RS-11-155, Additional Information Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit: Difference between revisions

From kanterella
Jump to navigation Jump to search
(Created page by program invented by StriderTol)
(Created page by program invented by StriderTol)
 
(2 intermediate revisions by the same user not shown)
Line 19: Line 19:


=Text=
=Text=
{{#Wiki_filter:Exelon Generation Company, LLCwww.exeloncorp.com 4300 Winfield Road Warrenville, IL 60555 clear 10 CFR 50.90 RS-11-155 September 21, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Unit 1 Renewed Facility Operating License No. DPR-29 NRC Docket No. 50-254
{{#Wiki_filter:Exelon Generation Company, LLC      www.exeloncorp.com 4300 Winfield Road                                                                                 clear Warrenville, IL 60555 10 CFR 50.90 RS-11-155 September 21, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Unit 1 Renewed Facility Operating License No. DPR-29 NRC Docket No. 50-254


==Subject:==
==Subject:==
Line 25: Line 25:


==References:==
==References:==
 
: 1. Letter from Mr. Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit," dated June 7, 2011
1.Letter from Mr. Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit," dated June 7, 2011 2.Letter from U. S. NRC to Mr. Michael J. Pacilio (Exelon Nuclear),"Quad Cities Nuclear Power Station, Unit 1 - Request for Additional Information Related to Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAG No. ME6383)," dated August 22, 2011 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Appendix A, Technical Specifications (TS), of Renewed Facility Operating License No.
: 2. Letter from U. S. NRC to Mr. Michael J. Pacilio (Exelon Nuclear),
                              "Quad Cities Nuclear Power Station, Unit 1 - Request for Additional Information Related to Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAG No. ME6383)," dated August 22, 2011 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Appendix A, Technical Specifications (TS), of Renewed Facility Operating License No.
DPR-29 for Quad Cities Nuclear Power Station (QCNPS), Unit 1. The proposed change revises the value of the single recirculation loop operation (SLO) safety limit minimum critical power ratio (SLMCPR) in TS Section 2.1.1, "Reactor Core SLs." Specifically, the proposed change would replace the current SLO SLMCPR requirement for QCNPS Unit 1 with a new SLMCPR requirement. This proposed change does not affect the QCNPS Unit 1 two recirculation loop operation (TLO) SLMCPR or either of the SLMCPR values for Unit 2. This change is needed to support the current cycle of operation (i.e., Cycle
DPR-29 for Quad Cities Nuclear Power Station (QCNPS), Unit 1. The proposed change revises the value of the single recirculation loop operation (SLO) safety limit minimum critical power ratio (SLMCPR) in TS Section 2.1.1, "Reactor Core SLs." Specifically, the proposed change would replace the current SLO SLMCPR requirement for QCNPS Unit 1 with a new SLMCPR requirement. This proposed change does not affect the QCNPS Unit 1 two recirculation loop operation (TLO) SLMCPR or either of the SLMCPR values for Unit 2. This change is needed to support the current cycle of operation (i.e., Cycle
: 22) for QCNPS Unit 1 for cycle exposure greater than 4000 MWd/MT, which is currently scheduled to occur in November 2011.
: 22) for QCNPS Unit 1 for cycle exposure greater than 4000 MWd/MT, which is currently scheduled to occur in November 2011.
In Reference 2, the NRC requested that EGC provide additional information in support of their review of Reference 1. The NRC request for additional information (RAI) and the
In Reference 2, the NRC requested that EGC provide additional information in support of their review of Reference 1. The NRC request for additional information (RAI) and the specific EGC responses are provided in Attachments 1 and 2 to this letter,


