RS-11-149, Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit

Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit

ML112860068
Person / Time
Site:	LaSalle
Issue date:	10/12/2011
From:	Gullott D Exelon Generation Co, Exelon Nuclear
To:	Office of Nuclear Reactor Regulation, Document Control Desk
Shared Package
ML112860067	List: RS-11-149, Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit
References
RS-11-149 GNF-0000-0138-9711-R0-NP
Download: ML112860068 (65)

Text

Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road clear Warrenville, IL 60555 Attachments 5, 6, and 7 contain Proprietary Information.

Withhold from public disclosure under 10 CFR 2.390.

When separated from Attachments 5, 6, and 7, this document is decontrolled.

RS-1 1-149 10 CFR 50.90 October 12, 2011 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001 LaSalle County Station, Unit 1 Facility Operating License No. NPF-1 1 NRC Docket Nos. 50-373

Subject:

Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC) requests an amendment to Appendix A, Technical Specifications (TS), of Facility Operating License No. NPF-1 1 for LaSalle County Station (LSCS), Unit 1. The proposed change revises the LSCS, Unit 1 minimum critical power ratio safety limit (MCPR SL) in TS Section 2.1.1, "Reactor Core SLs."

This change is required to support the upcoming LSCS, Unit 1, Cycle 15 operation. Cycle 15 will be the first cycle of operation with a mixed core containing the following fuel types: Fresh Global Nuclear Fuel (GNF) GNF2 fuel, and reloaded AREVA ATRIUM- 10 fuel.

The proposed amendment reflects an increase of the two recirculation loop MCPR SL limit from >_ 1.11 to > 1.13 and an increase in the single recirculation loop MCPR SL from > 1.12 to

>_ 1.15. The proposed change has been evaluated in accordance with 10 CFR 50.91(a)(1) using the criteria in 10 CFR 50.92(c), and it has been determined that the change does not involve a significant hazards consideration. The bases for these determinations are included in the attached submittal. contains the evaluation of the proposed change. Attachments 2 and 3 provide the marked up TS page and the retyped TS page, respectively. (i.e., letter from C. F. Lamb (GNF) to J. Wheeler (EGC), dated September 30, 2011) specifies the new MCPR SLs for LSCS, Unit 1, Cycle 15. Attachments 5, 6, and 7 contain information proprietary to GNF. GNF requests that these documents be withheld from Attachments 5, 6, and 7 contain Proprietary Information.

Withhold from public disclosure under 10 CFR 2.390.

When separated from Attachments 5, 6, and 7, this document is decontrolled.

October 12, 2011 U. S. Nuclear Regulatory Commission Page 2 public disclosure in accordance with 10 CFR 2.390(b)(4). An affidavit attesting to the proprietary nature of this information is contained in Attachment 4. Attachments 8 and 9 are non-proprietary versions of Attachments 5 and 6, respectively. Attachment 7 is proprietary in its entirety; therefore, no non-proprietary version is provided.

The attached amendment request is subdivided as follows:

Attachment 1 provides an evaluation of the proposed change.

Attachment 2 provides the current TS page with the proposed change indicated with markups.

Attachment 3 provides the "Clean" TS page that includes the proposed change.

Attachment 4 provides an affidavit from GNF supporting the request to withhold proprietary information from public disclosure in accordance with 10 CFR 2.390 Attachment 5 provides GNF Additional Information regarding the Requested Changes to the Technical Specifications MCPR SL for LSCS, Unit 1, Cycle 15 - Proprietary Version Attachment 6 provides supplemental Peach Bottom MCPR SL license amendment request supplemental information applied to LSCS, Unit 1, Cycle 15 - Proprietary Version Attachment 7 provides the GNF Fuel Application Overview - Proprietary Attachment 8 provides GNF Additional Information regarding the Requested Changes to the Technical Specifications MCPR SL for LSCS, Unit 1, Cycle 15 - Non-Proprietary Version Attachment 9 provides supplemental Peach Bottom MCPR SL license amendment request supplemental information applied to LSCS, Unit 1, Cycle 15 - Non-Proprietary Version Attachment 10 provides a power-to-flow map for LSCS, Unit 1, Cycle 14 and Expected for Cycle 15 EGC requests approval of the proposed amendment by February 10, 2012, to support the upcoming refueling outage. Once approved, the amendment shall be implemented after Cycle 14 is completed and prior to the operation of Cycle 15.

The proposed amendment has 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.

EGC is notifying the State of Illinois of this application for a change to the TS by sending a copy of this letter and its attachments to the designated State Official in accordance with 10 CFR 50.91, "Notice for public comment; State consultation," paragraph (b).

The proposed change does not include any new commitments. Should you have any questions concerning this letter, please contact Mr. Mitchel Mathews at (630) 657-2819.

October 12, 2011 U. S. Nuclear Regulatory Commission Page 3 I declare under penalty of perjury that the foregoing is true and correct. Executed on the 12th day of October 2011.

Respectfully, David M. Gullott Manager - Licensing and Regulatory Affairs Attachments: 1. Evaluation of Proposed Change

2. Mark-up of Proposed Technical Specifications Page Changes

3. Clean Technical Specifications Page

4. Global Nuclear Fuel Affidavit Supporting Request to Withhold from Public Disclosure Under 10 CFR 2.390 5 Global Nuclear Fuel Letter Containing Additional Information Regarding the Requested Changes to the Technical Specification MCPR SL LaSalle County Station, Unit 1, Cycle 15 - Proprietary Version

6. Supplemental Peach Bottom RAI Responses Applied to LaSalle County Station, Unit 1, Cycle 15 - Proprietary Version

7. Global Nuclear Fuel - Fuel Application Overview - Proprietary Version

8. Global Nuclear Fuel Letter Containing Additional Information Regarding the Requested Changes to the Technical Specification MCPR SL LaSalle County Station, Unit 1, Cycle 15 -

Non-Proprietary Version

9. Supplemental Peach Bottom RAI Responses Applied to LaSalle County Station, Unit 1, Cycle 15 - Non-Proprietary Version

10. LaSalle County Station, Unit 1, Cycle 14 and Expected Cycle 15, Power-to-Flow Map cc: Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change

Subject:

Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedents 4.3 No Significant Hazards Consideration 4.4 Conclusions 5.0 ENVIRONMENTAL CONSIDERATON

6.0 REFERENCES

Page 1 of 7

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION The proposed change would revise the LaSalle County Station (LSCS), Unit 1 Technical Specifications (TS) Section 2.1.1, "Reactor Core Safety Limits." Specifically, the proposed change modifies the LSCS, Unit 1 Minimum Critical Power Ratio Safety Limits (MCPR SLs) for both two loop and single loop recirculation operation in TS Section 2.1.1.2. The change to TS Section 2.1.1.2 is necessary as a result of LSCS, Unit 1, Cycle 15 cycle-specific analyses.

The analyses performed to support the proposed cycle-specific MCPR SL changes utilized NRC-approved methodology.

The proposed change is described in detail in Section 2.0 below.

The requested approval date of February 14, 2012, will allow time for the LSCS, Unit 1 Core Operating Limits Report (COLR) to be prepared prior to LSCS, Unit 1, Cycle 15 operation.

2.0 DETAILED DESCRIPTION The proposed change involves revising the MCPR SLs contained in TS Section 2.1.1.2 for both two recirculation loop and single recirculation loop operation for LSCS, Unit 1. Analysis determined that LSCS, Unit 1 MCPR SL value for two recirculation loop operation requires revision from >_ 1.11 to > 1.13. Additionally, the LSCS, Unit 1 MCPR SL value for single recirculation loop operation requires revision from >_ 1.12 to >_ 1.15. The major contributors to these revisions are discussed in Attachment 5, Section 2.1, "Major Contributors to SLMCPR Change."

3.0 TECHNICAL EVALUATION

LSCS, Unit 1, Cycle 15 is the first cycle of operation to utilize Global Nuclear Fuel (GNF) GNF2 fuel. The current Cycle 15 core design consists of 296 fresh GNF2 bundles and 468 reloaded ATRIUM-10 bundles. While Exelon Generation Company, LLC (EGC) does not expect this to change, any final core design changes will be evaluated to confirm that the proposed Technical Specifications (TS) changes remain valid.

