RS-05-012, Request for Amendment to Technical Specification Section 5.6.5, Core Operating Limits Report (Cola).

Request for Amendment to Technical Specification Section 5.6.5, Core Operating Limits Report (Cola).

ML050680274
Person / Time
Site:	LaSalle
Issue date:	03/07/2005
From:	Jury K Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-05-012
Download: ML050680274 (12)

Text

Exelon Generation www.exeloncorp .com 4300 Winfield Road Warrenville, IL 60555 RS-05-012 10 CFR 50.90 March 7, 2005 U . S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C . 20555 LaSalle County Station, Units 1 and 2 Facility Operating License Nos. NPF-1 1 and NPF-1 8 NRC Docket Nos. 50-373 and 50-374

Subject:

Request for Amendment to Technical Specifications Section 5.6.5, "Core Operating Limits Report (COLA)"

References:

(1) Letter from S. A. Richards (USNRC) to J. A. Mallay (Framatome),

"Acceptance for Referencing Power Corporation Topical Report EMF-2245(P), Revision 0, `Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel (TAC No. MA6438),"' dated August 30, 2000 (2) Letter from S. A. Richards (USNRC) to J . A. Mallay (Framatome),

"Acceptance for Referencing of Licensing Topical report EMF-2361 (P),

Revision VEXEM BWR-2000 ECCS Evaluation Model (TAC No.

MB0574),"' dated May 29, 2001 In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC, (EGC), requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos . NPF-1 1 and NPF-18 . Specifically, the proposed changes will add two NRC approved topical report references to the list of analytical methods in TS 5.6 .5, "Core Operating Limits Report (COLA)," that can be used to determine core operating limits.

The proposed changes are:

1. Add a NRC previously approved Siemens Power Corporation (SPC) topical report reference for determination of fuel assembly critical power for previously loaded Global Nuclear Fuel (GNF) GE14 fuel which will be co-resident with reload Framatome ANP ATRIUM-10 fuel .

2. Add a NRC previously approved Framatome Advanced Nuclear Power, Inc . (FAA-ANP) topical report reference for an updated methodology for evaluation of loss of coolant accident (LOCA) conditions .

The proposed changes are the result of a decision to utilize Framatome ANP ATRIUM-10 fuel during the Unit 1 Refueling Outage 11 currently scheduled for February 2006.

March 7, 2005 U. S. Nuclear Regulatory Commission Page 2 The ATRIUM-10 fuel is now manufactured and licensed by Framatome ANP, a subsidiary of the AREVA group. This organization has previously been known as Siemens Power Corporation (SPC) and subsequently as Framatome Advanced Nuclear Power (FRA-ANP).

The first added reference, "EMF-2245(P)(A) Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," will list an SPC method for determining the critical power of co-resident fuel when a SPC (Framatome ANP) fuel design is reintroduced into a reloaded core. This topical report has been previously reviewed and approved by the NRC for use by licensees in Reference 1 .

The second added reference, "EMF-2361 (P)(A) EXEM BWR-2000 ECCS Evaluation Model, will list a FRA-ANP updated methodology for evaluation of LOCA conditions. This topical report has also been previously reviewed and approved by the NRC for use by licensees in Reference 2.

The proposed changes have been reviewed by the LSCS Plant Operations Review Committee and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program .

In accordance with 10 CFR 50.91, "Notice for public comment ; State consultation," EGC is notifying the State of Illinois of this application for changes to the TS by transmitting a copy of this letter and its attachments to the designated State Official .

Should you have any questions or require additional information, please contact Ms. Alison Mackellar at (630) 657-2817 .

I declare under penalty of perjury that the foregoing is true and correct .

X Executed on I U_ 7, WX Keith R. Jury Director - Licensing and Regulatory Affairs Exelon Generation Company, LLC  : Evaluation of Proposed Changes : Mark-up of Proposed Technical Specifications Page Changes : Typed Pages for Technical Specifications Changes

ATTACHMENT I EVALUATION OF PROPOSED CHANGES INDEX 1 .0 DESCRIPTION

2.0 PROPOSED CHANGE

S

5.0 REGULATORY ANALYSIS

5.1 No Significant Hazards Consideration 5.2 Applicable Regulatory Requirements / Criteria 6 .0 ENVIRONMENTAL EVALUATION

7.0 REFERENCES

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES 1 .0 DESCRIPTION In accordance with 10 CFR 50.90, Exelon Generation Company, LLC, (EGC), requests the following amendment to Appendix A, Technical Specifications (TS), of Facility Operating License Nos. NPF-1 1 and NPF-1 8. Specifically, the proposed changes will add two NRC approved topical report references to the list of analytical methods in TS 5.6 .5, "Core Operating Limits Report (COLR)," that may be used to determine core operating limits .