specific EGC responses are provided in Attachments 1 and 2 to this letter, September 21, 2011 U. S. Nuclear Regulatory Commission Page 2 Attachment 1 provides the response to RAI 1 and Attachment 2 to this letter provides the responses to RAIs 2 through 6. Attachment 2 contains information proprietary to Westinghouse Electric Company, LLC that is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit, provided in Attachment 3, sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is requested that the information in Attachment 2 be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 4 to this letter provides a non-proprietary version of the RAI responses provided in Attachment 2.
September 21, 2011 U. S. Nuclear Regulatory Commission Page 2 provides the response to RAI 1 and Attachment 2 to this letter provides the responses to RAIs 2 through 6. Attachment 2 contains information proprietary to Westinghouse Electric Company, LLC that is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit, provided in Attachment 3, sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is requested that the information in Attachment 2 be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 4 to this letter provides a non-proprietary version of the RAI responses provided in Attachment 2.
In support of the RAI responses provided in Attachments 1 and 2, several technical reports have been provided as attachments to this letter. Attachments 5 and 6 provide copies of EGC procedures used in the reload control and the bundle and core design process, respectively. Attachment 7 provides a copy of the Cycle Design Inputs and Requirements (i.e., CDIR) document for Quad Cities Unit 1, Cycle 22. Attachment 8 provides EGC Calculation QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan." Attachment 10 provides the Quad Cities Unit 1, Cycle 22 Bundle Design Report and 1 provides the Reference Loading Pattern Report for Quad Cities Unit 1, Cycle 22. Westinghouse Electric Company, LLC considers the information provided in these two reports to be proprietary. Therefore, an affidavit signed by Westinghouse, the owner of the information, is provided in Attachment 9. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390.
In support of the RAI responses provided in Attachments 1 and 2, several technical reports have been provided as attachments to this letter. Attachments 5 and 6 provide copies of EGC procedures used in the reload control and the bundle and core design process, respectively. Attachment 7 provides a copy of the Cycle Design Inputs and Requirements (i.e., CDIR) document for Quad Cities Unit 1, Cycle 22. Attachment 8 provides EGC Calculation QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan." 0 provides the Quad Cities Unit 1, Cycle 22 Bundle Design Report and 1 provides the Reference Loading Pattern Report for Quad Cities Unit 1, Cycle 22. Westinghouse Electric Company, LLC considers the information provided in these two reports to be proprietary. Therefore, an affidavit signed by Westinghouse, the owner of the information, is provided in Attachment 9. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390.
Accordingly, it is requested that the information in Attachments 10 and 11 be withheld from public disclosure in accordance with 10 CFR 2.390. Since the documents provided in Attachments 10 and 11 to this letter are considered to be proprietary in their entirety, non-proprietary versions are not provided.
Accordingly, it is requested that the information in Attachments 10 and 11 be withheld from public disclosure in accordance with 10 CFR 2.390. Since the documents provided in Attachments 10 and 11 to this letter are considered to be proprietary in their entirety, non-proprietary versions are not provided.
Review of the original Quad Cities Unit 1 Cycle 22 SLMCPR report provided as  to Reference 1 indicated several corrections were required. The SLO and DLO SLMCPR values in the report had incorrectly been identified as proprietary. Since these values are included in QCNPS TS 2.1.1, the values should not be withheld as proprietary.
Review of the original Quad Cities Unit 1 Cycle 22 SLMCPR report provided as  to Reference 1 indicated several corrections were required. The SLO and DLO SLMCPR values in the report had incorrectly been identified as proprietary. Since these values are included in QCNPS TS 2.1.1, the values should not be withheld as proprietary. Therefore, the report has been revised to remove the proprietary brackets from around the SLMCPR values. Secondly, it was noted that two minor typographical errors needed correction. In the fifth paragraph on page 5 of 17, the SLMCPR at 95.3%
Therefore, the report has been revised to remove the proprietary brackets from around the SLMCPR values. Secondly, it was noted that two minor typographical errors needed correction. In the fifth paragraph on page 5 of 17, the SLMCPR at 95.3%
and 108% flow were updated to 1.1097 and 1.1029, respectively. The correction of these typographical errors has no impact on the results of the SLMCPR analysis. 2 to this letter provides Revision 2 of the SLMCPR report. Westinghouse Electric Company, LLC considers information provided in this report to be proprietary.
and 108% flow were updated to 1.1097 and 1.1029, respectively. The correction of these typographical errors has no impact on the results of the SLMCPR analysis. 2 to this letter provides Revision 2 of the SLMCPR report. Westinghouse Electric Company, LLC considers information provided in this report to be proprietary.
Therefore, an affidavit signed by Westinghouse, the owner of the information, is provided in Attachment 13. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, it is requested September 21, 2011 U. S. Nuclear Regulatory Commission Page 3 that the information in Attachment 12 be withheld from public disclosure in accordance with 10 CFR 2.390. A non
Therefore, an affidavit signed by Westinghouse, the owner of the information, is provided in Attachment 13. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, it is requested
-proprietary version of this report is provided in Attachment 14.EGC has reviewed the information supporting a finding of no significant hazards consideration that was provided to the NRC in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. No new regulatory commitments are established by this submittal.
 
September 21, 2011 U. S. Nuclear Regulatory Commission Page 3 that the information in Attachment 12 be withheld from public disclosure in accordance with 10 CFR 2.390. A non - proprietary version of this report is provided in Attachment 14.
EGC has reviewed the information supporting a finding of no significant hazards consideration that was provided to the NRC in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. No new regulatory commitments are established by this submittal.
If you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2804.
If you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2804.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 21St day of September 2011.
I declare under penalty of perjury that the foregoing is true and correct. Executed on the 21St day of September 2011.
Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC Attachments:
Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC Attachments:
1.Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit 2.Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Proprietary) 3.Westinghouse Affidavit for RAI 2 through 6 Responses 4.Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Non
: 1. Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit
-Proprietary) 5.EGC Procedure NF-AA-100, "Reload Control Procedure" 6.EGC Procedure NF-AB-1 10, "Bundle and Core Design (BWR)" 7.EGC Transmittal of Design Information document NF1000236 Rev. 1 8.EGC Calculation QDC-0000-N-1 804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" 9.Westinghouse Affidavit for Bundle Design Report and Reference Loading Pattern Report 10.Attachment to NF-BEX-1 0-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) 11.Attachment 1 to NF-BEX-1 0-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary) 12.Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Proprietary)
: 2. Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Proprietary)
September 21, 2011 U. S. Nuclear Regulatory Commission Page 4 13.Westinghouse Affidavit for Quad Cities Unit 1 Cycle 22 SLMCPR Report Revision 2 14.Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Non-Proprietary)
: 3. Westinghouse Affidavit for RAI 2 through 6 Responses
ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit In reviewing the Exelon Generation Company's (Exelon's) submittal dated June 7, 2011, related to the request to replace the current single recirculation loop operation (SLO) safety limit minimum critical power ratio (SLMCPR) requirement, for the Quad Cities Nuclear Power Station (QCNPS), Unit 1, the NRC staff has determined that the following information is needed in order to complete its review:
: 4. Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Non -
RAI-01 Provide: a.The details of the QCNPS, Unit 1, Cycle 22, analysis performed to obtain the final core loading pattern including procedure, guideline, criteria, and approved methodologies used for this analysis, andb.The design document for the QCNPS, Unit 1, Cycle 22, core loading pattern.
Proprietary)
Response to RAI-01 a The procedures and guidelines used for the QCNPS, Unit 1, Cycle 22 core loading pattern development are Exelon Generation Company, LLC (EGC) procedures NF-AA-100, "Reload Control Procedure" (Reference 1-1) and NF-AB-1 10, "Bundle and Core Design (BWR)," (Reference 1-2). EGC procedures documented in References 1-1 and 1-2 direct the bundle design and core reload process. The criteria are defined in the Cycle Design Inputs and Requirements (CDIR) document (Reference 1-3), which sets the design criteria for cycle energy, thermal margins, and other design constraints for the cycle.Following completion of the core loading pattern development, Westinghouse provides to EGC the Bundle Design Report (Reference 1-4) and the Reference Loading Pattern Report (Reference 1-5) which describe in detail the bundle design and loading pattern. In addition, Reference 1-5 includes an explanation of how the CDIR design requirements are satisfied.
: 5. EGC Procedure NF-AA-100, "Reload Control Procedure"
: 6. EGC Procedure NF-AB-1 10, "Bundle and Core Design (BWR)"
: 7. EGC Transmittal of Design Information document NF1000236 Rev. 1
: 8. EGC Calculation QDC-0000-N-1 804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan"
: 9. Westinghouse Affidavit for Bundle Design Report and Reference Loading Pattern Report
: 10. Attachment to NF-BEX-1 0-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary)
: 11. Attachment 1 to NF-BEX-1 0-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary)
: 12. Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Proprietary)
 