The criteria used in developing the LSCS, Unit 1, Cycle 15 core loading pattern is discussed in in section labeled "GNF Response to RAI Applied to LaSalle Unit 1." The design record file for the LSCS, Unit 1, Cycle 15 core loading pattern is included in , Figure 1, "Current Cycle Core Loading Diagram."

The proposed MCPR SL values for the upcoming operating Cycle 15 were developed with GNF's NRC-approved MCPR SL methodology. The methodology used is found in NEDE-2401 1 -P-A, "General Electric Standard Application for Reactor Fuel (GESTAR-1 I)," and NEDC-33106P, "GEXL97 Correlation for ATRIUM-10 Fuel." While Cycle 15 is the first cycle to introduce GNF2 fuel to the LSCS, Unit 1 core, all fuel types that were utilized in the design of the Cycle 15 core are fully compatible. Moreover, identical NRC-approved methodology has been used to analyze similar core designs as discussed in the referenced letters in Section 4.2 Page 2 of 7

ATTACHMENT 1 Evaluation of Proposed Change below. A discussion justifying that the results are conservative related to the methodology deviations, penalties, and/or uncertainties is provided in the Monte Carlo two loop operation and single loop operation results listed in Attachment 5, Table 3. Attachment 5, Section 2.2 provides a discussion of deviations from the NRC-approved values. There are no 10 CFR Part 21 issues associated with the LSCS, Unit 1, Cycle 15 analysis.

EGC is proposing that the LSCS, Unit 1 Operating License be amended to modify the MCPR SLs reported in Technical Specification 2.1.1.2. The proposed change is necessary in order to reflect the safety limit changes for the Cycle 15 core.

The MCPR SL is developed to assure compliance with General Design Criterion 10 of 10 CFR 50, Appendix A. The Bases to TS Section 2.1.1.2 states that "The MCPR SL ensures sufficient conservatism in the operating MCPR limit that, in the event of an AOO [Anticipated Operational Occurrence] from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition." Attachment 10 provides an updated version of the power-to-flow map for LSCS, Unit 1, Cycle 14 and expected Cycle 15 operation including the stability Option III features of scram region and controlled entry region for back-up stability protection.

Information to support the cycle specific MCPR SL is included in Attachment 5. This attachment summarizes the MCPR SL analysis: methodology, inputs, results, and the reasons for the increase in the MCPR SL. The LSCS, Unit 1, Cycle 15 core will consist of 296 bundles of fresh GNF2, 320 bundles of once burned ATRIUM-10, and 148 bundles of twice burned ATRIUM-10 fuel. The COLR references in TS section 5.6.5 reflect the methods and codes that apply to fuel types in the Cycle 15 core.

For two-loop operation, a MCPR SL of >_ 1.13 was demonstrated to be adequate to ensure that 99.9% of the rods in the core avoid boiling transition during the most limiting AOO. For single -

loop operation, this assurance is provided by a MCPR SL of >_ 1.15. Attachment 6, provides a discussion and indicates the fuel bundle groups, group exposure, number of bundles, fuel type, and percent contribution to the number of fuel rods that are subjected to boiling transition that depicts the 0.1 % of fuel bundles that may experience boiling transition for the limit MCPR SL case in the section entitled, "GNF Response to RAI Applied to LaSalle Unit 1."

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria The proposed change has been evaluated to determine whether applicable regulations and requirements continue to be met.

10 CFR 50.36, "Technical specifications," defines a safety limit as a limit upon important process variables that are found to be necessary to reasonably protect the integrity of certain physical barriers that guard against the uncontrolled release of radioactivity.

Page 3 of 7

ATTACHMENT 1 Evaluation of Proposed Change 4.2 Precedents

1. Letter from C. F. Lyon (U. S. NRC) to Vice President Entergy Operations, Inc., "Grand Gulf Nuclear Station, Unit 1 - Issuance of Amendment Re: Change to Minimum Critical Power Ratio Safety Limit (TAC No. ME2474)," dated March 25, 2010

2. Letter from A. B. Wang (U. S. NRC) to Vice President Entergy Operations, Inc., "River Bend Station, Unit 1 - Issuance of Amendment Re: Changes to Technical Specification 5.6.5, 'Core Operating Limits Report (COLR)' (TAC No. ME0157)," dated September 29, 2009

3. Letter from C. F. Lyon (U. S. NRC) to J. V. Parrish (Energy Northwest), "Columbia Generating Station - Issuance of Amendment Re: Core Operating Limits Report and Scram Time Testing (TAC No. MD9247)," dated May 5, 2009 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, "Application for amendment of license or construction permit," Exelon Generation Company, LLC (EGC) is requesting a change to the Technical Specifications (TS) of Facility Operating License Nos. NPF-1 1 for LaSalle County Station (LSCS), Unit 1.

According to 10 CFR 50.92, "Issuance of amendment," paragraph (c), a proposed amendment to an operating license involves no significant hazards consideration if operation of the facility in accordance with the proposed amendment would not:

(1) Involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Involve a significant reduction in a margin of safety.

In support of this determination, an evaluation of each of the three criteria set forth in 10 CFR 50. 92 is provided below:

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

Response: No.

The Minimum Critical Power Ratio Safety Limit (MCPR SL) is defined in the TS Bases Section B 2.1.1 as that limit "that, in the event of an AOO [Anticipated Operational Occurrence] from the limiting condition of operation, at least 99.9% of the fuel rods in the core would be expected to avoid boiling transition." The MCPR SL satisfies the requirements of General Design Criterion 10 of Appendix A to 10 CFR 50 regarding Page 4 of 7

ATTACHMENT 1 Evaluation of Proposed Change acceptable fuel design limits. The MCPR SL is reevaluated for each reload using NRC-approved methodologies. The analyses for LSCS, Unit 1, Cycle 15 have concluded that a two-loop MCPR SL of ? 1.13, based on the application of Global Nuclear Fuel's (GNF's)

NRC-approved MCPR SL methodology, will ensure that this acceptance criterion is met.

For single-loop operation, a MCPR SL of >_ 1.15 also ensures that this acceptance criterion is met. The MCPR operating limits are presented and controlled in accordance with the LSCS, Unit 1 Core Operating Limits Report (COLR).

The requested Technical Specification changes do not involve any plant modifications or operational changes that could affect system reliability or performance or that could affect the probability of operator error. The requested changes do not affect any postulated accident precursors, do not affect any accident mitigating systems, and do not introduce any new accident initiation mechanisms.

Therefore, the changes to the Minimum Critical Power Ratio safety limit do not involve a significant increase in the probability or consequences of any accident previously evaluated.

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

Response: No.

The GNF2 fuel to be used in Cycle 15 is of a design compatible with the co-resident Areva ATRIUM-10 fuel. Therefore, the introduction of GNF2 fuel into the Cycle 15 core will not create the possibility of a new or different kind of accident. The proposed change does not involve any new modes of operation, any changes to setpoints, or any plant modifications.

The proposed revised MCPR SLs have accounted for the mixed fuel core and have been shown to be acceptable for Cycle 15 operation. Compliance with the criterion for incipient boiling transition continues to be ensured. The core operating limits will continue to be developed using NRC approved methods which also account for the mixed fuel core design.

The proposed MCPR SLs or methods for establishing the core operating limits do not result in the creation of any new precursors to an accident.

Therefore, this change does not create the possibility of a new or different kind of accident from any previously evaluated.

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

Response: No.

The MCPR SLs have been evaluated in accordance with GNF's NRC-approved cycle-specific limit methodology to ensure that during normal operation and during AOO's at least 99.9% of the fuel rods in the core are not expected to experience transition boiling. The proposed revised MCPR SLs have accounted for the mixed fuel core and have been shown to be acceptable for Cycle 15 operation. Compliance with the criterion for incipient boiling Page 5 of 7

ATTACHMENT 1 Evaluation of Proposed Change transition continues to be ensured. On this basis, the implementation of the change to the MCPR SLs does not involve a significant reduction in a margin of safety.

Therefore, the proposed change does not involve a significant reduction in a margin of safety.

