RS-12-036, License Amendment Request to Technical Specifications Section 5.6.5, Core Operating Limits Report (Colr).

License Amendment Request to Technical Specifications Section 5.6.5, Core Operating Limits Report (Colr).

ML12094A328
Person / Time
Site:	Dresden, Quad Cities
Issue date:	04/03/2012
From:	Gullott D Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RS-12-036
Download: ML12094A328 (13)

Text

Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road Nuclear Warrenville, IL 60555 RS-12-036 10 CFR 50.90 April 3, 2012 U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265

Subject:

License Amendment Request to Technical Specifications Section 5.6.5, "Core Operating Limits Report (COLR)"

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 Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS),

Units 1 and 2. The proposed change will add an NRC approved topical report reference to the list of analytical methods in TS Section 5.6.5, "Core Operating Limits Report (COLR)," that are used to determine the core operating limits. Specifically, the following reference will be added:

Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application."

This TR will be used for its proposed method of determining the end of lower plenum flashing for analysis of a BWR Loss-of-Coolant Accident (LOCA).

The proposed change is necessary in order to use the most recent Westinghouse methodology to determine overall core operating limits for future core configurations.

U. S. Nuclear Regulatory Commission April 3, 2012 Page 2 This attached amendment request is subdivided as follows:

Attachment 1 provides a description and evaluation of the proposed change.

Attachments 2 and 3 provide the marked-up TS pages for DNPS and QCNPS, respectively, with the proposed change indicated.

The proposed change has been reviewed by the DNPS and QCNPS Plant Operations Review Committees and approved by the EGC 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 change to the TS by transmitting a copy of this letter and its attachments to the designated State Official.

EGC requests approval of the proposed amendment by April 3, 2013. Once approved, the amendment will be implemented within 30 days.

The proposed change does not include any new commitments. Should you have any questions concerning this letter, please contact Ms. Dwi Murray at (630) 657-3695.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 3rd day of April 2012.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC Attachments: 1. Evaluation of Proposed Changes

2. Markup of Proposed Technical Specifications Pages for DNPS

3. Markup of Proposed Technical Specifications Pages for QCNPS cc: USNRC Project Manager, NRR - Dresden and Quad Cities Nuclear Power Stations USNRC Regional Administrator - Region III USNRC Senior Resident Inspector - Dresden Nuclear Power Station USNRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Changes

Subject:

License Amendment Request to Technical Specifications Section 5.6.5, "Core Operating Limits Report (COLR)"

1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

4.1 Applicable Regulatory Requirements/Criteria 4.2 Precedent 4.3 No Significant Hazards Consideration 4.4 Conclusions

6.0 REFERENCES

Page 1 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 1.0

SUMMARY

DESCRIPTION This evaluation supports Exelon Generation Company, LLC, (EGC) request to amend Appendix A, Technical Specifications (TS), of Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS),

Units 1 and 2, in accordance with 10 CFR 50.90. The proposed change will add an NRC approved topical report reference to the list of analytical methods in TS Section 5.6.5, "Core Operating Limits Report (COLR)," that are used to determine core operating limits.

The proposed addition of the Westinghouse topical report to TS 5.6.5 is necessary in order to use the most recent Westinghouse methodology to determine overall core operating limits for future core configurations.

2.0 DETAILED DESCRIPTION The proposed change will add the following reference to TS 5.6.5:

Westinghouse Electric Company (Westinghouse) Topical Report (TR) WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application."

This TR will be used for its method of determining the end of lower plenum flashing for analysis of a BWR Loss-of-Coolant Accident (LOCA).

As stated in DNPS and QCNPS TS 5.6.5.b, "The COLR will contain the complete identification for each of the TS referenced topical reports used to prepare the COLR (Le., report number, title, revision, date, and any supplements)." Therefore, the marked-up TS pages contained in Attachments 2 and 3 will reference the new topical report as "WCAP-16865-P-A, 'Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application.'"

3.0 TECHNICAL EVALUATION

DNPS and QCNPS TS 5.6.5 requires that a COLR be established and the analytical methods used to determine the core operating limits shall be those previously reviewed and approved by the NRC. TS 5.6.5.b lists the analytical codes, or the topical reports describing these codes, that are used to calculate operating parameters and predict the core behavior under normal and accident conditions. Specifically, the approved analytical methods listed in TS 5.6.5.b support operation of the SVEA-96 Optima2 fuel contained in the reactor core.