September 21, 2011 U. S. Nuclear Regulatory Commission Page 4
: 13. Westinghouse Affidavit for Quad Cities Unit 1 Cycle 22 SLMCPR Report Revision 2
: 14. Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Non-Proprietary)
 
ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit
 
ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit In reviewing the Exelon Generation Company's (Exelon's) submittal dated June 7, 2011, related to the request to replace the current single recirculation loop operation (SLO) safety limit minimum critical power ratio (SLMCPR) requirement, for the Quad Cities Nuclear Power Station (QCNPS), Unit 1, the NRC staff has determined that the following information is needed in order to complete its review:
 
===RAI-01===
Provide:
: a. The details of the QCNPS, Unit 1, Cycle 22, analysis performed to obtain the final core loading pattern including procedure, guideline, criteria, and approved methodologies used for this analysis, and
: b. The design document for the QCNPS, Unit 1, Cycle 22, core loading pattern.
Response to RAI-01 a The procedures and guidelines used for the QCNPS, Unit 1, Cycle 22 core loading pattern development are Exelon Generation Company, LLC (EGC) procedures NF-AA-100, "Reload Control Procedure" (Reference 1-1) and NF-AB-1 10, "Bundle and Core Design (BWR)," (Reference 1-2). EGC procedures documented in References 1-1 and 1-2 direct the bundle design and core reload process. The criteria are defined in the Cycle Design Inputs and Requirements (CDIR) document (Reference 1-3), which sets the design criteria for cycle energy, thermal margins, and other design constraints for the cycle. Following completion of the core loading pattern development, Westinghouse provides to EGC the Bundle Design Report (Reference 1-4) and the Reference Loading Pattern Report (Reference 1-5) which describe in detail the bundle design and loading pattern. In addition, Reference 1-5 includes an explanation of how the CDIR design requirements are satisfied.
The approved Westinghouse methodologies used to evaluate the core loading are documented in Reference 1-7. The results of implementing these methods are used to generate References 1-4 and 1-5.
The approved Westinghouse methodologies used to evaluate the core loading are documented in Reference 1-7. The results of implementing these methods are used to generate References 1-4 and 1-5.
EGC Procedures NF-AA-100, NF-AB-1 10, and the EGC Transmittal of Design Information document NF1000236 Rev. 1 (i.e., Q1C22 CDIR) are provided as Attachments 5, 6, and 7, respectively.
EGC Procedures NF-AA-100, NF-AB-1 10, and the EGC Transmittal of Design Information document NF1000236 Rev. 1 (i.e., Q1C22 CDIR) are provided as Attachments 5, 6, and 7, respectively.
Response to RAI-01 b Westinghouse documentation of the final core loading pattern is provided in References 1-4 and 1-5. The EGC documentation of the final as-loaded core loading pattern is documented in Calculation QDC-0000-N-1 804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" (Reference 1-6). Reference 1-6 documents the as-loaded Q1C22 core aswell as associated technical information about the core loading used for EGC internal activities.
Response to RAI-01 b Westinghouse documentation of the final core loading pattern is provided in References 1-4 and 1-5. The EGC documentation of the final as-loaded core loading pattern is documented in Calculation QDC-0000-N-1 804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" (Reference 1-6). Reference 1-6 documents the as-loaded Q1C22 core as well as associated technical information about the core loading used for EGC internal activities.
Page 1 of 2 ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Westinghouse proprietary References 1-4 and 1-5 are provided as Attachments 10 and 11 to this letter, respectively. EGC Calculation QDC-0000-N-1 804 is also provided as  to this letter.
Page 1 of 2
References RAI-01 1-1 NF-AA-100, "Reload Control Procedure," Revision 13 1-2 NF-AB-1 10, "Bundle and Core Design (BWR)," Revision 10 1-3 Cycle Design Inputs and Requirements for Quad Cities Unit 1, Cycle 22, dated November 30, 2010 1-4 Attachment to NF-BEX-10-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22," dated November 2010 1-5 Attachment 1 to NF-BEX-1 0-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22," dated December 2010 1-6 QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" 1-7 CENPD-300-P-A, Revision 0, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996 Page 2 of 2 ATTACHMENT 3 Westinghouse Affidavit for RAI 2 through 6 Responses Westinghouse Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory CommissionDirect tel:
 