Based on the above, EGC has determined that operation of the facility in accordance with the proposed change does not involve a significant hazards consideration as defined in 10 CFR 50.92(c), in that it:

(1) Does not involve a significant increase in the probability or consequences of an accident previously evaluated; or (2) Does not create the possibility of a new or different kind of accident from any accident previously evaluated; or (3) Does not involve a significant reduction in a margin of safety.

Based on the above evaluation, Exelon Generation Company, LLC concludes that the proposed amendment presents no significant hazards consideration under the standards set forth in 10 CFR 50.92(c).

4.4 Conclusions In conclusion, based on the considerations discussed above, (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

5.0 ENVIRONMENTAL CONSIDERATION

EGC has evaluated this proposed operating license amendment consistent with the criteria for identification of licensing and regulatory actions requiring environmental assessment in accordance with 10 CFR 51.21, "Criteria for and identification of licensing and regulatory actions requiring environmental assessments." EGC has determined that this proposed change meets the criteria for a categorical exclusion set forth in paragraph (c)(9) of 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review," and as such, has determined that no irreversible consequences exist in accordance with paragraph (b) of 10 CFR 50.92, "Issuance of amendment." This determination is based on the fact that this change is being proposed as an amendment to the license issued pursuant to 10 CFR 50, "Domestic Licensing of Production and Utilization Facilities," which changes 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," or which changes an inspection or a surveillance requirement, and the amendment meets the following specific criteria:

Page 6 of 7

ATTACHMENT 1 Evaluation of Proposed Change (i) The amendment involves no significant hazards consideration.

As demonstrated in Section 5.1, "No Significant Hazards Consideration," the proposed change does not involve any significant hazards consideration.

(ii) There is no significant change in the types or significant increase in the amounts of any effluent that may be released offsite.

The proposed change does not result in an increase in power level, does not increase the production nor alter the flow path or method of disposal of radioactive waste or byproducts. It is expected that all plant equipment would operate as designed in the event of an accident to minimize the potential for any leakage of radioactive effluents; thus, there will be no change in the amounts of radiological effluents released offsite.

Based on the above evaluation, the proposed change will not result in a significant change in the types or significant increase in the amounts of any effluent released offsite.

(iii) There is no significant increase in individual or cumulative occupational radiation exposure.

There is no net increase in individual or cumulative occupational radiation exposure due to the proposed change. The proposed action will not change the level of controls or methodology used for processing of radioactive effluents or handling of solid radioactive waste, nor will the proposed action result in any change in the normal radiation levels within the plant.

Based on the above information, there will be no increase in individual or cumulative occupational radiation exposure resulting from this change.

6.0 REFERENCES

1. NEDE-33106P, "GEXL97 Correlation Report for ATRIUM-10 Fuel"

2. NEDE-2401 1 -P-A, "GESTAR-II, General Electric Standard Application for Reactor Fuel" Page 7of7

ATTACHMENT 2 LASALLE COUNTY STATION UNIT 1 Docket No. 50-373 License No. NPF-1 1 Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Markup of Proposed Technical Specifications Page Changes MARKED UP TS PAGE 2.0-1

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs For Unit 1, MCPR shall be >_1.13 for 2.1.1.1 With the reactor steam dome pressure < 785 psig or core two recirculation flow < 10% rated core flow:

loop operation or

>1.15 for single THERMAL POWER shall be <_ 25% RTP.

recirculation loop operation. 2.1.1.2 With the reactor steam dome pressure >_ 785 psig and core flow >_ 10% rated core flow:

For Unit 2, MCPR shall be >_1.11 for -

two recirculation trop operation or >_1.12 for single 2.1.1.3 Reactor vessel water level shall be greater than the top recirculation loop of active irradiated fuel.

operation.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be <_ 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

LaSalle 1 and 2 2.0-1 Amendment No. 44444-34

ATTACHMENT 3 LASALLE COUNTY STATION UNIT 1 Docket No. 50-373 License No. NPF-1 1 Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Clean TS Page REVISED TS PAGE 2.0-1

SLs 2.0 2.0 SAFETY LIMITS (SLs) 2.1 SLs 2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam dome pressure < 785 psig or core flow < 10% rated core flow:

THERMAL POWER shall be <_ 25% RTP.

2.1.1.2 With the reactor steam dome pressure >_ 785 psig and core flow >_ 10% rated core flow:

For Unit 1, MCPR shall be >_ 1.13 for two recirculation loop operation or >_ 1.15 for single recirculation loop operation.

For Unit 2, MCPR shall be >_ 1.11 for two recirculation loop operation or >_ 1.12 for single recirculation loop operation.

2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.

2.1.2 Reactor Coolant System Pressure SL Reactor steam dome pressure shall be <_ 1325 psig.

2.2 SL Violations With any SL violation, the following actions shall be completed within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:

2.2.1 Restore compliance with all SLs; and 2.2.2 Insert all insertable control rods.

LaSalle 1 and 2 2.0-1 Amendment No.

ATTACHMENT 4 LASALLE COUNTY STATION UNIT 1 Docket No. 50-373 License No. NPF-1 1 Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Global Nuclear Fuel Affidavit Supporting Request to Withhold from Public Disclosure Under 10 CFR 2.390

ENCLOSURE6 CFA-EXN-HA1-11-120 Affidavit

Global Nuclear Fuel - Americas AFFIDAVIT I, Atul A. Karve, state as follows:

(1) I am Engineering Manager, Methods, Global Nuclear Fuel - Americas, LLC (GNF-A), and have been delegated the function of reviewing the information described in paragraph (2) which is sought to be withheld, and have been authorized to apply for its withholding.

(2) The information sought to be withheld is contained in Enclosures 1, 3, and 5 of GNF's letter, CFL-EXN-HA1-11-120, C. Lamb (GNF-A) to J. Wheeler (Exelon Generation),

entitled "GNF Additional Information for SLMCPR Technical Specification Submittal Letter for LaSalle Unit 1 Cycle 15," dated September 28, 2011. GNF-A proprietary information in Enclosure 1, which is entitled "GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR, LaSalle Unit 1 Cycle 15," and Enclosure 3, which is entitled "Supplemental Peach Bottom RAI Responses Applied to LaSalle Unit 1 Cycle 15," is identified by a dotted underline inside double square brackets.

((This sentence is an example. {3})) A "((" marking at the beginning of a table, figure, or paragraph closed with a "))" marking at the end of the table, figure or paragraph is used to indicate that the entire content between the double brackets is proprietary. The information in Enclosure 5, which is entitled "Fuel Application Overview," is proprietary in its entirety.

The header of each page in this enclosure carries the notation "GNF Proprietary Information

- Class III (Confidential){3} " In each case, the superscript notation {3} refers to Paragraph (3) of this affidavit, which provides the basis for the proprietary determination.

(3) In making this application for withholding of proprietary information of which it is the owner or licensee, GNF-A relies upon the exemption from disclosure set forth in the Freedom of Information Act ("FOIA"), 5 USC Sec. 552(b)(4), and the Trade Secrets Act, 18 USC Sec. 1905, and NRC regulations 10 CFR 9.17(a)(4), and 2.390(a)(4) for "trade secrets" (Exemption 4). The material for which exemption from disclosure is here sought also qualify under the narrower definition of "trade secret", within the meanings assigned to those terms for purposes of FOIA Exemption 4 in, respectively, Critical Mass Energy Project v. Nuclear Regulatory Commission, 975 F2d 871 (DC Cir. 1992), and Public Citizen Health Research Group v. FDA, 704 F2d 1280 (DC Cir. 1983).

(4) Some examples of categories of information which fit into the definition of proprietary information are:

a. Information that discloses a process, method, or apparatus, including supporting data and analyses, where prevention of its use by GNF-A's competitors without license from GNF-A constitutes a competitive economic advantage over other companies;

b. Information which, if used by a competitor, would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing of a similar product;

c. Information which reveals aspects of past, present, or future GNF-A customer-funded development plans and programs, resulting in potential products to GNF-A; CFL-EXN-HA 1-11-120 Enclosures 1, 3, and 5 Affidavit Page 1 of 3

d. Information which discloses patentable subject matter for which it may be desirable to obtain patent protection.

The information sought to be withheld is considered to be proprietary for the reasons set forth in paragraphs (4)a. and (4)b. above.

