, the proposed changes to technical specifications do not create the possibility of a new or different kind of accident from any accident previously evaluated.

3. The proposed changes to Technical Specifications do not involve a significant reduction in a margin of safety.The margin of safety as defined in the TS bases will remain the same. The new SLMCPR was derived using NRC approved methods which are in accordance with the current fuel design and licensing criteria.

The SLMCPR remains high enough to ensure that greater than 99.9% of all fuel rods in the core will avoid transition boiling if the limit is not violated, thereby preserving the fuel cladding integrity. Therefore, the proposed changes to technical specifications do not involve a significant reduction in the margin of safety.The proposed changes have been reviewed and recommended for approval by the Pilgrim Station On-site Safety Review Committee (OSRC) and Off-site Safety Review Committee (SRC).ENVIRONMENTAL IMPACT The proposed Technical Specification changes were reviewed against the criteria of 10 CFR 51.22 for environmental considerations. The proposed changes do not involve a significant hazards consideration, a significant increase in the amounts of effluents that may be released offsite, or a significant increase in individual or cumulative occupational radiation exposure.Based on the foregoing, Entergy concludes the proposed Technical Specifications meet the criteria in 10 CFR 51 .22(c)(9) for a categorical exclusion from the requirements for an Environmental Impact Statement. Page 3 of 3 ATTACHMENT 2 Technical Specification Marked Up Pages (3 pages)

2.0 SAFETY

LIMITS 2.1 Safety Limits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% of rated core flow: THERMAL POWER shall be < 25% of RATED THERMAL POWER.2.1.2 With the reactor steam dome pressure > 785 psig and core flow >10% of rated core flow: MINIMUM CRITICAL POWER RATIO shall be > 4-106 1.08 for two recirculation loop operation ei and > 4-.08 1.11 for single recirculation loop operation.

2.1.3 Whenever

the reactor is in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the normal active fuel zone.2.1.4 Reactor steam dome pressure shall be < 1325 psig at any time when irradiated fuel is present in the reactor vessel.2.2 Safety Limit Violation With any Safety Limit not met within two hours the following actions shall be met: 2.2.1 Restore compliance with all Safety Limits, and 2.2.2 Insert all insertable control rods.Amendment No. 15, 27, 42, 72, 433, 146, 17!, 191, 219, 223 2-1 BASES: 2.0 SAFETY LIMITS (Cont)FUEL CLADDING INTEGRITY (2.1.1)(Cont)MINIMUM CRITICAL POWER RATIO (2.1.2)Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (2), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. Instead of the standard GETAB model uncertainties, revised uncertainties in accordance with references 3 and 4 were used to calculate the SLMCPR.The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) -Boilin GEXL, correlation. The range of validity thfe GEXL correlation is speci ie in References 5 and 6.The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical power at (Cont)Revision N 1 51-Amendment No. 15 4,t, 72, 105ý 13, 171, 49 B2-2 BASES: 2.0 SAFETY LIMITS (Cont)REACTOR STEAM DOME PRESSURE (2.1.4)REFERENCES The Safety Limit for the reactor steam dome pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code (1965 Edition, including the January 1966 Addendum), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor coolant system is designed to ASME Section III for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1148 psig at 562 0 F for suction piping and 1241 psig at 562 0 F for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes.1) "General Electric Standard Application for Reactor Fuel," NEDE-2401 1-P-A (through the latest approved amendment at the time the reload analyses are performed as specified in the CORE OPERATING LIMITS REPORT).2) General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Co. BWR Systems Department, January 1977, NEDE-1 0958-PA and NEDO-10958-A.

3) "Methodology

& Uncertainties for SLMCPR Evaluations," NEDC-32601-P-A (August 1999).4) "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," NEDC-32694-P-A (August 1999).5) "GE 11 Cmpliap with Amendm.nt 22 of GESTAR7A" NEDE 31917P (April 1991).6}"GE 14 Geomaliancc wIth Amendment 22 of GESTAR II," NEDO 32868P (D eVe FnbeF4998) 7 Revision 2-Amendment No. 15, 133, 16, 171, 191 B2-4 ATTACHMENT 3 Technical Specification Revised Pages (3 pages)

2.0 SAFETY

LIMITS 2.1 Safety Limits 2.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% of rated core flow: THERMAL POWER shall be < 25% of RATED THERMAL POWER.2.1.2 With the reactor steam dome pressure > 785 psig and core flow >10% of rated core flow: MINIMUM CRITICAL POWER RATIO shall be > 1.08 for two recirculation loop operation and > 1.11 for single recirculation loop operation.