Application of a new NRC approved Westinghouse methodology to support operation of the SVEA-96 Optima2 fuel contained in the reactor core, including overall core operating limits for future core configurations, requires a revision to TS 5.6.5.b. The proposed change will also allow DNPS and QCNPS to use the most current Westinghouse methodology for determination Page 2 of 7

ATTACHMENT 1 Evaluation of Proposed Changes of the APLHGR limits associated with Technical Specifications 3.2.1, "Average Planar Linear Heat Generation Rate (APLHGR)."

The NRC, in Reference 2, reviewed topical report WCAP-16865-P/wCAP-16865-NP, Revision 1, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application." This topical report is to be used for its proposed method of determining the end of lower plenum flashing for analysis of a BWR LOCA. The change credits steam cooling for a longer time than the previous evaluation model resulting in a corresponding reduction in the peak clad temperature (PCT). The NRC concluded that the topical report was acceptable for referencing in licensing applications to the BWR LOCA ECCS Evaluation Model to the extent specified, and subject to the limitations and conditions described in the final safety evaluation.

There are two limitations and conditions specified in Reference 2, Section 5, "Limitations and Conditions." Note that the two specific limitations and conditions contain proprietary information and are not repeated in this section.

EGC confirms that the method to determine the end of lower plenum flashing proposed for use in the DNPS and QCNPS LOCA analyses will include the two specific requirements described in Reference 2, Section 5, "Limitations and Conditions." These two requirements will be implemented by the use of the appropriate Westinghouse codes and procedures.

The validation of the proposed methodology described in Reference 1 included methodology code comparisons to Two-Loop Test Apparatus (TLTA) and The Rig of Safety Assessment (ROSA-III) test results to demonstrate that the code properly accounts for the physical phenomena associated with steam cooling, lower plenum flashing and flow splits between bundles having different powers. This demonstration used DNPS and QCNPS as the sample model for BWRl3 plants. There were no conditions placed on using the model that would restrict the application of Reference 1 to any particular BWR design. The validations provided by Westinghouse demonstrate that the proposed method described in Reference 1 to determine the end of lower plenum flashing is acceptable because it will continue to result in a conservative safety analysis.

Also note that there are no power level restrictions associated with application of the methodology described in Reference 1, therefore, the methodology is appropriate for use under future planned power uprate conditions at both DNPS and QCNPS.

4.0 REGULATORY EVALUATION

4.1 Applicable Requlatory Requirements/Criteria Technical Specifications (TS) Section 5.6.5 lists the NRC approved analytical methods used at DNPS and QCNPS to determine the core operating limits. The listed NRC approved analytical methods provide the necessary administrative controls to ensure operation of the facility in a safe manner and thus, are required for inclusion in the DNPS and QCNPS Technical Specifications in accordance with 10 CFR 50.36, "Technical Specifications," paragraph (c)(5).

Page 3 of 7

ATTACHMENT 1 Evaluation of Proposed Changes 4.2 Precedents

1. Letter from K. R. Jury (EGC) to USNRC, "Request for Amendment to Technical Specifications Section 5.6.5, 'Core Operating Limits Report (COLR),"' dated March 7, 2005

2. Letter from S. P. Sands (USNRC) to C. M. Crane (EGC), "LaSalle County Station, Units 1 and 2 - Issuance of Amendments Re: Technical Specifications Section 5.6.5,

'Core Operating Limits Report (COLR),"' dated February 1, 2006

3. Letter from K. R. Jury (EGC) to USNRC, "Request for License Amendment to Technical Specifications Section 5.6.5, 'Core Operating Limits Report (COLR),"' dated April 4, 2006

4. Letter from S. P. Sands (USNRC) to C. M. Crane (EGC), "LaSalle County Station, Units 1 and 2 - Issuance of Amendments Re: Technical Specifications Section 5.6.5,

'Core Operating Limits Report,'" dated February 15, 2007 4.3 No Significant Hazards Consideration In accordance with 10 CFR 50.90, Exelon Generation Company, LLC, (EGC), is requesting a change to Appendix A, Technical Specifications (TS), of Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2.

The proposed change will add an NRC approved topical report reference 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. Specifically, the following reference will be added:

Westinghouse TR WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application" This TR will be used for its method of determining the end of lower plenum flashing for analysis of a BWR Loss-of-Coolant Accident (LOCA).

EGC has evaluated whether or not a significant hazards consideration is involved with the proposed amendment by focusing on the three standards set forth in 10 CFR 50.92(c),

"Issuance of amendment," as discussed below:

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

Response: No.

TS 5.6.5 lists NRC approved analytical methods used at DNPS and QCNPS to determine core operating limits. The proposed change adds an NRC approved topical report reference to the list of analytical methods in TS 5.6.5, "Core Operating Limits Report (COLR)."