(5) To address 10 CFR 2.390 (b) (4), the information sought to be withheld is being submitted to NRC in confidence. The information is of a sort customarily held in confidence by GNF-A, and is in fact so held. The information sought to be withheld has, to the best of my knowledge and belief, consistently been held in confidence by GNF-A, no public disclosure has been made, and it is not available in public sources. All disclosures to third parties including any required transmittals to NRC, have been made, or must be made, pursuant to regulatory provisions or proprietary agreements which provide for maintenance of the information in confidence. Its initial designation as proprietary information, and the subsequent steps taken to prevent its unauthorized disclosure, are as set forth in paragraphs (6) and (7) following.

(6) Initial approval of proprietary treatment of a document is made by the manager of the originating component, the person most likely to be acquainted with the value and sensitivity of the information in relation to industry knowledge, or subject to the terms under which it was licensed to GNF-A. Access to such documents within GNF-A is limited on a "need to know" basis.

(7) The procedure for approval of external release of such a document typically requires review by the staff manager, project manager, principal scientist or other equivalent authority, by the manager of the cognizant marketing function (or his delegate), and by the Legal Operation, for technical content, competitive effect, and determination of the accuracy of the proprietary designation. Disclosures outside GNF-A are limited to regulatory bodies, customers, and potential customers, and their agents, suppliers, and licensees, and others with a legitimate need for the information, and then only in accordance with appropriate regulatory provisions or proprietary agreements.

(8) The information identified in paragraph (2) is classified as proprietary because it contains details of GNF-A' s fuel design and licensing methodology. The development of this methodology, along with the testing, development and approval was achieved at a significant cost to GNF-A.

The development of the fuel design and licensing methodology along with the interpretation and application of the analytical results is derived from an extensive experience database that constitutes a major GNF -A asset.

(9) Public disclosure of the information sought to be withheld is likely to cause substantial harm to GNF-A's competitive position and foreclose or reduce the availability of profit-making opportunities. The information is part of GNF-A's comprehensive BWR safety and technology base, and its commercial value extends beyond the original development cost.

The value of the technology base goes beyond the extensive physical database and analytical methodology and includes development of the expertise to determine and apply the appropriate evaluation process. In addition, the technology base includes the value derived from providing analyses done with NRC-approved methods.

CFL-EXN-HA 1-11-120 Enclosures 1, 3, and 5 Affidavit Page 2 of 3

The research, development, engineering, analytical, and NRC review costs comprise a substantial investment of time and money by GNF-A.

The precise value of the expertise to devise an evaluation process and apply the correct analytical methodology is difficult to quantify, but it clearly is substantial.

GNF-A's competitive advantage will be lost if its competitors are able to use the results of the GNF-A experience to normalize or verify their own process or if they are able to claim an equivalent understanding by demonstrating that they can arrive at the same or similar conclusions.

The value of this information to GNF-A would be lost if the information were disclosed to the public. Making such information available to competitors without their having been required to undertake a similar expenditure of resources would unfairly provide competitors with a windfall, and deprive GNF-A of the opportunity to exercise its competitive advantage to seek an adequate return on its large investment in developing and obtaining these very valuable analytical tools.

I declare under penalty of perjury that the foregoing affidavit and the matters stated therein are true and correct to the best of my knowledge, information, and belief.

Executed on this 28th day of September 2011.

Atul A. Karve Engineering Manager, Methods Global Nuclear Fuel - Americas, LLC CFL-EXN-HA 1-11-120 Enclosures 1, 3, and 5 Affidavit Page 3 of 3

ATTACHMENT 8 LASALLE COUNTY STATION UNIT 1 Docket No. 50-373 License No. NPF-1 1 Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Global Nuclear Fuel Letter Containing Additional Information Regarding the Requested Changes to the Technical Specification MCPR SL LaSalle County Station, Unit 1, Cycle 15 -

Non-Proprietary Version

ENCLOSURE2 CFL-EXN-HA1-11-120 GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR, LaSalle Unit 1 Cycle 15 Non-Proprietary Information - Class I (Public)

INFORMATION NOTICE This is a non -proprietary version of CFL-EXN-HA1 120 Enclosure 1, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside an open and closed bracket as shown here ((

)).

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September 19, 2011 GNF-0000-01 38-971 1-R0-NP eDRFSection: 0000-0138-9711 RO GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR LaSalle Unit 1 Cycle 15 Copyright 2011 Global Nuclear Fuel - Americas, LLC All Rights Reserved LaSalle Unit 1 Cycle 15 Page 1 of 26

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Information Notice This is a non-proprietary version of the document GNF-0000-0138-9711-RO-P, which has the proprietary information removed. Portions of the document that have been removed are indicated by an open and closed bracket as shown here (( )).

Important Notice Regarding Contents of this Report Please Read Carefully The design, engineering, and other information contained in this document is furnished for the purpose of providing information regarding the requested changes to the Technical Specification SLMCPR for Exelon LaSalle Unit 1. The only undertakings of Global Nuclear Fuel - Americas, LLC (GNF-A) with respect to information in this document are contained in contracts between GNF-A and Exelon, and nothing contained in this document shall be construed as changing the contract. The use of this information by anyone other than Exelon, or for any purposes other than that for which it is intended is not authorized; and with respect to any unauthorized use, GNF-A makes no representation or warranty, express or implied, and assumes no liability as to the completeness, accuracy, or usefulness of the information contained in this document.

Page 2 of 26

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Table of Contents 1.0 METHODOLOGY .......................................................................................................................................... 4 2.0 DISCUSSION ................................................................................................................................................... 4 2.1. MAJOR CONTRIBUTORS TO SLMCPR CHANGE ..............................................................................................4 2.2. DEVIATIONS IN NRC-APPROVED UNCERTAINTIES .........................................................................................5 2.2.1. R-Factor ................................................................................................................................................. 5 2.2.2. Core Flow Rate and Random Effective TIP Reading ............................................................................. 6 2.3. DEPARTURE FROM NRC-APPROVED METHODOLOGY ....................................................................................7 2.4. FUEL AXIAL POWER SHAPE PENALTY ............................................................................................................7 2.5. METHODOLOGY RESTRICTIONS ...................................................................................................................... 8 2.6. MINIMUM CORE FLOW CONDITION ................................................................................................................8 2.7. LIMITING CONTROL ROD PATTERNS ..............................................................................................................9 2.8. CORE MONITORING SYSTEM ..........................................................................................................................9 2.9. POWER/FLOW MAP .........................................................................................................................................9 2.10. CORE LOADING DIAGRAM ..........................................................................................................................9 2.11. FIGURE REFERENCES ..................................................................................................................................9 2.12. ADDITIONAL SLMCPR LICENSING CONDITIONS ......................................................................................10 2.13. 10 CFR PART 21 EVALUATION .................................................................................................................10 2.14.

SUMMARY

................................................................................................................................................10

3.0 REFERENCES

..............................................................................................................................................11 List of Figures FIGURE 1. CURRENT CYCLE CORE LOADING DIAGRAM ...............................................................................................12 FIGURE 2. PREVIOUS CYCLE CORE LOADING DIAGRAM ....................... ....................................................................... 13 FIGURE 3. FIGURE 4.1 FROM NEDC-32601P -A ...........................................................................................................14 FIGURE 4. FIGURE 111.5-1 FROM NEDC-3260 1P -A ......................................................................................................15 FIGURE 5. RELATIONSHIP BETWEEN MIP AND CPR MARGIN ......................................................................................16 List of Tables TABLE 1. DESCRIPTION OF CORE .................................................................................................................................17 TABLE 2. SLMCPR CALCULATION METHODOLOGIES ................................................................................................. 18 TABLE 3. MONTE CARLO CALCULATED SLMCPR VS. ESTIMATE ...............................................................................19 TABLE 4. NON-POWER DISTRIBUTION UNCERTAINTIES ...............................................................................................21 TABLE 5. POWER DISTRIBUTION UNCERTAINTIES .......................................................................................................23 TABLE 6. CRITICAL POWER UNCERTAINTIES ...............................................................................................................25 Table of Contents Page 3 of 26

Non-Proprietary Information - Class I (Public) 1.0 Methodology Global Nuclear Fuel (GNF) performs Safety Limit Minimum Critical Power Ratio (SLMCPR) calculations in accordance with NEDE-24011-P-A "General Electric Standard Application for Reactor Fuel" (Revision 18) using the following Nuclear Regulatory Commission (NRC)-approved methodologies and uncertainties:

• NEDC-32601P-A, "Methodology 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 GE11, GE12 and GE13 Fuel,"

Revision 1, July 1999.