2.1.3 Whenever

the reactor is in the cold shutdown condition with irradiated fuel in the reactor vessel, the water level shall not be less than 12 inches above the top of the normal active fuel zone.2.1.4 Reactor steam dome pressure shall be < 1325 psig at any time when irradiated fuel is present in the reactor vessel.2.2 Safety Limit Violation With any Safety Limit not met within two hours the following actions shall be met: 2.2.1 Restore compliance with all Safety Limits, and 2.2.2 Insert all insertable control rods.Amendment No. 15, 27, 12, 72, 133, 146, 171, 191, 219, 223 2-1 BASES: 2.0 SAFETY LIMITS (Cont)FUEL CLADDING INTEGRITY (2.1.1)(Cont)MINIMUM CRITICAL POWER RATIO (2.1.2)Since the pressure drop in the bypass region is essentially all elevation head, the core pressure drop at low power and flows will always be greater than 4.5 psi. Analyses show that with a bundle flow of 28 x 103 lbs/hr, bundle pressure drop is nearly independent of bundle power and has a value of 3.5 psi. Thus, the bundle flow with a 4.5 psi driving head will be greater than 28 x 103 lbs/hr. Full scale ATLAS test data taken at pressures from 14.7 psia to 800 psia indicate that the fuel assembly critical power at this flow is approximately 3.35 MWt. With the design peaking factors, this corresponds to a THERMAL POWER of more than 50% of RATED THERMAL POWER. Thus, a THERMAL POWER limit of 25% of RATED THERMAL POWER for reactor pressure below 785 psig is conservative. The Safety Limit MCPR is determined using the General Electric Thermal Analysis Basis, GETAB (2), which is a statistical model that combines all of the uncertainties in operating parameters and the procedures used to calculate critical power. Instead of the standard GETAB model uncertainties, revised uncertainties in accordance with references 3 and 4 were used to calculate the SLMCPR.The probability of the occurrence of boiling transition is determined using the General Electric Critical Quality (X) -Boiling Length (L), GEXL, correlation. The fuel cladding integrity Safety Limit is set such that no fuel damage is calculated to occur if the limit is not violated.Since the parameters which result in fuel damage are not directly observable during reactor operation, the thermal and hydraulic conditions resulting in a departure from nucleate boiling have been used to mark the beginning of the region where fuel damage could occur. Although it is recognized that a departure from nucleate boiling would not result in damage to BWR fuel rods, the critical power at (Cont)Revision Amendment No. 15, 42, 72, 105, 129, 133, 165, 171, !9!132-2 BASES: 2.0 SAFETY LIMITS (Cont)REACTOR STEAM DOME PRESSURE (2.1.4)REFERENCES The Safety Limit for the reactor steam dome pressure has been selected such that it is at a pressure below which it can be shown that the integrity of the system is not endangered. The reactor pressure vessel is designed to Section III of the ASME Boiler and Pressure Vessel Code (1965 Edition, including the January 1966 Addendum), which permits a maximum pressure transient of 110%, 1375 psig, of design pressure 1250 psig. The Safety Limit of 1325 psig, as measured by the reactor steam dome pressure indicator, is equivalent to 1375 psig at the lowest elevation of the reactor coolant system. The reactor coolant system is designed to ASME Section III for the reactor recirculation piping, which permits a maximum pressure transient of 120% of design pressures of 1148 psig at 562 0 F for suction piping and 1241 psig at 562 0 F for discharge piping. The pressure Safety Limit is selected to be the lowest transient overpressure allowed by the applicable codes.1) "General Electric Standard Application for Reactor Fuel," NEDE-2401 1-P-A (through the latest approved amendment at the time the reload analyses are performed as specified in the CORE OPERATING LIMITS REPORT).2) General Electric Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, General Electric Co. BWR Systems Department, January 1977, NEDE-10958-PA and NEDO-10958-A.

3) "Methodology

& Uncertainties for SLMCPR Evaluations," NEDC-32601-P-A (August 1999).4) "Power Distribution Uncertainties for Safety Limit MCPR Evaluations," NEDC-32694-P-A (August 1999).Revision Amendment No. 15, 133, 146, 171, 191 132-4 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment 10/16/2008 eDRFSection: 0000-0090-6001 GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR Pilgrim Cycle 18 Pilgjim Cycle 18 Page I of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Proprietary Information Notice This document is the GNF non-proprietary version of the GNF proprietary report. From the GNF proprietary version, the information denoted as GNF proprietary (enclosed in double brackets) was deleted to generate this version.Proprietary Information Notice Pag-e 2 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table of Contents 1.0 M ET H O DO LO G Y .......................................................................................................................................... 4 2.0 D ISC U SSIO N ................................................................................................................................................... 4 2.1. MAJOR CoNT11I1.JTOH.H S"T1O SLMCPRCII:\ ................ ................ ............. ................................. 4 2.2. Dji:v\TlNS IN NRC-Ai'Pii;,oI) UNCERTAINTIES ........... ..................................................... 5 2.2 .1. R -a cto ..... ...........................................................................................................................................