(412) 374-4643 Document Control DeskDirect fax:
ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Westinghouse proprietary References 1-4 and 1-5 are provided as Attachments 10 and 11 to this letter, respectively. EGC Calculation QDC-0000-N-1 804 is also provided as  to this letter.
(724) 720-0754 11555 Rockville Pikee-mail: greshaja@westinghouse.com Rockville, MD 20852Proj letter:
References RAI-01 1-1   NF-AA-100, "Reload Control Procedure," Revision 13 1-2   NF-AB-1 10, "Bundle and Core Design (BWR)," Revision 10 1-3   Cycle Design Inputs and Requirements for Quad Cities Unit 1, Cycle 22, dated November 30, 2010 1-4   Attachment to NF-BEX-10-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22," dated November 2010 1-5 Attachment 1 to NF-BEX-1 0-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22," dated December 2010 1-6 QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" 1-7   CENPD-300-P-A, Revision 0, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996 Page 2 of 2
NF-BEX-11-142 CAW-1 1-3247 September 15, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
 
ATTACHMENT 3 Westinghouse Affidavit for RAI 2 through 6 Responses
 
Westinghouse                                                       Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission                                  Direct tel: (412) 374-4643 Document Control Desk                                              Direct fax: (724) 720-0754 11555 Rockville Pike                                                  e-mail: greshaja @ westinghouse.com Rockville, MD 20852                                                Proj letter: NF-BEX-11-142 CAW-1 1-3247 September 15, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE


==Subject:==
==Subject:==
USBWR-11-25-P-Attachment, "Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No.ME6383)" (Proprietary)
USBWR- 11-25 -P-Attachment, "Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383)" (Proprietary)
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3247 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3247 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3247, and should be addressed to J.A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3247, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Very truly yours, J.A. Gresham, Manager Regulatory Compliance Enclosures CAW- 11-3247 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
Very truly yours, J. A. Gresham, Manager Regulatory Compliance Enclosures
COUNTY OF BUTLER: Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
 
A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 145 ^4 day of September 2011 Notary Public COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member, Pennsylvania Association of Notaries ss 2CAW-11-3247
CAW- 11-3247 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
{1)I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse, (2)I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit, (3)I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
ss COUNTY OF BUTLER:
(4)Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.(i)The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse, (ii)The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes
A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 145 ^4 day of September 2011 Notary Public COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member, Pennsylvania Association of Notaries
 
2                                      CAW-11-3247
{1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse, (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit, (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)     The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse, (ii)     The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a)      The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of


Westinghouse policy and provides the rational basis required.
3                                        CAW-11-3247 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a)The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of 3CAW-11-3247 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.(b)It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.(c)Its use 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 a similar product.(d)It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.(e)It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.(f)It contains patentable ideas, for which patent protection may be desirable.
(b)     It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
There are sound policy reasons behind the Westinghouse system which include the following: (a)The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.(b)It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.(c)Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(c)     Its use 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 a similar product.
4CAW-11-3247 (d)Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.(e)Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the
(d)     It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)     It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)     It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a)     The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)     It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c)     Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.


competition of those countries.(f)The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage, (iii)The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.(iv)The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.(v)The proprietary information sought to be withheld in this submittal is that which is appropriately marked in USBWR-11-25-P-Attachment, "Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383)" (Proprietary), for submittal to the Commission, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.
4                                    CAW-11-3247 (d)      Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
This information is part of that which will enable Westinghouse to: (a)Support Exelon's use of Westinghouse Fuel at Quad Cities.
(e)      Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
5CAW-11-3247 (b)Assist the customer to obtain license change.
(f)     The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage, (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.
Further this information has substantial commercial value as follows: (a)Westinghouse can use this information to further enhance their licensing position with their competitors.(b)The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)   The proprietary information sought to be withheld in this submittal is that which is appropriately marked in USBWR-11-25-P-Attachment, "Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383)" (Proprietary), for submittal to the Commission, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.
This information is part of that which will enable Westinghouse to:
(a)     Support Exelon's use of Westinghouse Fuel at Quad Cities.
 
5                                      CAW-11-3247 (b)     Assist the customer to obtain license change.
Further this information has substantial commercial value as follows:
(a)     Westinghouse can use this information to further enhance their licensing position with their competitors.
(b)     The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
 
applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
Further the deponent sayeth not.
Proprietary Information Notice Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.
 
Proprietary Information Notice Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant -specific review and approval.
In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).
In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).
Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding.
Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
 
ATTACHMENT 4Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Non-Proprietary)
ATTACHMENT 4 Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Non-Proprietary)
 
Westinghouse Non-Proprietary Class 3 USBWR 11-25-NP-Attachment Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383) (Non-Proprietary)
Westinghouse Non-Proprietary Class 3 USBWR 11-25-NP-Attachment Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383) (Non-Proprietary)
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066 J 2011 Westinghouse Electric Company LLC All Rights Reserved Page 2 of 10 USB WR-11-25-NP-Attachment RAI-02 Provide the fuel bundle critical power ratio distribution in the core for the limiting point in the cycle.
Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066 J 2011 Westinghouse Electric Company LLC All Rights Reserved
Response to RAI-02 Figure RAI-02-1 shows the fuel assembly minimum critical power ratio values for the most limiting point in the cycle according to Figure 4 of Attachment 3 of the Exelon submittal. The values are I a,c shown [
 