The proposed changes are:

1 . Add a NRC previously approved Siemens Power Corporation (SPC) topical report reference for determination of fuel assembly critical power for previously loaded Global Nuclear Fuel (GNF) GE14 fuel which will be co-resident with reload Framatome ANP ATRIUM-10 fuel.

2 . Add a NRC previously approved Framatome Advanced Nuclear Power, Inc .

(FRA-ANP) topical report reference for an updated methodology for evaluation of loss of coolant accident (LOCA) conditions .

The proposed changes are the result of a decision to reinsert Framatome ANP ATRIUM-10 fuel during the Unit 1 Refueling Outage 11, currently scheduled for February 2006.

The ATRIUM-10 fuel is now manufactured and licensed by Framatome ANP (FRA-ANP),

a subsidiary of the AREVA group. This organization has previously been known as Siemens Power Corporation (SPC) and subsequently as Framatome Advanced Nuclear Power (FRA-ANP). The Unit 1 Cycle 12 core reload will be analyzed by Framatome ANP.

2.0 PROPOSED CHANGE

S The first added reference, "EMF-2245(P)(A) Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel," will list an SPC method for determining the critical power of co-resident fuel when an SPC (Framatome ANP) fuel design is reintroduced into a reloaded core. This topical report has been previously reviewed and approved by the NRC for use by licensees in Reference 1 .

The second added reference, "EMF-2361 (P)(A) EXEM BWR-2000 ECCS Evaluation Model," will list a FRA-ANP updated methodology for evaluation of LOCA conditions .

This topical report has also been previously reviewed and approved by the NRC for use by licensees in Reference 2 .

The marked-up and retyped TS pages are contained in Attachments 2 and 3.

Page 2 of 6

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES 3 .0 BACKGROUND The analytical methods currently listed in TS 5 .6.5 support the determination of core operating limits for both units by using GNF or FRA-ANP methodology. LaSalle County Station, (LSCS), Unit 1 currently uses a mixture of ATRIUM-913, ATRIUM-10 and GE14 fuel in the core . The determination of fuel assembly critical power for ATRIUM-913 and ATRIUM-10 fuel for the current operating cycle is determined with a GNF critical power correlation and overall core operating limits were determined using GNF methodology.

currently LSCS Unit 2 uses a mixture of ATRIUM-913 and ATRIUM-10. The overall core operating limits are currently determined using FRA-ANP methodology. During the upcoming Unit 2 Refueling Outage 10, scheduled for February 2005, LSCS Unit 2 will also load a mixture of ATRIUM-913, ATRIUM-10 and GE14 fuel into the core.

During this Unit 2 Operating Cycle 11, the determination of fuel assembly critical power for ATRIUM-913, ATRIUM-10 and GE1 4 fuel will be determined with a GNF critical power correlation and overall core operating limits will be determined using GNF methodology .

EGC has recently decided to load Framatome ANP ATRIUM-10 fuel during the subsequent Unit 1 Refueling Outage 11 currently scheduled for February 2006 . LSCS intends to use the most recent FRA-ANP methodologies to determine overall core operating limits for future core configurations. This change will require the listing of additional analytical methodologies for evaluating LOCH conditions and analyzing the critical power performance of the GE14 fuel with the FRA-ANP methodology. Thus, the proposed changes will allow LSCS to use FRA-ANP's most recent LOCA analytical methods W ATRIUM-10 fuel and SPC critical power correlations to determine the critical power for the co-resident GE14 fuel .

4.0 TECHNICAL ANALYSIS

TS Section 5.6 .5, requires that a COLA be established and that the analytical methods used to determine the core operating limits be those previously reviewed and approved by the NRC . The approved analytical methods are listed in TS Section 5.6.5.b. The analytical methods listed in this section support operation of certain types of fuel contained in the reactor core and list the analytical codes used to calculate operating parameters . The analytical codes are utilized to predict core behavior under normal and accident conditions . The proposed additional codes are discussed below.

EMF-2245(P)(A) Topical Report The NRC, in Reference 1, approved topical report EMF-2245(P), Revision 0, "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel ." This topical report presented two processes for the application of an approved SPC critical power correlation to pre-existing co-resident fuel, when an SPC (Framatome ANP) fuel design is reintroduced into a reload core.