Table 2 identifies the actual methodologies used for the LaSalle Unit 1 Cycle 15 and the Cycle 15 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.

2.1. 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 Minimum Critical Power Ratio (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 effect 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 current cycle along with the TLO SLMCPR estimate using the MIPRIP correlation. If the 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 effects on the TLO SLMCPR due to methodology deviations, penalties, and/or uncertainty deviations from approved values. Based on the MIPRIP correlation and any effects due to deviations from approved values, a final estimated TLO SLMCPR is determined. Table 3 also provides the Methodology Page 4 of 26

Non-Proprietary Information - Class I (Public) actual calculated Monte Carlo SLMCPRs. Due to the increased uncertainty included in the Monte Carlo SLMCPR calculation due to the use of the GNF2 and ATRIUM-10 GEXL models, which have greater critical power uncertainties than GE 14 and earlier GNF fuel designs shown in Table 6, and the inherent assumption in the MIPRIP correlation of a normal distribution of bundle MCPRs in the core, while the Monte Carlo SLMCPR uses actual core MCPR distributions, the LaSalle Unit I Cycle 15 calculated Monte Carlo TLO SLMCPR using rated core power and rated core flow conditions is conservative compared to the corresponding estimated TLO SLMCPR value.

The intent of the final estimated TLO SLMCPR is to provide an estimate to check the reasonableness of the Monte Carlo result. It is not used for any other purpose. The methodology and final SLMCPR is based on the rigorous Monte Carlo analysis.

The items in Table 3 that result in the increase of the estimated SLMCPR are discussed in Section 2.2.

Cycle 15 will be the first full reload of GNF2 for LaSalle Unit 1. The critical power uncertainty for GNF2 is defined in Table 6. As seen in Table 6, the critical power uncertainty for GNF2 is lower than the previous cycle's fuel type (ATRIUM-10). As such, the GEXL uncertainty of the new fuel type tends to make the final SLMCPR lower than that which would be calculated using only the critical power uncertainties for the previous cycle's fuel type. However, GNF did not supply reload fuel or determine any design or licensing calculations for the previous cycle. As such, comparisons of SLMCPR analysis bases, assumptions and results are limited.

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 the NRC-approved values. Also, estimated effect 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 ((

)) to account for an increase in channel bow due to the emerging unforeseen phenomena called control blade shadow corrosion-induced channel bow, which is not accounted for in the channel bow uncertainty component of the approved R-Factor uncertainty. The step "a RPEAK" in Figure 4.1 from NEDC-32601P-A, which has been provided for convenience in Figure 3, 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 (( )).

LaSalle Unit I has experienced control blade shadow corrosion-induced channel bow to the extent that an increase in the NRC-approved R-Factor uncertainty (( )) is deemed prudent to address its effect. Accounting for the control blade shadow corrosion-induced channel bow, Discussion Page 5 of 26

Non-Proprietary Information - Class I (Public) the LaSalle Unit 1 Cycle 15 analysis shows an expected channel bow uncertainty of

(( )), which is bounded by a GEXL R-Factor uncertainty of (( )). Thus, the use of a GEXL R-Factor uncertainty of (( )) adequately accounts for the expected control blade shadow corrosion-induced channel bow for LaSalle Unit 1 Cycle 15.

2.2.2. Core Flow Rate and Random Effective TIP Reading In Reference 5 GNF committed to the expansion of the state points used in the determination of the SLMCPR. Consistent with the Reference 5 commitments, GNF performs 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. The approved SLMCPR methodology is applied at each state point that is analyzed.

For the TLO calculations performed at 82.8% core flow, the approved uncertainty values for the core flow rate (2.5%) and the random effective Traversing In-Core Probe (TIP) reading (1.2%)

are conservatively adjusted by dividing them by 82.8/100. The steps "o CORE FLOW" and "6 TIP (INSTRUMENT)" in Figure 4.1 from NEDC-32601P-A, which has been provided for convenience in Figure 3, are affected by this deviation, respectively.

Historically, these values have been construed to be somewhat dependent on 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 because 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.

Discussion Page 6 of 26

Non-Proprietary Information - Class I (Public) 2.3. Departure from NRC-Approved Methodology No departures from NRC-approved methodologies were used in the LaSalle Unit 1 Cycle 15 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 (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:

11 11 For the Atrium-10 fuel product from AREVA, no axial power shape penalty is applied (Reference 10). 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 with the following table:

11 11 If the limiting bundles in the SLMCPR calculation exhibit an axial power shape identified by this table, GNF penalizes the GEXL critical power uncertainties to conservatively account for the effect of the axial power shape. Table 6 provides a list of the GEXL critical power uncertainties determined in accordance with the NRC-approved methodology contained in NEDE-24011-P-A along with values actually used.

Discussion Page 7 of 26

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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 LaSalle Unit I Cycle 15 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-24011-P-A (March 11, 1999) are addressed in References 1, 2, 3, and 9.

The four restrictions for GNF2 were determined acceptable by the NRC review of "GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II)," NEDC-33270P, Revision 0, March 14, 2007. Specifically, in the NRC audit report ML081630579 for the said document, Section 3.4.1 (page 59) states:

"The NRC staffs SE of NEDC-32694P-A (Reference 19 of NEDC-33270P) provides four actions to follow whenever a new fuel design is introduced. These four conditions are listed in Section 3.0 of the SE. The analysis and evaluation of the GNF2 fuel design was evaluated in accordance with the limitations and conditions stated in the NRC staff' s SE, and is acceptable."

GNF's position is that GNF2 is an evolutionary fuel product based on GE14. It is not considered a new fuel design as it maintains the previously established 10x10 array and 2 water rod makeup, as stated by the NRC audit report ML081630579, Section 3 .4.2.2.1 (page 59):

"The NRC staff finds that the calculational methods, evaluations and applicability of the OLMCPR and SLMCPR are in accordance with existing NRC-approved methods and thus valid for use with GNF2 fuel."

As such, no new GNF fuel designs are being introduced in LaSalle Unit 1 Cycle 15; therefore, the NEDC-32505P-A statement "...if new fuel is introduced, GENE must confirm that the revised R-Factor method is still valid based on new test data" is not applicable.

2.6. Minimum Core Flow Condition For LaSalle Unit I Cycle 15, the minimum core flow SLMCPR calculation performed at 82.8%

core flow and rated core power condition was limiting as compared to the rated core flow and rated core power condition. For convenience, Figures III.5-1 and 111.5-2 from NEDC-32601P-A have been provided in Figures 4 and 5, respectively, to show this minimum core flow condition relative relationship to the data on these figures. For this condition, the MIP ((

Discussion Page 8 of 26

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)) 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 pattern used to calculate the SLMCPR at 100% rated power/82.8%

rated flow 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 LaSalle Unit 1 Cycle 15. Consequently, the SLMCPR value calculated from the 82.8% core flow and rated core power condition limiting MCPR distribution reasonably bounds this mode of operation for LaSalle Unit 1 Cycle 15.

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 LaSalle Unit 1 Cycle 15.

2.8. Core Monitoring System The utility has requested that GNF perform the SLMCPR calculation applying the GETAB power distribution methodology and uncertainties. Due to presence of third party proprietary information, the utility has provided, in a separate attachment, the basis that the GETAB power distribution methodology and uncertainties are applicable for the POWERPLEX-III core monitoring system.

2.9. Power/Flow Map The utility has provided the current and previous cycle power/flow map in a separate attachment.

2.10. Core Loading Diagram Figures 1 and 2 provide the core-loading diagram for the current and previous cycle, respectively, which are the Reference Loading Pattern as defined by NEDE-24011-P-A. Table I provides a description of the core.