Page 3 of 10 USB WR-11-25-NP-Attachment Figure RAI-02 QCNPS, Unit 1, Cycle 22 - Minimum CPR distribution for the limiting point in the cycle a,c Page 4 of 10 US B W R- 11-2 5-NP-Attachment Figure RAI-02 QCNPS, Unit 1, Cycle 22 -
Page 2 of 10 USB WR-11-25-NP-Attachment
Minimum CPR distribution for the limiting point in the cycle a, c Page 5 of 10 USBWR-1 1-25-NP-Attachment RAI-03 Provide: a.Verification that the power to flow map bounds QCNPS, Unit 1, Cycle 22, operation including stability Option III features of scram region and controlled entry region for back up stability projection, and b.A list of approved methodologies used to perform the above stability calculations.
 
===RAI-02===
Provide the fuel bundle critical power ratio distribution in the core for the limiting point in the cycle.
Response to RAI-02 Figure RAI-02-1 shows the fuel assembly minimum critical power ratio values for the most limiting point in the cycle according to Figure 4 of Attachment 3 of the Exelon submittal. The values are a,c shown [                                                 I
 
Page 3 of 10 USB WR-11-25-NP-Attachment Figure RAI-02 QCNPS, Unit 1, Cycle 22 - Minimum CPR distribution for the limiting point in the cycle a,c
 
Page 4 of 10 US B W R- 11-2 5-NP-Attachment Figure RAI-02 QCNPS, Unit 1, Cycle 22 - Minimum CPR distribution for the limiting point in the cycle a, c
 
Page 5 of 10 USBWR-1 1-25-NP-Attachment
 
===RAI-03===
Provide:
: a. Verification that the power to flow map bounds QCNPS, Unit 1, Cycle 22, operation including stability Option III features of scram region and controlled entry region for back up stability projection, and
: b. A list of approved methodologies used to perform the above stability calculations.
Response to RAI-03a Figures RAI-03-1 and RAI-03-2 show the QCNPS, Unit 1, Cycle 22 BSP Stability results for nominal and reduced feedwater heating respectively. These figures show that the Cycle 22 stability results are bounded by the backup stability protection scram and controlled entry regions.
Response to RAI-03a Figures RAI-03-1 and RAI-03-2 show the QCNPS, Unit 1, Cycle 22 BSP Stability results for nominal and reduced feedwater heating respectively. These figures show that the Cycle 22 stability results are bounded by the backup stability protection scram and controlled entry regions.
Figure RAI-03-1 QCNPS, Unit1, Cycle 22 BSP Stability Results - Nominal FWT Band Page 6 of 10 U S B W R-11-25-NP-Attachment Figure RAI-03-2 QCNPS, Unit 1, Cycle 22 BSP Stability Results - Reduced FWT Band Response to RAI-03b The methodologies used to perform the Stability Analysis for QCNPS, Unit 1, Cycle 22 are included in the following references.
Figure RAI-03-1 QCNPS, Unit1, Cycle 22 BSP Stability Results - Nominal FWT Band
3-1 CENPD-295-P-A, "Thermal-Hydraulic Stability Methodology for Boiling Water Reactors", July 1996.
 
3-2 GE-NE-0000-0028-9741-R1, "Plant-Specific Regional Mode DIVOM Procedure Guideline", June 2005.
Page 6 of 10 U S B W R-11-25-NP-Attachment Figure RAI-03-2 QCNPS, Unit 1, Cycle 22 BSP Stability Results - Reduced FWT Band Response to RAI-03b The methodologies used to perform the Stability Analysis for QCNPS, Unit 1, Cycle 22 are included in the following references.
Page 7 of 10 U SB WR-11-25-NP-Attachment RAI-04 Provide the rationale why a 30.4 percent reload batch fraction for SVEA-96 Optimal fuel caused the proposed SLMCPR increment of 0.01 for SLO and no change for TLO for the proposed loading pattern in Figure 1 of Attachments 3 and 5 of the Exelon submittal.
3-1   CENPD-295-P-A, "Thermal-Hydraulic Stability Methodology for Boiling Water Reactors",
July 1996.
3-2   GE-NE-0000-0028-9741-R1, "Plant-Specific Regional Mode DIVOM Procedure Guideline",
June 2005.
 
Page 7 of 10 U SB WR-11-25-NP-Attachment
 
===RAI-04===
Provide the rationale why a 30.4 percent reload batch fraction for SVEA-96 Optimal fuel caused the proposed SLMCPR increment of 0.01 for SLO and no change for TLO for the proposed loading pattern in Figure 1 of Attachments 3 and 5 of the Exelon submittal.
Response to RAI-04 The batch change in the reload fraction for SVEA-96 Optimal did not cause the change in the result for SLMCPR for SLO and no change for TLO. The reason for the change in the result for SLMCPR is described in detail in the response to Question RAI-05c.
Response to RAI-04 The batch change in the reload fraction for SVEA-96 Optimal did not cause the change in the result for SLMCPR for SLO and no change for TLO. The reason for the change in the result for SLMCPR is described in detail in the response to Question RAI-05c.
Page 8 of 10 USB WR-11-25-NP-Attachment RAI-05 Provide: a.Or make available for audit, copy of the McSLAP computer code including theory and user's manual;b.A description of the relationship between the McSLAP computer code and CENPD-300-P-A; c.Details of two errors found in Westinghouse's McSLAP computer code including applicable portions of the methodology described in CENPD-300-P-A, and their impact on the SLMCPR calculations including those at exposure beyond 4000 MWd/MTU; and d.Or make available for audit, a flow chart consisting of approved methodologies used for this SLMCPR calculation with respect to their input to each sub-routine calculation.
 