2.2.2. C'ore

hm. Rale d R(ol d oalln , / tix e "'11' Reading ...................... ....................... 2.3. DEPARI, RElw FiRi NRC-APiP, I'ID) Mn IOT Il)Ol,(X;Y ................................................... 6 2.4. F uii ,. A XI.m\ , POWV IR SII,\PI- E T ................................................

....................................................

6 2.5. i io i .R IC i N S ........................................ ............ ........................................... 7 2.6 .M IN I tI I C o m -. F Io w C O ND\D ITIO N ................................................................................................................ 8 2.7. L iIm T IMN ; C O N T RO l. R o i) P.A II R NS ................................................ ............................................................ 2.8 .C o Ru 5 M O N I'IO R INC S Y S, T I.:m .......................................................................................................................... 8 2.9. Powm ;I. /FI,.o\v M A I .................................................................... ..................................... ....8 2 .I0 .C o R ' L o ..\i)iN (, D I.\.O -\ I ................................. ........ .. ....... ................. ............................... .. .... ......9 2.11. Fi..uiz.: Ri:FI:RI c: S ............................................. ....... ....... .............. ............ ........... )2.12. AlI)ION-\r. SLMCPR LICE:NSING CONDITIONS ................................... ........... 9 2 .13 .S I JIX .I A N. ................................................................................... ............ ........................... ........9 3.0 R E FER EN C ES ............................................................................................................................................... 10 List of Figures FIoiE.F I. CuI',I:Nr C-cii.; CoRF LOADINo DIAioR.. ........... ................... ............................................... I I FI(iUI( 2. PREIvious CYCI:, CoRi: LOA1ING DIARAM .............................. ................................. 123, Fio iu;i : 4.1 ilizox NEDC-32601-P-A ................................................................ .................. 13 Fic;u;izl., 4. .FiRiEm 111.5-I FRoM NEDC-3260 I P-A ..................................... ........... ........... ... 14 FiGiuRE 5. .I 1.5'-2 nio:. NEDC-32601 P-A .................................................................. ................ 15 List of Tables T.mBEE 1. DESCR I PTION OF CORE ................. ......... ............ ......... ......... ... .................................. ....... 16 T.\IAi.,I:

2. SLMCPR Cl,\I.C IIo.N iolt s ..............

........... ....... ................ 17 T.\1L. 3. MON'1 C, ARLO CAI.,CuUII.x'r;D SLMCPR vs. ESTiIx.:Vf- .. .................................................................. 18 T,:\i3Ei 4. N ON-PoWvi{R D isTR ii3iTI-i N UNCERTAINTIES .......................................... ............................... 20 T.2\imE 5. PowmIu DISTRIBUTION UNCERTAINTi-S ................ ........ ..................................... ..... 22 T.,i i.E 6. C RITICA L. POWVER U NCIEIR. INTIE-S ............................. ............................. ....................................... 24 Table of Contents Page 3 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachmcnt

1.0 Methodology

GNF performed the Pilgrim Cycle 18 Safety Limit Minimum Critical Power Ratio (SLN4CPR)calculation in accordance to NEDE-2401 I-P-A "General Electric Standard Application for Reactor Fuel" (Revision

16) using: the following NRC-approved methodologies and uncertainties:

• NEDC-32601P-A "MN'ethodology and Uncertainties for Safety Limit MCPR Evaluations" (August 1999)." NEDC-32694P-A "Power Distribution Uncertainties for Safety Limit MCPR Evaluations" (August 1999).* NEDC-32505P-A "R-Factor Calculation Method for GE II, GE 12 and GE 13 Fuel" (Revision 1, July 1999)." NEDO-10958-A "General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application" (January 1977).Table 2 identifies the actual methodologies used for the previous cycle I7 and the current cycle 18 SLMCPR calculations.