2.11. Figure References Figure 3 is Figure 4 .1 from NEDC-32601P-A. Figure 4 is Figure III.5-1 from NEDC-32601P-A.

Figure 5 is based on Figure 111.5 -2 from NEDC-32601P-A, and has been updated with GE14 and GNF2 data.

Discussion Page 9 of 26

Non-Proprietary Information - Class I (Public) 2.12. Additional SLMCPR Licensing Conditions For LaSalle Unit 1 Cycle 15, no additional SLMCPR licensing conditions are included in the analysis.

2.13. 10 CFR Part 21 Evaluation There are no known 10 CFR Part 21 factors that affect the LaSalle Unit I Cycle 15 SLMCPR calculations.

2.14. Summary The requested changes to the Technical Specification SLMCPR values are 1.13 for TLO and 1.15 for SLO for LaSalle Unit I Cycle 15.

Discussion Page 10 of 26

Non-Proprietary Information - Class I (Public) 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), "Confinnation of 10x10 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 GEXL14 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 31 - August 1, 2002," FLN-2002-015, October 31, 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 10X10 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 GE14 Critical Power Correlation for Outlet Peaked Axial Power Shapes,"

FLN-2007-031, September 18, 2007.

9. Letter, Andrew A. Lingenfelter (GNF-A) to U.S. Nuclear Regulatory Commission Document Control Desk with cc to SS Philpott (NRC), "Amendment 33 to NEDE-24011-P, General Electric Standard Application for Reactor Fuel (GESTAR II) and GNF2 Advantage Generic Compliance with NEDE-24011-P-A (GESTAR II),

NEDC-33270P, Revision 3, March 2010," MFN 10-045, March 5, 2010.

10. Global Nuclear Fuel, "GEXL97 Correlation Applicable to ATRIUM-10 Fuel",

NEDC-33106P, Revision 2, June 2004.

References Page 11 of 26

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11 Figure 3. Figure 4.1 from NEDC-32601P-A Figure 3. Figure 4.1 from NEDC-32601 P-A Page 14 of 26

Non-Proprietary Information - Class I (Public) 11 11 Figure 4. Figure III.5-1 from NEDC-32601P-A Figure 4. Figure III .5-1 from NEDC-32601P-A Page 15 of 26

Non-Proprietary Information - Class I (Public) 11 11 Figure 5. Relationship Between MIP and CPR Margin Figure 5. Relationship Between MIP and CPR Margin Page 16 of 26

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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 the Core 764 764 Limiting Cycle Exposure Point (i.e., Beginning of Cycle (BOC)/Middle of N/A N/A FOR FOR Cycle (MOC)/End of Cycle (EOC))

Cycle Exposure at Limiting Point N/A N/A 13971 13971 (MWd/STU)

% Rated Core Flow N/A N/A 82.8 100.0 Reload Fuel Type ATRIUM-10 GNF2 Latest Reload Batch Fraction, % 41.9 38.7 Latest Reload Average Batch Weight % 4.03 3.97 Enrichment Core Fuel Fraction:

ATRIUM-10 1 0.613 GNF2 0 0.387 Core Average Weight %

Enrichment 3 . 97 3 . 99 Table 1. Description of Core Page 17 of 26

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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 Uncertainty N/A NEDC-32601-P-A Power Distribution Methodology N/A NEDC-32601-P-A Power Distribution Uncertainty N/A NEDC-32601-P-A Core Monitoring System POWERPLEX-III POWERPLEX-III R-Factor Calculation Methodology N/A NEDC-32505P-A Table 2. SLMCPR Calculation Methodologies Page 18 of 26

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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 11 Table 3. Monte Carlo Calculated SLMCPR vs. Estimate Page 19 of 26

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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 11 Table 3. Monte Carlo Calculated SLMCPR vs. Estimate Page 20 of 26

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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

+/-s(%) Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case GETAB Feedwater Flow 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 6.0 SLO/2.5 TLO N/A N/A N/A N/A Measurement Channel Flow Area 3.0 N/A N/A N/A N/A Variation Friction Factor 10.0 N/A N/A N/A N/A Multiplier Channel Friction 5.0 N/A N/A N/A N/A Factor Multiplier Table 4. Non-Power Distribution Uncertainties Page 21 of 26

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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

+/- t6 (%) Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case NEDC-32601P-A Feedwater Flow Measurement N/A N/A Feedwater Temperature N/A N/A Measurement Reactor Pressure Measurement N/A N/A Core Inlet Temperature 0.2 N/A N/A 0.2 0.2 Measurement Total Core Flow Measurement 6.0 SLO/2.5 TLO N/A N/A 3.02 2.5 Channel Flow Area Variation N/A N/A Friction Factor Multiplier N/A N/A Channel Friction 5.0 N/A N/A 5.0 5.0 Factor Multiplier Table 4. Non-Power Distribution Uncertainties Page 22 of 26

Non-Proprietary Information - Class I (Public)

Table 5. Power Distribution Uncertainties Nominal (NRC- Previous Cycle Previous Cycle Current Cycle Current Cycle Description Approved) Value Minimum Core Rated Core Flow Minimum Core Rated Core Flow f a (%) Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case GETAB/NEDC -32601P-A GEXL R-Factor (( )) N/A N/A (( )) (( ))

Random Effective 2.85 SLO/1.2 TLO N/A N/A 1.45 1.2 TIP Reading Systematic Effective TIP Reading 8.6 N/A N/A 5.56 5.56 NEDC-32694P-A, 3DMONICORE GEXL R-Factor (( )) N/A N/A N/A N/A Random Effective 2 . 85 SLO/1.2 TLO N/A N/A N/A N/A TIP Reading TIP Integral (( )) N/A N/A N/A N/A Four Bundle Power Distribution (( ))

Surrounding TIP N/A N/A N/A N/A Location Table 5. Power Distribution Uncertainties Page 23 of 26

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Table 5. Power Distribution Uncertainties Nominal (NRC- Previous Cycle Previous Cycle Current Cycle Current Cycle Description Approved) Value Minimum Core Rated Core Flow Minimum Core Rated Core Flow f a (%) Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case Contribution to Bundle Power Uncertainty Due to Local Power Range (( )) N/A N/A N/A N/A Monitor (LPRM)

Update Contribution to Bundle Power Due to (( )) N/A N/A N/A N/A Failed TIP Contribution to Bundle Power Due to (( )) N/A N/A N/A N/A Failed LPRM Total Uncertainty in Calculated Bundle (( )) N/A N/A N/A N/A Power Uncertainty of TIP Signal Nodal (( )) N/A N/A N/A N/A Uncertainty Table 5. Power Distribution Uncertainties Page 24 of 26

Non-Proprietary Information - Class I (Public)

Table 6. Critical Power Uncertainties Nominal Value Previous Cycle Previous Cycle Current Cycle Current Cycle Description Minimum Core Rated Core Flow Minimum Core Rated Core Flow f a ^0^Q^

Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case

))

Table 6. Critical Power Uncertainties Page 25 of 26

Non-Proprietary Information - Class I (Public)

Table 6. Critical Power Uncertainties Nominal Value P revious Cycle Previous Cycle Current Cycle Current Cycle Description Minimum Core Rated Core Flow Minimum Core Rated Core Flow 6 (%)

Flow Limiting Case Limiting Case Flow Limiting Case Limiting Case 11 11 Table 6. Critical Power Uncertainties Page 26 of 26

ATTACHMENT 9 LASALLE COUNTY STATION UNIT 1 Docket No. 50-373 License No. NPF-11 Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Supplemental Peach Bottom RAI Responses Applied to LaSalle County Station, Unit 1, Cycle 15 - Non-Proprietary Version

ENCLOSURE 4 CFL-EXN-HH1-11-120 Supplemental Peach Bottom RAI Responses Applied to LaSalle Unit 1 Cycle 15 Non-Proprietary Information - Class I (Public)

INFORMATION NOTICE This is a non-proprietary version of CFL-EXN- HH1-11-120 Enclosure 3, which has the proprietary information removed. Portions of the document that have been removed are indicated by white space inside an open and closed bracket as shown here (( I].