LSCS has determined that the use of this SPC topical report and the processes for the application of the critical power correlation to pre-existing co-resident fuel Page 3 of 6

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES as described in Reference 1 are appropriate for LSCS Units 1 and 2, and provides an equivalent level of protection as that currently provided .

EMF-2361 (P)(A) Topical Report The NRC, in Reference 2, approved topical report EMF-2361 (P), Revision 0, "EXEM BWR-2000 ECCS Evaluation Model ." This topical report submitted by FRA-ANP describes revisions made to the methodology for evaluation of LOCA conditions. LSCS has determined that the use of FRA-ANP EXEM BWR-2000 ECCS Evaluation Model contained in this topical report as described in Reference 2 is appropriate for LSCS Units 1 and 2, and provides an equivalent level of protection as that currently provided.

5.0 REGULATORY ANALYSIS

5 .1 No Significant Hazards Consideration EGC has evaluated the proposed changes to the TS for LSCS, Unit 1 and Unit 2, and has determined that the proposed changes do not involve a significant hazards consideration and is providing the following supporting information.

Does the change involve a significant increase in the probability or consequences of an accident previously evaluated?

Response : No The proposed changes will add two additional NRC approved topical report references to the list of administratively controlled analytical methods in TS 5.6 .5, "Core Operating Limits Report (COLR)," that can be used to determine core operating limits . TS 5.6.5 lists NRC approved analytical methods used at LaSalle County Station (LSCS) to determine core operating limits .

LSCS Unit 1 is scheduled to reload Framatome ANP ATRIUM-10 fuel during the Unit 1 Refueling Outage 11 currently scheduled for February 2006 . The proposed changes to TS Section 5.6 .5 will add FRA-ANP methodologies to determine overall core operating limits for future core configurations . This change will require the listing of additional analytical methods for evaluating LOCA conditions and determining the critical power performance of the GE14 fuel . Thus, the proposed changes will allow LSCS to use the most recent FRA-ANP LOCA methodology for evaluation of ATRIUM-10 fuel and SPC critical power correlations to determine the critical power for the GE1 4 fuel .

The addition of approved methods to TS Section 5.6.5 has no effect on any accident initiator or precursor previously evaluated and does not change the manner in which the core is operated . The methods have been reviewed to ensure that the output accurately models predicted core behavior, have no effect on the type or amount of radiation released, and have no effect on predicted Page 4 of 6

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES offsite doses in the event of an accident. Additionally the methods do not change any key core parameters that influence any accident consequences. Thus, the proposed changes do not have any effect on the probability of an accident previously evaluated .

The methodology conservatively establishes acceptable core operating limits such that the consequences of previously analyzed events are not significantly increased.

The proposed changes in the administratively controlled analytical methods do not affect the ability of LSCS to successfully respond to previously evaluated accidents and does not affect radiological assumptions used in the evaluations.

Thus, the radiological consequences of any accident previously evaluated are not increased .

Therefore, the proposed changes do not involve a significant increase in the probability or consequences of an accident previously evaluated.

Does the change create the possibility of a new or different kind of accident from any accident previously evaluated?

Response : No The proposed changes to TS Section 5.6.5 do not affect the performance of any LSCS structure, system, or component credited with mitigating any accident previously evaluated . The insertion of fuel, which has been analyzed with NRC approved methodologies, will not affect the control parameters governing unit operation or the response of plant equipment to transient conditions . The proposed changes do not introduce any new modes of system operation or failure mechanisms.

Therefore, the proposed changes do not create the possibility of a new or different kind of accident from any previously evaluated.

Does the change involve a significant reduction in a margin of safety?

Response : No The proposed changes will add two additional references to the list of administratively controlled analytical methods in TS 5 .6.5 that can be used to determine core operating limits . The proposed changes do not modify the safety limits or setpoints at which protective actions are initiated and do not change the requirements governing operation or availability of salty equipment assumed to operate to preserve the margin of safety . Therefore, LSCS has determined that the proposed changes provide an equivalent level of protection as that currently provided.

Page 5 of 6

ATTACHMENT 1 EVALUATION OF PROPOSED CHANGES Therefore, the proposed changes do not involve a significant reduction in a margin of safety .

Based on the above information, EGC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c), and, accordingly, a finding of "no significant hazards consideration" is justified .