Page 4 of 7

ATTACHMENT 1 Evaluation of Proposed Changes The proposed change to TS Section 5.6.5 will add a Westinghouse methodology to determine the end of lower plenum flashing for analysis of a BWR LOCA. The proposed change will allow DNPS and QCNPS to use the most current Westinghouse methodology for determination of the APLHGR limits associated with Technical Specifications 3.2.1, "Average Planar Linear Heat Generation Rate (APLHGR)."

The addition of an approved analytical method in 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 NRC approved method ensures that the analysis output accurately models the predicted core behavior, has no effect on the type or amount of radiation released, and has no effect on predicted offsite doses in the event of an accident.

Additionally, the NRC approved method does not change any key core parameters that influence any accident consequences. Thus, the proposed change does 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 increased.

The proposed change in the list of analytical methods does not affect the ability of DNPS and QCNPS to successfully respond to previously evaluated accidents and does not affect the radiological assumptions used in the evaluations. Thus, the radiological consequences of any accident previously evaluated are not increased.

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

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

Response: No.

The proposed change to TS Section 5.6.5 does not affect the performance of any DNPS and QCNPS structure, system or component credited for mitigating any accident previously evaluated. The NRC approved analytical methodology for evaluating the APLHGR limits will not affect the control parameters governing unit operation or the response of the plant equipment to transient conditions. The proposed change does not introduce any new accident precursors, modes of system operation, or failure mechanisms.

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

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

Response: No.

The proposed change will add a reference to the list of analytical methods in TS 5.6.5 that can be used to determine core operating limits. The new methodology has been previously approved by the NRC and accurately establishes the appropriate APLHGR limits. The Page 5 of 7

ATTACHMENT 1 Eval uation of Proposed Changes proposed change does not modify the safety limits or setpoints at which protective actions are initiated and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. Therefore, the proposed change does not impact the level of protection currently provided.

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

Based on the above, EGC concludes that the proposed amendment does not involve a 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.

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 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 processed 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 surveillance requirement and the amendment meets the following specific criteria:

(i) The amendment involves no significant hazards consideration.

As demonstrated in Section 4.3 above, "No Significant Hazards Consideration," the proposed change does not involve any significant hazards consideration.

Page 6 of 7

ATTACHMENT 1 Evaluation of Proposed Changes (ii) There is no significant change in the types or significant increase in the amounts of any effluent that maybe 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; therefore, 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 change 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.

Therefore, in accordance with 10 CFR 51.22(b), no environmental impact statement or environmental assessment need be prepared in connection with the proposed amendment.

6.0 REFERENCES

1. Westinghouse Electric Company Topical Report WCAP-16865-P-A, Revision 1, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application," dated October 2011

2. Letter from J. F. Quichocho (USNRC) to J. A. Gresham (Westinghouse), "Final Safety Evaluation for Westinghouse Electric Company Topical Report WCAP-16865-P/wCAP-16865-NP, Revision 1, 'Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application' (TAC No. ME2901)," dated September 29, 2011 Page 7 of 7

ATTACHMENT 2 DRESDEN NUCLEAR POWER STATION, UNITS 2 AND 3 RENEWED FACILITY OPERATING LICENSE NOS. DPR-19 AND DPR-25 Mark-up of Proposed Technical Specifications Page 5.6-4

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

9. WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors, Supplement 1 to CENPD-287."

10. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

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).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle revisions or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Dresden 2 and 3 5.6-4 Amendment No. 234/227

ATTACHMENT 3 QUAD CITIES NUCLEAR POWER STATION, UNITS 1 AND 2 RENEWED FACILITY OPERATING LICENSE NOS. DPR-29 AND DPR-30 Mark-up of Proposed Technical SpeCifications Page 5.6-4

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 CORE OPERATING LIMITS REPORT (COLR) (continued)

8. WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors - Supplement I."

9. WCAP-15942-P-A, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENPD-287."

10. CENPD-390-P-A, "The Advanced PHOENIX and POLCA Codes for Nuclear Design of Boiling Water Reactors."

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).

c. The core operating limits shall be determined such that all applicable limits (e.g., fuel thermal mechanical limits, core thermal hydraulic limits, Emergency Core Cooling Systems (ECCS) limits, nuclear limits such as SDM, transient analysis limits, and accident analysis limits) of the safety analysis are met.

d. The COLR, including any midcycle reV1Slons or supplements, shall be provided upon issuance for each reload cycle to the NRC.

5.6.6 Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days. The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

Quad Cities 1 and 2 5.6-4 Amendment No. 246/241