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 1 of 14 Nuclear Regulatory Commission (NRC) Docket No. 50-278 contains Requests for Additional Information (RAIs) related to the License Amendment Request (LAR) for Technical Specification changes to the Safety Limit Minimum Critical Power Ratio (SLMCPR) values for Peach Bottom Atomic Power Station (PBAPS) Unit 3. In order to assist the review of the LaSalle Unit 1 SLMCPR Technical Specification LAR, Global Nuclear Fuel (GNF) is including responses to these RAIs as applied to LaSalle Unit 1. Please note that references to "Attachment 4" in the PBAPS RAls are equivalent to Enclosure 1 to CFL-EXN-HA1-11-120 entitled "GNF Additional Information Regarding the Requested Changes to the Technical Specification SLMCPR,"

hereafter referred to as Enclosure 1.

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 2 of 14 PBAPS Unit 3 - RAI-01:

Provide the cycle-specific fuel quantity for each fuel type and state when the specific fuel types are loaded in the core (i.e., fresh, once, or twice burn) as depicted in Figure 1 of Attachment 4 for the Cycle 19 core loading diagram.

GNF RESPONSE TO RAI.01. 1 - Applied to LaSalle Unit 1:

GNF provides the following table for clarification.

Table RAI-01-1 Figure 1 of Enclosure I (Cycle 15) Core Description - Fuel Type, Bundle Name, Number of Bundles, and Cycle Loaded Fuel' Number of Circle Bundle Ns

Bundles Loaded Irradiated:

7 ATRM10-PlOCAZB391-12GZ-1000 400M-9WR-149-16-3094 16 13 8 ATRM10-P10CAZB404-14GZ-1000 400M-9WR-149-T6-3095 72 13 9 ATRM10-P10CAZB392-14GZ-1000 400M-9WR-149-T6-3096 28 13 10 ATRM10-P10CAZB406-14GZ-1000-400R-9WR-149-16-3998 64 14 11 ATRM 10-P10CAZB397-16GZ-1000-400R-9WR-149-T6-3999 120 14 12 ATRM10-P10CAZB406-16GZ-1000-400R-9WR-149-T6-4004 136 14 17 ATRM10-PlOCAZB391-12GZ-1000 400M-9WR-149-16-3094 23 13 19 ATRM10-PlOCAZB392-14GZ-1000 400M-9WR-149-T6-3096 9 13 Fresh:

1 GNF2-P10CG2B410-17GZ-12012-150-T6-4070 72 15 2 GNF2-P10CG2B408-12GZ-120T2-150-16-4069 72 15 3 GNF2-P10CG2B395-15GZ-12012-150-T6-4067 48 15 4 GNF2-P10CG2B395-13GZ-12012-150-T6-4066 16 15 5 GNF2-P10CG2B395-15GZ-12012-150-T6-4067 48 15 6 GNF2-P10CG2B329-13GZ-12012-150-T6-4068-LTD 24 15 13 GNF2-P10CG2B395-13GZ-120T2-150-16-4066 8 15 20 GNF2-P10CG2B408-12GZ-12012-150-T6-4069 8 15

CFL-EXN-HA1 120 Non -Proprietary Information - Class I (Public)

Enclosure 4 Page 3 of 14 PBAPS Unit 3 - RAI-02:

Provide the information to obtain a final core loading pattern as shown in Figure 1 of Attachment 4 including procedures, guidelines, criteria, and approved methodologies used for this analysis.

GNF RESPONSE TO RAI Applied to LaSalle Unit 1:

The loading pattern is developed collaboratively by GNF and Exelon based on Exelon input.

Among the inputs are:

• Cycle Energy Requirements - fuel bundle design (nuclear) and loading patterns

• Thermal Limit Margins

• Reactivity Margins - minimum shutdown margin, minimum and maximum hot excess reactivity

• Discharge Exposure Limitations and Other Limits as established by safety analysis

• Desired Control Rod Patterns - sequences and durations

• Channel Distortion Minimization Methods used to analyze the core-loading pattern are in accordance with GESTAR II.

GESTAR II is the umbrella for all procedures, guidelines, criteria, and methodologies used for this analysis. There is no change in approved methodologies. This is a SLMCPR Technical Specifications change within approved methodologies. SLMCPR is not the primary driver in developing the fuel cycle core design. The energy plan, reactivity, and thermal margins are the primary drivers.

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 4 of 14 PBAPS Unit 3 - RAI-03:

Provide the rationale for why a 35.1 % reload batch fraction for GNF2 fuel caused the proposed SLMCPR to change by 0.02 for two recirculation loop operation and 0.03 for single recirculation loop operation for the proposed loading pattern in Figure 1 of Attachment 4.

GNF Response to RAI Applied to LaSalle Unit 1:

Section 2.1 of Enclosure 1 describes the major contributors to the SLMCPR change. In addition to the explanation provided in this section, Table 6 of Enclosure 1 provides a list of the GEXL critical power uncertainties determined in accordance to the NRC-approved methodology contained in NEDE-2401 1 -P-A along with the values actually used. LaSalle Unit 1 Cycle 15 uses GNF methods, as opposed to Cycle 14 which did not, and due to the use of GNF methods the legacy fuel in the core has a much higher critical power uncertainty value than is typical for GNF fuel, which contributes to an increase in the SLMCPR.

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 5 of 14 PBAPS Unit 3 - RAI-04:

Confirm that the fuel-related coefficients and constants are the same in the approximation of the correlation for the MCPR Importance Parameter (MIP) and the R-factor Importance Parameter (RIP) for all of the fuels shown in Figure 5 of Attachment 4.

GNF Response to RAI Applied to LaSalle Unit 1:

All of the fuels shown in Figure 5 of Enclosure 1 use the same coefficients and constants in the approximation of the correlation for the MIP and RIP. The correlation provides an estimate to check the reasonableness of the Monte Carlo result. It is not used for any other purpose. The methodology and final SLMCPR is based on the rigorous Monte Carlo analysis. A description of the correlation used for the Two Loop Operation (TLO) SLMCPR estimate using the MIPRIP correlation is provided below.

((

11

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public)

Enclosure 4 Page 6 of 14 Background for RAI-05.1 -RAI-05.3:

Section 2.1, "Major Contributors to SLMCPR Change," states that Table 3 presents estimated impacts on the TLO SLMCPR due to methodology deviations, penalties, uncertainties and/or deviations from approved values.

PBAPS Unit 3 - RAI-05.1:

Provide calculation details and justify that the results listed in Table 3 are conservative related to methodology deviations, penalties, uncertainties and/or deviations from approved values.

GNF Response to RAI -05.1 - Applied to LaSalle Unit 1:

The Monte Carlo TLO and Single Loop Operation (SLO) results listed in Table 3 of Enclosure 1 are conservative related to methodology deviations, penalties, and/or uncertainties deviations from approved values. Section 2.2 of Enclosure 1 discusses the deviations from the NRC-approved values.

PBAPS Unit 3 - RAI-05.2:

Provide a qualitative explanation of the impact on the SLMCPR estimate at rated power and rated flow versus minimum core using MIPRIP correlation as described in Section 2.1 of .

GNF Response to RAI -05.2 - Applied to LaSalle Unit 1:

Section 2.1 of Enclosure 1 states that if the 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 cases, and that this is done only to provide some reasonable assessment basis for the minimum core flow trend. As described in Table 3 of Enclosure 1, an additional uncertainty of 0.003 is applied to the estimated SLMCPR for the low core flow case as a result.

PBAPS Unit 3 - RAI-05.3:

Provide a justification that all affected factors including any fuel-related Part 21 issues are reflected in Table 3.

GNF Response to RAI - 05.3 - Applied to LaSalle Unit 1:

There are no known 10 CFR Part 21 factors that affect the LaSalle Unit 1 Cycle 15 SLMCPR calculations.

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 7 of 14 PBAPS Unit 3 - RAI-06:

Provide a reactor core map that depicts the 0.1 percent of fuel bundles that may experience boiling transition for the limiting SLMCPR case. Include information regarding the fuel bundle group, group exposure, the number of bundles, fuel type and the percent contribution to the number of fuel rods that are subjected to boiling transition.