2.0 Discussion

In this discussion, the TLO nomenclature is used for two recirculation loops in operation, and the SLO nomenclature is used for one recirculation loop in operation. 2A1. Major Contributors to SLMCPR Change In general, the calculated safety limit is dominated by two key parameters: (1) flatness of the core bundle-by-bundle MCPR distribution, and (2) flatness of the bundle pin-by-pin power/R-Factor distribution. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher calculated SLMCPR. MIP (MCPR Importance Parameter) measures the core bundle-by-bundle MCPR distribution and RIP (R-Factor Importance Parameter) measures the bundle pin-by-pin power/R-Factor distribution. The impact of the fuel loading pattern on the calculated TLO SLMCPR using rated core power and rated core flow conditions has been correlated to the parameter MIPRIP, which combines the MIP and RIP values.Table 3 presents the MIP and RIP parameters for the previous cycle and the current cycle along with the TLO SLMCPR estimate using the MIPRIP correlation. If tile minimum core flow case is applicable, the TLO SLMCPR estimate is also provided for that case although the MIPRIP correlation is only applicable to the rated core flow case. This is done only to provide some reasonable assessment basis of the minimum core flow case trend. In addition, Table 3 presents estimated impacts on the TLO SLMCPR due to methodology deviations, penalities, and/or uncertainties deviations from approved values. Based on the MIPRIP correlation and any Methodology Paoe 4 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment impacts due to deviations from approved values, a final estimated TLO SLMCPR is determined. Table 3 also provides the actual calculated Monte Carlo SLMCPRs. Given the bias and uncertainty in the MIPRIP correlation [[f 1l]and the inherent variation in the Monte Carlo results '3', the change in the Pilgrim Cycle 18 calculated Monte Carlo TLO SLMCPR using rated core power and rated core (low conditions is consistent with the corresponding estimated TLO SLMCPR value.2.2. Deviations in NRC-Approved Uncertainties Tables 4 and 5 provide a list of NRC-approved uncertainties along with values actually used. A discussion of deviations from these NRC-approved values follows; all of which are conservative relative to NRC-approved values. Also, estimated impact on the SLMCPR is provided in Table 3 for each deviation. 2.2.1. R-Factor At this time, GNF has generically increased the GEXL R-Factor uncertainty from 131 to account for an increase in channel bow due to the emerging unforeseen phenomena called control blade shadow corrosion-induced channel bow,ýwhich is not accounted for in the channel bow uncertainty component of the approved R-Factor uncertainty. The step "o RPEAK" in Figure 4.1 from NEDC-32601P-A, which has been provided for convenience in Figure 3 of this attachment, is affected by this deviation. Reference 4 technically justifies that a GEXL R-Factor uncertainty of [[ ]] accounts for a channel bow uncertainty of up to [1 Pilgiim has experienced control blade shadow corrosion-induced channel bow to the extent that an increase in the NRC-approved R-Factor uncertainty 3 is deemed prudent to address its impact. Accounting for the control blade shadow corrosion-induced channel bow, the Pilgrim Cycle 18 analysis shows an expected channel bow uncertainty of [[ 3 which is bounded by a GEXL R-Factor uncertainty of [[ 1:fl., Thus the use of a GEXL R-Factor uncertainty of [[ f l] adequately accounts for the expected control blade shadow corrosion-induced channel bow for Pilgrim Cycle 18.2.2.2. Core Flow Rate and Random Effective TIP Reading At this time, GNF has not been able to show that the NRC-approved process to calculate the SLMCPR only at the rated core power and rated core flow condition is adequately bounding relative to the SLMCPR calculated at rated core power and minimum core flow, see Reference 5.The minimum core flow condition can be more limiting due to the control rod pattern used.GNF has modified the NRC-approved process for determining the SLMCPR to include analyses at the rated core power and minimum licensed core flow point in addition to analyses at the rated core power and rated core flow point. GNF believes this modification is conservative and may in the future provide justification that the original NRC-approved process is adequately bounding.For the TLO calculations performed at 76.7% core flow, the approved uncertainty values for the core flow rate (2.5%)' and the random effective TIP reading (1.2%) are conservatively adjusted Discussion Page 5 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment by dividing them by 76.7/100, The steps "a CORE FLOW" and "a TIP (INSTRUMENT)" in Figure 4.1 from NEDC-32601P-A (which has been provided for convenience in Figure 3 of this attachment) are affected by this deviation, respectively. Historically, these values have been construed to be somewhat dependent oil the core flow conditions as demonstrated by the fact that higher values have always been used when performing SLO calculations. It is for this reason that GNF determined that it is appropriate to consider an increase in these two uncertainties when the core flow is reduced. The amount of increase is determined in a conservative way. For both parameters it is assumed that the absolute uncertainty remains the same as the flow is decreased so that the percentage uncertainty increases inversely proportional to the change in core flow. This is conservative relative to the core flow uncertainty since the variability in the absolute flow is expected to decrease somewhat as the flow decreases. For the random effective TIP uncertainty, there is no reason to. believe that the percentage uncertainty should increase as the core flow decreases for TLO.Nevertheless, this