5 .2 Applicable Regulatory Requirements / Criteria TS 5.6.5 lists NRC approved analytical methods used at LSCS to determine core operating limits . The list of NRC approved analytical methods to be used to determine core operating limits provides the necessary administrative controls to assure operation of the facility in a safe manner and thus in accordance with 10 CFR 50.36, "Technical specifications," paragraph (c)(5), the additional proposed analytical methods must be included in the TS .

6 .0 ENVIRONMENTAL EVALUATION A review has determined that the proposed amendment would not change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation," and would not change an inspection or surveillance requirement. The proposed amendment does not involve 0) a significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure.

Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51 .22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible 110 categorical exclusion or otherwise not requiring environmental review," paragraph (c)(9) .

7.0 REFERENCES

1 . S. A. Richards (USNRC) to J. A. Mallay (Framatome), "Acceptance for Referencing Power Corporation Topical Report EMF-2245(P), Revision 0, `Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel (TAC No.

MA6438),"' dated August 30, 2000

2. S. A. Richards (USNRC) to J . A. Mallay (Framatome), "Acceptance for Referencing of Licensing Topical report EMF-2361(P), Revision 0,'EXEM BWR-2000 ECCS Evaluation Model (TAC No. MB0574),"' dated May 29, 2001

ATTACHMENT 2 Mark-up of Proposed Technical Specifications Page Changes REVISED TS PAGE 5.6-4

Reporting Requirements 5 .6 5 .6 Reporting Requirements 5 .6 .5 CORE OPERATING LIMITS REPORT (COLR) (continued) 12 . ANF-91-048(P)(A), "ANF Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model ."

13 . EMF-2209(P)(A), "SPCB Critical Power Correlation ."

14 . ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs ."

15 . NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel ."

16 . NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods ."

17 . EMF-1125(P)(A), "ANFB Critical Power Correlation Application for Co-Resident Fuel ."

18 . ANF-1125(P)(A), "ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties ."

19 . EMF-85-74(P)(A), - RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model ."

20 . EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors : Evaluation and Validation of CASMO-4/MICROBURN-B2 ."

Correlation 21 . NEDC-32981P(A), "GEXL96 for Atrium-9B Fuel ."

22 . NEDC-33106P, "GEXL97 Correlation for Atrium-10 Fuel ."

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i .e ., report number, title, revision, date, and any supplements) .

(continued)

LaSalle 1 and 2 5 .6-4 Amendment No . 164/150

ATTACHMENT 3 Typed Pages for Technical Specifications Changes REVISED TS PAGE 5.6-4

Reporting Requirements 5 .6 5 .6 Reporting Requirements 5 .6 .5 CORE OPERATING LIMITS REPORT-11OLRI (continued) 12 . ANF-91-048(P)(A), "ANF Corporation Methodology for Boiling Water Reactors EXEM BWR Evaluation Model ."

13 . EMF-2209(P)(A), "SPCB Critical Power Correlation ."

14 . ANF-89-98(P)(A), "Generic Mechanical Design Criteria for BWR Fuel Designs ."

15 . NEDE-24011-P-A, "General Electric Standard Application for Reactor Fuel ."

16 . NFSR-0091, "Benchmark of CASMO/MICROBURN BWR Nuclear Design Methods ."

17 . EY-1125(P)(A), "ANFB Critical Power Correlation Application for Co-Resident Fuel ."

18 . ANF-1125(P)(A), "ANFB Critical Power Correlation Determination of ATRIUM-9B Additive Constant Uncertainties ."

19 . EMF-85-74(P)(A), "RODEX2A (BWR) Fuel Rod Thermal-Mechanical Evaluation Model ."

20 . EMF-2158(P)(A), "Siemens Power Corporation Methodology for Boiling Water Reactors : Evaluation and Validation of CASMO-4/MICROBURN-B2 ."

21 . NEDC-32981P(A), "GEXL96 Correlation for Atrium-9B Fuel ."

22 . NEDC-33106P, "GEXL97 Correlation for Atrium-10 Fuel ."

23 . EMF-2245(P)(A), "Application of Siemens Power Corporation's Critical Power Correlations to Co-Resident Fuel ."

24 . EMF-2361(P)(A), "EXEM BWR-2000 ECCS Evaluation Model ."

The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (i .e ., report number, title, revision, date, and any supplements) .

(continued)

LaSalle 1 and 2 5 .6-4 Amendment No . xxx/xxx