GNF Response to RAI Applied to LaSalle Unit 1:

The bundle groupings for the TLO SLMCPR calculations are shown in Table RAI-06-1, along with the number of bundles in the group, their contribution to percent number of rods in boiling transition (NRSBT) and the group average exposure at the analysis point. The 2-dimensional core map of the bundle groupings is shown in Figure RAI-06-1 for the upper left hand quadrant in the core. The bundle groupings for the SLO SLMCPR calculations are shown in Table RAI-06-2, along with the number of bundles in the group, their contribution to the percent NRSBT and the group average exposure at the analysis point. ((

)). Both the TLO and SLO are ((

CFL-EXN-HA1 120 Non-Proprietary Information - Class I (Public) Page 8 of 14 Table RAI-06-1 Bundle Group, Number of Bundles, Bundle Type, % Contribution to NRSBT, and Group Exposure for TLO

((

11

((

11 Figure RAI-06-1 Two-Dimensional Map of the Bundle Groupings for Percent Contribution to NRSBT for TLO

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 9 of 14 Table RAI-06-2 Bundle Group, Number of Bundles, Bundle Type, % Contribution to NRSBT, and Group Exposure for SLO

((

11

((

11 Figure RAI-06-2 Two-Dimensional Map of the Bundle Groupings for Percent Contribution to NRSBT for SLO

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 10 of 14 PBAPS Unit 3 - RAI-08:

Identify the sections/pages of Reference 02-1: Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011 P-A-18 and NEDE-24011 P-A-18-US, April 2011, that are applicable to the inputs identified in the response to RAI-02. In addition, provide a quantitative range for the referenced thermal limit margins and the reactivity margins.

GNF Response to RAI Applied to LaSalle Unit 1:

While GESTAR II is the umbrella regulatory document for reload activities from a safety criteria and work scope standpoint; it does not specify all design considerations or parameters.

Therefore, a one-to-one correspondence with the cycle "design" considerations as listed in the RAI-02 response does not exist.

GESTAR II, NEDE-24011 P-A-18, Section 3.1 (Page 3-1) includes the safety criteria to be met by the core design:

3.1.1 Reactivity Basis The nuclear design shall meet the following basis: The core shall be capable of being made subcritical at any time or at any core condition with the highest worth control rod fully withdrawn.

3.1.2 Overpower Bases The Technical Specification limits on Minimum Critical Power Ratio (MCPR), the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) and the Linear Heat Generation Rate (LHGR) are determined such that the fuel will not exceed required licensing limits during abnormal operational occurrences or accidents.

Table 3-1 of GESTAR II (Page 3-12) contains more details of the thermal limits applied to nuclear designs.

The design considerations as listed in the RAI-02 response relate to customer specified parameters for the cycle design. Each reload design must meet the GESTAR II safety criteria.

These criteria are reflected in the plant Technical Specifications and are monitored for compliance. The design considerations do not affect the requirement to meet the safety criteria or the methodology used in the analysis.

Based on customer input, the following quantitative design values were applied to the LaSalle Unit 1 Cycle 15 core design:

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public) Page 11 of 14

• Cycle Energy Requirement 0 ((

11

• Thermal Limit Margins 0 ((

11

• Reactivity Margins 0 ((

11

• Discharge Exposure Limitations 0 ((

11

• Desired Control Rod Patterns o Collaborative decision making process with Exelon based on core management philosophy

• Channel Distortion Minimization o Collaborative decision making process with Exelon based on fuel channel to control blade interference monitoring philosophy The determination of the SLMCPR, which is the subject of the Technical Specification LAR, is performed in the same manner regardless of the specified core design requirements and margins.

CFL-EXN-HA1 120 Non-Proprietary Information - Class I (Public)

Enclosure 4 Page 12 of 14 PBAPS Unit 3 - RAI-09:

Identify the sections/pages of Reference 03-1: Global Nuclear Fuel, "General Electric Standard Application for Reactor Fuel (GESTAR II)," NEDE-24011P-A-18 and NEDE-24011P-A-18-US, April 2011, that describe the SLMCPR calculation related to GNF2 fuel assemblies as discussed in the response to RAI-03.

GNF Response to RAI Applied to LaSalle Unit 1:

There appears to be some clarification needed in the parsing of this statement in the RAI-03 response:

"In addition to the explanation provided in this section, Table 6 of Attachment 4 provides a list of the GEXL critical power uncertainties determined in accordance to the NRC-approved methodology contained in NEDE-2401 1 -P-A along with the values actually used."

1St: Table 6 of Enclosure 1 includes the list of uncertainties and actual values used for the LaSalle Unit 1 Cycle 15 SLMCPR LAR.

2"d: The NRC-approved methodologies are contained in GESTAR 11 as well as being documented in Enclosure 1. The plant and product line specific parameters are not contained in GESTAR II.

An overview of the SLMCPR process is presented in Section 1.1.5 of GESTAR II, NEDE-24011P-A-18 (Page 1-4). Sections 4.3.1, 4.3.1.1, 4.3.1.1.1 and 4.3.1.1.2 (Pages 4-7 and 4-8) contain more details on the calculational method to derive the SLMCPR.

As described in GESTAR II Section 4.3.1.1.1, further details on the procedure are presented in Appendix IV of GESTAR II Reference 4-9 and Section 4 of GESTAR 11 Reference 4-36. The uncertainties used for the LaSalle Unit 1 cycle-specific statistical analyses are presented in GESTAR II Reference 4-37, as well as Table 4, Table 5 and Table 6 of Enclosure 1.

GESTAR II provides the approved SLMCPR process and methodology references; therefore, there is no specific mention of the GNF2 product line. The GNF2 fuel product was licensed via the provisions of GESTAR II Section 1.1, Fuel Licensing Acceptance Criteria, which culminated in the Compliance Report for GNF2, NEDC-33270P. Revision 0 of NEDC-33270P was submitted in March 2007 and was subsequently audited by the NRC. As noted in Section 2.5 of , the NRC has reviewed the applicability of GNF2 to the SLMCPR process and audit report ML081630579, Section 3.4.2.2.1 page 59 states:

"The NRC staff finds that the calculational methods, evaluations and applicability of the OLMCPR and SLMCPR are in accordance with existing NRC-approved methods and thus valid for use with GNF2 fuel."

CFL-EXN-HA1 120 Non- Proprietary Information - Class I (Public) Page 13 of 14

References:

GESTAR 11 4-9: General Electric BWR Thermal Analysis Basis (GETAB): Data, Correlation and Design Application, NEDE-10958-PA and NEDO-10958-A, January 1977.

GESTAR 11 4-36: Methodology and Uncertainties for Safety Limit MCPR Evaluation, NEDC-32601P-A, August 1999.

GESTAR 114-37: Power Distribution Uncertainties for Safety Limit MCPR Evaluations, NEDC-32694P-A, August 1999.

CFL-EXN-HA1-11-120 Non-Proprietary Information - Class I (Public)

Enclosure 4 Page 14 of 14 PBAPS Unit 3 - RAI-10:

Provide the calculation details requested in RAI-05.1 that justify the results listed in Table 3 of from the June 8, 2011, submittal.

GNF Response to RAI Applied to LaSalle Unit 1:

Please see Enclosure 5, which contains an overview of the core design process. Note that this enclosure is a modified version of a presentation that was provided to the NRC on August 10, 2010 in response to an NRC Audit on the Fitzpatrick C20 SLMCPR Technical Specification Change letter. Slight modifications were made to the presentation in order to include data specific to the LaSalle Unit 1 Cycle 15 core design.

ATTACHMENT 10 LASALLE COUNTY STATION UNIT 1 Docket No. 50-373 License No. NPF-1 1 Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit LaSalle County Station, Unit 1, Cycle 14 and Expected Cycle 15, Power/Flow Map

Core Flow (Percent) 0 5 10 15 20 25 30 35 40 45 50 55 60 65 70 75 80 85 90 95 100 105 120 115 110 REGION 2 105 100 95 90 85 80 REGION 1 75 m 70 65 c

a 60 55 50 45 40 35 0 209fi FW Flow ^ a 30 25 20 Low Recirc Pump Speed 15 Max Valve Pos.

10 (Approximate) 5 0

Core Flow (M#/hr)

Attachment 10: LaSalle County Station, Unit 1, Cycle 14 and Expected Cycle 15, Power-to-Flow Map