Response to RAI-05a and RAI-05d Westinghouse will make available for audit in the Westinghouse Rockville office a copy of theMcSLAP computer code including the theory / user's manual. The documents contain a general overview flow chart used for SLMCPR calculation.
Page 8 of 10 USB WR-11-25-NP-Attachment
 
===RAI-05===
Provide:
: a. Or make available for audit, copy of the McSLAP computer code including theory and user's manual;
: b. A description of the relationship between the McSLAP computer code and CENPD-300-P-A;
: c. Details of two errors found in Westinghouse's McSLAP computer code including applicable portions of the methodology described in CENPD-300-P-A, and their impact on the SLMCPR calculations including those at exposure beyond 4000 MWd/MTU; and
: d. Or make available for audit, a flow chart consisting of approved methodologies used for this SLMCPR calculation with respect to their input to each sub-routine calculation.
Response to RAI-05a and RAI-05d Westinghouse will make available for audit in the Westinghouse Rockville office a copy of the McSLAP computer code including the theory / user's manual. The documents contain a general overview flow chart used for SLMCPR calculation.
If the NRC needs to perform code executions, Westinghouse needs to be informed in advance in order to coordinate sending personnel to the Rockville office.
If the NRC needs to perform code executions, Westinghouse needs to be informed in advance in order to coordinate sending personnel to the Rockville office.
Response to RAI-05b The McSLAP code uses a statistical approach for sensitivity and uncertainty analysis of MCPR using a Monte Carlo technique. The McSLAP code integrates the Monte Carlo approach described in Section 5.3.2 of Reference 5-1.
Response to RAI-05b The McSLAP code uses a statistical approach for sensitivity and uncertainty analysis of MCPR using a Monte Carlo technique. The McSLAP code integrates the Monte Carlo approach described in Section 5.3.2 of Reference 5-1.
Line 110: Line 187:
== Description:==
== Description:==


a,c Page 9 of 10 USB WR-11-25-NP-Attachment Ia,c McSLAP Issue 2
a,c
 
Page 9 of 10 USB WR-11-25-NP-Attachment Ia,c McSLAP Issue 2
 
== Description:==
 
I a,c McSLAP Issue 3


== Description:==
== Description:==


I a,c McSLAP Issue 3 Descri p tion: I Mc SLAP Issues Impact:
I a,c I
References RAI-05:
Mc SLAP Issues Impact:
5-1 CENPD-300-P-A, Revision 0, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996 I a,c J a,c Page 10 of 10 U SB WR-11-25-NP-Attachment RAI-06Describe any Part 21 issues relating to the fuel design applied to the QCNPS, Unit 1, Cycle 22, fuel assemblies. Also, with respect to the Part 21 issues, identify all affected factors, and quantify their impact on the parameters shown in Table 2 of Attachment 3.
J a,c References RAI-05:
5-1 CENPD-300-P-A, Revision 0, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996
 
Page 10 of 10 U SB WR-11-25-NP-Attachment
 
===RAI-06===
Describe any Part 21 issues relating to the fuel design applied to the QCNPS, Unit 1, Cycle 22, fuel assemblies. Also, with respect to the Part 21 issues, identify all affected factors, and quantify their impact on the parameters shown in Table 2 of Attachment 3.
Response to RAI-06 There are no Part 21 reportable issues related to the fuel design applied to the QCNPS, Unit 1, Cycle 22 fuel assemblies. With respect to the issue associated with this Technical Specification change request, it was determined not to be reportable tinder IOCFR21 due to the existence of offsetting conservatisms.
Response to RAI-06 There are no Part 21 reportable issues related to the fuel design applied to the QCNPS, Unit 1, Cycle 22 fuel assemblies. With respect to the issue associated with this Technical Specification change request, it was determined not to be reportable tinder IOCFR21 due to the existence of offsetting conservatisms.
ATTACHMENT 5 EGC Procedure NF-AA-100, "Reload Control Procedure" Exelon, Nuclear NF-AA-100 Revision 13 Page 1 of 54 Level 3 - Information Use RELOAD CONTROL PROCEDURE 1.PURPOSE 1.1.This procedure defines the steps to be followed for the effective specification, design, review, approval and implementation of reload fuel assemblies, core designs and related services (including core component management) for Exelon nuclear reactors. (Reference NF-AA-10) 1.2.This procedure defines the requirements for documenting Core Reload and Cycle Management configuration changes. There are T&RMs that detail the processing of Fuel Change Package (FCP) type Engineering Changes (EC - Passport) or Engineering Change Requests (ECR - PIMS) for Core Reload and Cycle Management configuration changes. (Reference NF-AA-100-1000) 1.3.This document is not a detailed specification of the reload procedures to be followed for each unit. That detail is provided by other corporate and site specific procedures and T&
 
RMs. Nor does this procedure repeat the detail of applicable general topics such as: task control; documentation control; engineering reviews; etc., which are the subject of other procedures.
ATTACHMENT 5 EGC Procedure NF-AA-100, "Reload Control Procedure"
1.4.This procedure is also applicable for mid-cycle outages that implement core re-designs (e.g. replacement of failed fuel). Although not all steps within this procedure may be required to be executed, each step should be evaluated for applicability.
 