uncertainty is also increased as is done in the more extreme case for SLO primarily to preserve the historical precedent established by the SLO evaluation. Note that the TLO condition is different than the SLO condition because for TLO there is no expected tilting of the core radial power shape.The treatment of the core flow and random effective TIP reading uncertainties is based on the assumption that the signal to noise ratio deteriorates as core flow is reduced. GNF believes this is conservative and may in the future provide justification that the original uncertainties (non-flow dependent) are adequately bounding.The core flow and random TIP reading uncertainties used in the SLO minimum core flow SLMCPR analysis remain the same as in the rated core flow SLO SLMCPR analysis because these uncertainties (which are substantially larger than used in the TLO analysis) already account for the effects of operating at reduced core flow.2.3. Departure from NRC-Approved Methodology No departures from NRC-approved methodologies were used in the Pilgrim Cycle 18 SLMCPR calculations. 2.4. Fuel Axial Power Shape Penalty At this time, GNF has determined that higher uncertainties and non-conservative biases in the GEXL correlations for the various types of axial power shapes (i.e., inlet, cosine, outlet and double hump) could potentially exist relative to the NRC-approved methodology values, see References 3, 6, 7 and 8. The following table identifies, by marking with an 'X", this potential for each GNF product line currently being offered: Discussion Page 6 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment 1]3: 1.1 Axial bundle power shapes corresponding to the limiting SLMCPR control blade patterns are determined using the PANACEA 3D core simulator. These axial power shapes are classified in accordance to the following table: If the limiting bundles in the SLMCPR calculation exhibit an axial power shape identified by this table, GNF penalizes the GEXL critical power uncertaintics to conservatively account for the impact of the axial power shape. Table 6 provides a list of the GEXL critical power uncertainties determined in accordance to the NRC-approved methodology contained in NEDE-2401 I-P-A along with values actually used.For the limiting bundles, the fuel axial power shapes in the SLMCPR analysis were examined to determine the presence of axial power shapes identified in the above table. These power shapes were not found; therefore, no power shape penalties were applied to the calculated Pilgrim Cycle 18 SLMCPR values.2.5. Methodology Restrictions The four restrictions identified on Page 3 of NRC's Safety Evaluation relating to the General Electric Licensing Topical Reports NEDC-32601P, NEDC-32694P, and Amendment 25 to NEDE-2401 I-P-A (March I1, 1999) are addressed in References 1,2, and 3.Discussion Page 7 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment No new GNF fuel designs are being introduced in Pilgrim Cycle 18; therefore, the NEDC-32505-P-A statement ",,.if new fuel is introducted, GENE must confirm that the revised R-Factor method is still valid based on new test data" is not applicable. The GNF2 product line is not considered a new fuel design relative to the GE14 product line, as both consist of 10 x 10 lattice designs.2.6. Minimum Core Flow Condition For Pilgrim Cycle 18, the minimum core flow SLMCPR calculation performed at 76.7% core flow and rated core power condition was limiting as compared to the rated core flow and rated core power condition. For convenience, Figures 111.5-1 and 111.5-2 from NEDC-32601P-A have been provided in Figures 4 and 5, respectively, in order to show this minimum core flow conditon relative relationship to the data on these figures. For this condition the MIP ,, Therefore, this demonstrates that the MIP criterion for determining what constitutes a reasonably bounding limiting rod pattern is still valid for this minimum core flow condition. Hence, the rod patterns used to calculate the SLMCPR at 100 percent rated power/76.7 percent rated flow reasonably assures that at least 999% of the fuel rods in the core would not be expected to experience boiling transition during normal operation or anticipated operational occurrences during the operation of Pilgrim Cycle 18. Consequently, the SLMCPR value calculated from the 76,7% core flow and rated core power condition limiting MCPR distribution reasonably bounds this mode of operation for Pilgrim Cycle 18.2.7. Limiting Control Rod Patterns The limiting control rod patterns used to calculate the SLMCPR reasonably assures that at least 99.9% of the fuel rods in the core would not be expected to experience boiling transition during normal operation or anticipated operational occurrences during the operation of Pilgrim Cycle 18.2.8. Core Monitoring System For Pilgrim Cycle 18, the 3D Monicore system will be used as the core monitoring system.2.9. Power/Flow Map The utility has provided the current and previous cycle power/flow map in a separate attachment. Discussion Page 8 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment 2.10. Core Loading Diagram Figures I and 2 provide the core loading diagram for the current and previous cycle respectively. which are the Reference Loading Pattern as defined by NEDE-2401 I-P-A. Table I provides a description of the core.2.11. Figure References Figure 3 is Figure 4.1 friom NEDC-32601-P-A. Figure 4 is Figure 111.5-1 from NEDC-32601P-A. Figure 5 is Figure 111.5-2 from NEDC-32601 P-A.2.12. Additional SLMCPR Licensing Conditions For Pilgrim Cycle 18, no additional SLMCPR licensing conditions are included in the analysis.2.13. Summary The requested changes to the Technical Specification SLMCPR values are 1.08 for TLO and Il I for SLO for Pilgrim Cycle 18.Discussion Page 9 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment

3.0 References

1 Letter., Glen A. Watford (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to R. Pulsifer (NRC), "Confirmation of lOx 10 Fuel Design Applicability to Improved SLMCPR, Power Distribution and R-Factor Methodologies", FLN-2001-016, September 24, 2001.2. Letter, Glen A. Watford (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to J. Donoghue (NRC), "Confirmation of the Applicability of the GEXL 14 Correlation and Associated R-Factor Methodology for Calculating SLMCPR Values in Cores Containing GE14 Fuel", FLN-2001-017, October 1, 2001.3. Letter, Glen A. Watford (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to Joseph E. Donoghue (NRC), "Final Presentation Material for GEXL Presentation -February 11, 2002", FLN-2002-004, February 12, 2002.4. Letter, John F. Schardt (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to Mel B. Fields (NRC), "Shadow Corrosion Effects on SLMCPR Channel Bow Uncertainty", FLN-2004-030, November 10, 2004.5. Letter, Jason S. Post (GENE) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to Chief, Information Management Branch, et al. (NRC), "Part 21 Final Report: Non-Conservative SLMCPR", MFN 04-108, September 29, 2004.6. Letter, Glen A. Watford (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to Alan Wang (NRC), "NRC Technology Update -Proprietary Slides -July 31I -August I, 2002", FLN-2002-015, October 311, 2002.7. Letter, Jens G. Munthe Andersen (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with attention to Alan Wang (NRC), "GEXL Correlation for lOX 10 Fuel", FLN-2003-005, May 31, 2003.8. Letter, Andrew A. Lingenfelter (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with cc to MC Honcharik (NRC), "Removal of Penalty Being Applied to GE 14 Critical Power Correlation for Outlet Peaked Axial Power Shapes", FLN-2007-031, September 18, 2007, References Page 10 of24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment 52 SIQ]so Fo--jKj %E EALE 12 1 0 -EE]Ci[M 38 -ME]J417 M 32----[DFE[N MK f[II] E IDlj f~l9 M ME r I I I I][E2 E6FI :f X F A z H J[ ]EM[Df O MlED[ Jl[ f I L%@% ýt ti[I -LM11-0 M]E T -M M 7 N FD [C E [] E] M D f W6 r6 FR]E]I]II11M 11~n n+E]%[EFR] HH ll] 2 [K E]R==M %@[flS% E F2 W-AFI ffi[E OMQ E [@%@M6ME EMEDE [ý ]EME Eýv týýOD[1 L1E [:][:] J :C 6M6NE FL 6][f 103LY 0 MEA% E[f EM L TBE 1 3 5 7 9 u T13 15 17 19 21 25 25 21 29 31 33 35 37 39 4114 45 1 49 51:ueI Type A = GEl 4-P1 ODNAB404-5G6.01 0G5.0-1 0OT-1 45-T6-2960 B = GE 14-P 1 ODNAB398-5G6.0/1 GS.0-1 COT-1 45-T6-2958 C = GEl 4-P10DNAB397-10G6.013G5.0-100T-145-T6-2613 D = GEl 4-PI ODNAB399-1 0G6.0/3G5.0/1 G2.0-1 OT-1 45-T6-2828 E = GNF2-P1 0DG2B388-6G6.O07G5.0-1 00T2-1 45-T6-3143 F = GE 14-P 1 0DNAB398-8G6.0/5G5.0/1 G2.0-1 OT-1 45-T6-2614 G = GEl 4-P10 DNAB400-13GZ-1 COT-1 45-T6-2959 H = GE14-P10DNAB399-15GZ-100T-145-T6-2957 I = GNF2-P10DG2B389-6G6.018G50-10CT2-145-T6-3141 J = GE 1 4-P 10DNAB398-8G6.0/5G5.C12G4O-1 00T-1 45-T6-2829 K GE 14-Pi0DNAB397-14GZ-10OT-145-T6-2621 L GEl 4-PI 0DNAB400-12G6.0!3G5.0-10OT-145-T6-2961 M GNF2-PiODG2B389-6G6.0/2G5.0/6G4.0-100T2-145-T6-3142 Figure 1. Current Cycle Core Loading Diagram Figure 1. Current Cycle Core Loading Diagrama Page I1I of 24 GNF NON-PROPRIETARY INFORMATION Cla GNF Ati 5250E D ID 50 FI:I OGD 36ý0o E'VE El'l-S E 268- 21 tmW M V M fl8 f [fl ] [I] n1 ELI 3 0[lE[ V ©o [f I B[%i1 [-]a-S F2] ovJF, 32-- E EA[][ [ ] []F [] ]20- --%% -[T-- O-rE,] %___,L,% ,@ t'[V 4%-- I I 12 F ] H f [E-lf-f] EK IF [f IE IE [-' IG tachmncnt~E -%FlJW EK jLK] AD Q [H+% +[A1Fk ME 1WW I :- I I' I ETLK] [ @,ff Ei-r EA ] [ýt