Those steps that assure safe, economic and reliable plant operation and those steps that ensure communication of the redesign details and their impact on subsequent plant operations are essential.
Exelon,                                                                       NF-AA-100 Revision 13 Page 1 of 54 Nuclear                                        Level 3 - Information Use RELOAD CONTROL PROCEDURE
2.TERMS AND DEFINITIONS 2.1.50.59 Review -a review performed in accordance with 1 OCFR50.59 to determine if a proposed change, test or experiment requires prior NRC approval via license amendment under 1 OCFR50.90. (Reference LS-AA-1 04) 2.2.Advance Work Authorization (AWA) - authorization to proceed at risk with installation work activities in the field without issuance of a completely approved configuration change. The advance work does not affect any in-service equipment and the equipment is not placed in service and is not declared operable, and has no impact on operating or in-service equipment until the configuration change is approved.2.3.Affected Document List (ADL) - an electronic database list that identifies the controlled documents affected by configuration change packages.
: 1. PURPOSE 1.1. This procedure defines the steps to be followed for the effective specification, design, review, approval and implementation of reload fuel assemblies, core designs and related services (including core component management) for Exelon nuclear reactors. (Reference NF-AA-10) 1.2. This procedure defines the requirements for documenting Core Reload and Cycle Management configuration changes. There are T&RMs that detail the processing of Fuel Change Package (FCP) type Engineering Changes (EC - Passport) or Engineering Change Requests (ECR - PIMS) for Core Reload and Cycle Management configuration changes. (Reference NF-AA-100-1000) 1.3. This document is not a detailed specification of the reload procedures to be followed for each unit. That detail is provided by other corporate and site specific procedures and T& RMs. Nor does this procedure repeat the detail of applicable general topics such as: task control; documentation control; engineering reviews; etc., which are the subject of other procedures.
2.4.Beginning of Cycle (BOC) - the timeframe when an operating fuel cycle is initiated.
1.4. This procedure is also applicable for mid-cycle outages that implement core re-designs (e.g. replacement of failed fuel). Although not all steps within this procedure may be required to be executed, each step should be evaluated for applicability. Those steps that assure safe, economic and reliable plant operation and those steps that ensure communication of the redesign details and their impact on subsequent plant operations are essential.
NF-AA-100 Revision 13 Page 2 of 542.5.Compensatory Actions - a real commitment of effort or material to reduce the probability or consequences of a risk; part of a Risk Management Assessment.2.6.Configuration Change a change to the design basis, design documentation, or physical plant configuration2.7.Contingency Plans -the action in response to a "risk trigger" (without necessarily committing to the actual effort or material to execute that plan); part of a Risk Management Assessment.2.8.Core Components - components that are not intimately incorporated into fuel assemblies. They are either not located in the fuel assembly or could be re-used many times in different fuel assemblies. Therefore their procurement/
: 2. TERMS AND DEFINITIONS 2.1. 50. 59 Review - a review performed in accordance with 1 OCFR50.59 to determine if a proposed change, test or experiment requires prior NRC approval via license amendment under 1 OCFR50.90. (Reference LS-AA-1 04) 2.2. Advance Work Authorization (AWA) - authorization to proceed at risk with installation work activities in the field without issuance of a completely approved configuration change. The advance work does not affect any in-service equipment and the equipment is not placed in service and is not declared operable, and has no impact on operating or in-service equipment until the configuration change is approved.
use is not intimately linked to the reload fuel. Examples of core components are: (PWR):
2.3. Affected Document List (ADL) - an electronic database list that identifies the controlled documents affected by configuration change packages.
2.4. Beginning of Cycle (BOC) - the timeframe when an operating fuel cycle is initiated.
 