ff[D[:A I M nI I DL 1H I1 E-L] [K] [: ] [E.] F 5DI- ITO -lE[Afl EFýE (T m T O D R If R Ell7, E~li I-- QE M WE-KýElh EID ME~FM RFf[EE]i~1 3 5 7 9 Ii 13 15 17 19 21 23 25 27 29 3t 33 35 37 39 1 'ý3 i5 /17 ý9 51=uel Type A = GEl 4-P1 ODNAB404-5G6.0/1 0G5.0-1 OOT-1 45-T6-2960 B = GE 14-P 10 DNA B398-5G6.0/1 0G5.0-1 00T-1 45-T6-2958 C = GEl 4-P1 0DNAB399-10 G6.0/3G5.0/1 G2.0-1 OOT-1 45-T6-2828 D = GEl 4-P IODNAB400-13GZ-1OOT-145-T6-2959 E = G E14-P1ODNAB399-15GZ-10OT-145-T6-2957 F = GE14-P1ODNAB412-16GZ-10OT-145-T6-3901 G = GE 14-P 1 0DNAB397-1 0G6./3G5.0-1 00T-1 45-T6-2613 H = GE1 4-P10 D NAB398-8G6.0/5G5.0/2G4.0-1 OT-1 45-T6-2829 I = GEl 4-P1 CCNAB398-5G6.0tIOG5.0-1 0OT-145-T6-2958 J = GE14-P1OCNAB400-1 2G6.O/3G5.0-ICOT-1 45-T6-2961 K= GE1 4-P10DNAB397-14GZ-10OT-145-T6-2621 L = GEi4-P10DNAB400-12G6.0/3G5.0-10OT-145-T6-2961 M = GE 1 4-P 1i DNAB398-8G6.05G5.0/1 G2.0-1 COT-1 45-T6-2614 Figure 2. Previous Cycle Core Loading Diagram Figure 2. Previous Cycle Core Loading Diagram Page 12 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment [[P ý ] ]Figure 3. Figure 4.1 from NEDC-32601-P-A Figure 3. Figure 4.1 from NEDC-3260 1-P-A Palge I') of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment K Figure 4. Figure [11.5-1 from NEDC-32601P-A Figure 4. Figure 111.5-1 from NEDC-32601P-A I Page 14 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment' [[Figure 5. Figure [11.5-2 from NEDC-32601P-A Figure 5. Figure 111. 5-2 from NEDC-,32601P-A Page 15 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 1. Description of Core Previous Cycle Previous Cycle Rated Current Cycle Current Cycle Rated Description Minimum Core Flow Core Flow Limiting Minimum Core Flow Core Flow Limiting Limiting Case Case Limiting Case Case Number of Bundles in tile 580 580 Core Limiting Cycle Exposure Point (i.e. EOC EOC EOC BOC BOC/MOC/EOC) Cycle Exposure at Limiting Point 10600 10600 9500 200 (MWd/STU) I% Rated Core Flow 76.7 100 76.7 100 Reload Fuel Type GE 14 GNF2 Latest Reload Batch 27.6 26.9 Fraction, %Latest Reload Average Batch Weight % 4.00 3.89 Enrichment Core Fuel Fraction: GE 14 1.0 0.731 GNF2 0.0 0269 Core Average Weight % 4.01 3.96 Enrichment -I.9 Table 1. Description of Core.Page 16 o1-24 -J GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 2. SLMCPR Calculation Methodologies Previous Cycle Previous Cycle Rated Current Cycle Current Cycle Rated Description Minimum Core Flow Core Flow Limiting Minimum Core Flow Core Flow Limiting Limiting Case Case Limiting Case Case Non-power Distribution NEDC-32601-P-A NEDC-32601-P-A Uncertainty Power Distribution NEDC-3260 I -P-A NEDC-32601 -P-A Methodology Power Distribution NEDC-32694-P-A NEDC-32694-P-A Uncertainty Core Monitoring System 3D Monicore 3D Monicore Table 2. SLMCPR Calculation Methodologies Pa-e 17 of.'24 v GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 3. Monte Carlo Calculated SLMCPR vs. Estimate Previous Cycle Previous Cycle Rated Current Cycle Current Cycle Rated Description Minimum Core Flow Core Flow Limiting Minimum Core Flow Core Flow Limiting Limiting Case Case Limiting Case Case[[Table 3. Monte Carlo Calculated SLMCPR vs. Estimate Page 18 ot'24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 3. Monte Carlo Calculated SLMCPR vs. Estimate Previous Cycle Previous Cycle Rated Current Cycle Current Cycle Rated Description Minimum Core Flow Core Flow Limiting Minimnu Core Flow Core Flow Limiting Limiting Case Case Limiting Case Case Table 3. Monte Carlo Calculated SLMCPR vs. Estimate Page 19 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 4. Non-Power Distribution Uncertainties Nominal (NRC- Previous Cycle Previous Cycle Current Cycle Current Cycle Approved) Value Niiimuin Core Rated Core Flow lMinimum Core Rated Core Flow++/- (%) Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case GETAB Feedwater Flow Measur F 1.76 N/A N/A N/A N/A Measurement Feedwater Temperature 0.76 N/A N/A N/A N/A Measurement Reactor Pressure 0.50 N/A N/A N/A N/A Measurement Core Inlet Temperature 0.20 N/A N/A N/A N/A Measurement Total Core Flow ToalCremFlow6.0 SLO/2.5 TLO N/A N/A N/A N/A Measurement Channel Flow Area Chatn 3.0 N/A N/A N/A N/A Variation Friction Factor N/A Multiplier " l0.0 N/A N/A NiA Channel .Friction FactorMutipi 5.0 N/A N/A N/A N/A Factor Multiplier Table 4. Non-Power Distribution Uncertainties Palge 20 of'24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 4. Non-Power Distribution Uncertainties Nominal (NRC- Previous Cycle Previous Cycle Current Cycle Current Cycle Approved) Value Minimum Core Rated Core Flow Minimum Core Rated Core Flow+ (%) Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case NE DC-3260 I-P-A Feedw ater Flo wv [1*[1 1-1l Measurement Feedwater Temperature 3 [[ [3}.1 [[ ,i [ I Measurement Reactor Pressure Measurement [1 [[ [[ 11 [[ i1 [[ .Core Inlet Temperature 0.2 0.2 0.2 0.2 0:2 Measurement Total Core Flow 6.0 SLO/2.5 TLO 6.0 SLO/3.26 TLO 6.0 SLO/2.5 TLO 6.0 SLO/3.26 TLO 6.0 SLO/2.5 TLO Measurement Channel Flow Area 3[]]Variation Friction Factor [ ii]]Multiplier Channel Friction Factor Multiplier 5.0 5.0 5.0 5.0 5.0 Table 4.. Non-Power Distribution Uncertainties Page 21 of 24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 5. Power Distribution Uncertainties Nominal (NRC- Previous Cycle Previous Cycle Current Cycle Current Cycle Description Approved) Value Minimum Core Rated Core Flow Mininium Core Rated Core Flow+/- a (%) Flow Limiting Case Limiting Case Flow Liniiing Case Limiting Case GETAB/NEDC-3260 I-P-A GEXL R-Factor , N/A N/A N/A N/A Random Effective Rad ive 2.85 SLO/l.2 TLO N/A N/A N/A N/A TIP Reading Systematic Effective 8.6 N/A N/A TIP Reading N/A N/A NEDC-32694-P-A, 3DMONICORE GEXL R-Factor 3 [[ .jl [[ ,31l1 [[ .[1 ;31 Random Effective 2.85 SLO/1.2 TLO 2.85 SLO/1.56 TLO 2.85 SLO/1 .2 TLO 2.85 SLO/1.56 TLO 2.85 SLO/1.2 TLO TIP Reading _________ ________TIP Integral 131 [[[ II Four Bundle Power Distribution i[ 3" Surrounding TIP Location Contribution to Bundle Power Uncertainty Due to i [[Jl 1 III [[ "I LPRM Update Table 5. Power Distribution Uncertainties Paue 22 of'24 C GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 5. Power Distribution Uncertainties Nominal (NRC- Previous Cycle Previous Cycle Current Cycle Current Cycle Description Approved) Value Minimum Core Rated Core Flow Mininimum Core Rated Core Flow+/- F (%) Flow Limiting Case Limiting Case Flow Limiting Case. Limiting Case Contribution to Bundle Power Due to ., [[ 3:"1 E[ 1I [[Failed TIP Contribution to Bundle .Power Due to [[ ] [[ [[ Er 1I E[ Failed LPRM Total Uncertainty in Calculated Bundle [[I II [ II E[ .-Power Uncertainty of TIP Signal Nodal E[ H.]] [[ II [[ I1 E[ : 1 EE Uncertainty Table 5. Power Distribution Uncertainties Paue 23 ot'24 GNF NON-PROPRIETARY INFORMATION Class I GNF Attachment Table 6. Critical Power Uncertainties Nominal Value Previous Cycle Previous Cycle Current Cycle Current Cycle Descri ption Minimum Core Rated Core Flow Minimum Core Rated Core Flow+/- a Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case Table 6. Critical Power Uncertainties Page 24 ol'24}}