NF-AA-100 Revision 13 Page 2 of 54 2.5. Compensatory Actions - a real commitment of effort or material to reduce the probability or consequences of a risk; part of a Risk Management Assessment.
2.6. Configuration Change a change to the design basis, design documentation, or physical plant configuration 2.7. Contingency Plans - the action in response to a "risk trigger" (without necessarily committing to the actual effort or material to execute that plan); part of a Risk Management Assessment.
2.8. Core Components - components that are not intimately incorporated into fuel assemblies. They are either not located in the fuel assembly or could be re-used many times in different fuel assemblies. Therefore their procurement/ use is not intimately linked to the reload fuel. Examples of core components are: (PWR):
control rods; thimble plug assemblies; primary sources; secondary sources (BWR):
control rods; thimble plug assemblies; primary sources; secondary sources (BWR):
control blades.2.9.Core Loading Plan CLP) -
control blades.
a document that contains the full core layout that indicates core-loading positions of all reload assemblies/bundles using the vendor-supplied identifiers and supporting information.2.10.Core O p eratin g Limits Re portort Ct7LR -a document that contains the core thermal and reactivity limits required by Technical Specifications for an operating fuel cycle.2.11.Core Reload -the process of performing the engineering work necessary to refuel a reactor core. This work includes the specification, calculations, documentation, reviews, testing, and any updates to computer codes, databases, Technical Specifications, UFSAR, procedures and training necessary to operate the fuel cycle under design.2.12.Cycle Design Inputs and Requirements (CDIR) - a formal document(s) containing cycle design targets and the inputs and acceptable outputs that define the boundaries of the cycle design and licensing analyses. The document(s) may be identified by vendor-specific naming conventions (see Attachment 2).2.13.Cycle Management -
2.9. Core Loading Plan CLP) - a document that contains the full core layout that indicates core-loading positions of all reload assemblies/bundles using the vendor-supplied identifiers and supporting information.
the process of performing the engineering work necessary to implement changes to the cycle specific configurations. This work includes the specification, calculations, documentation, reviews, testing, and any updates to computer codes, databases, Technical Specifications, UFSAR, procedures and training necessary to implement changes to the current operating fuel cycle design.2.14.Design Technical Review (DTR) - a technical review of the PFCD and Risk Management Assessment performed by the Reload Design Overview Team.2.15.End of Cycle (EOC) - the timeframe when an operating fuel cycle is concluded.
2.10. Core O peratin g Limits Re portort Ct7LR - a document that contains the core thermal and reactivity limits required by Technical Specifications for an operating fuel cycle.
NF-AA-1 00 Revision 13 Page 3 of 542.16.End of Rated (EOR) - the cycle exposure at which there is insufficient reactivity to achieve rated thermal power without using cycle extension maneuvers.
2.11. Core Reload - the process of performing the engineering work necessary to refuel a reactor core. This work includes the specification, calculations, documentation, reviews, testing, and any updates to computer codes, databases, Technical Specifications, UFSAR, procedures and training necessary to operate the fuel cycle under design.
2.17.Energy Utilization Plan (EUP) -
2.12. Cycle Design Inputs and Requirements (CDIR) - a formal document(s) containing cycle design targets and the inputs and acceptable outputs that define the boundaries of the cycle design and licensing analyses. The document(s) may be identified by vendor-specific naming conventions (see Attachment 2).
the approved schedule of core thermal energy requirements (i.e., cycle length, power level, effective full power days, and EOC extension options such as coastdown) and outage dates input to the reload design.2.18.Final Fuel Cycle Design (FFCD) - the approved reload design that will be used for an operating fuel cycle.2.19.Fuel - the assemblies (PWR) or bundles (BWR) of fuel rods assembled in a retaining structure. In this document, references to reload fuel should be taken to encompass both the reload-specific fuel and fuel related components.2.20.Fuel Change Package (FCP) - a type of EC/ECR used to document, approve and implement core reload and cycle management activities.2.21.Fuel Related Components - components that are intimately related to a single fuel assembly, i.e., their entire in-core lifetime is usually associated with installation in only one fuel assembly. Therefore their procurement/use is intimately linked to the
2.13. Cycle Management - the process of performing the engineering work necessary to implement changes to the cycle specific configurations. This work includes the specification, calculations, documentation, reviews, testing, and any updates to computer codes, databases, Technical Specifications, UFSAR, procedures and training necessary to implement changes to the current operating fuel cycle design.
 
2.14. Design Technical Review (DTR) - a technical review of the PFCD and Risk Management Assessment performed by the Reload Design Overview Team.
reload fuel. Examples of fuel related components are (PWR): removable burnable poison rod assemblies (various designs such as Removable Burnable Poison Assemblies (RBPAs), Burnable Poison Rod Assemblies (BPRAs), Vibration Suppressors, or Wet Annular Burnable Absorbers (WABAs)); (BWR): channels, channel fasteners. Design and process changes to Fuel Related Components are evaluated as changes to Fuel Assemblies.2.22.Lead Test Assembly (LTA) - a fuel assembly or core component with no preexisting in-reactor operating experience.2.23.Lead Use Assembly (LUA) - a fuel assembly or core component with pre-existing in-reactor experience but may be its first application at an Exelon reactor.2.24.Like-for-Like Replacement I Identical Replacement Item - the replacement item is considered like-for-like if there is a high level of confidence that there have been no changes in the design, material, or manufacturing process since the procurement of the item being replaced.2.25.Monitoring Plan - a compensatory action to actively look for Risk Triggers and initiate contingency plans.2.26.Nuclear Fuels (NF) - the Exelon Nuclear department responsible for fuel related engineering activities.2.27.Nuclear Safety Review Board (NSRB) -
2.15. End of Cycle (EOC) - the timeframe when an operating fuel cycle is concluded.
a committee that oversees the management and operating practices of Exelon Nuclear.
NF-AA-100 Revision 13 Page 4 of 542.28.Operational Experience (OE) - nuclear industry information based on plant performance issues, lessons learned, regulatory reports, internal Exelon reports or via industry information exchange forums, such as OPEX from INPO.2.29.Passport Engineering Change
/PIMS Engineering Change Request (EC/ECR) -
approved design packages that govern configuration changes to the plants.2.30.Plant Operations Review Committee (PORC) - a multi-disciplined committee responsible for review of activities that have potential to affect nuclear safety.2.31.Preliminary Fuel Cycle Design (PFCD) -
the preliminary reload design that is presented to the RDOT for a design technical review and to the RRB for NF management approval prior to finalizing the reload design for licensing.2.32.Re-insert Fuel -
fuel used in the reload cycle that has been previously irradiated and was stored in the spent fuel pool during the previous cycle.2.33.Re-use Fuel -fuel used in the reload cycle that was operating in the reactor core in the previous cycle.2.34.Reload-Desi g n Review Criteria -
Criteria relating to Core Reload design and operating strategy changes and their impacts on: safety, operations, performance, monitoring, modeling, training, procedures and licensing, against which all reload designs must be reviewed. These criteria are generally vendor/unit specific and are specified in lower-level procedures and/or guidelines (see Attachment 3).2.35.Risk Management Assessment -
the end deliverable from the Risk Management Process: a document that identifies key risk management parameters and associated actions.2.36.Risk Trigger - A deviation from a nominal range of acceptable values that will activate a contingency plan.2.37.Senior Management Design Initialization (SMDI) - a meeting held early in the Reload Design process to ensure that senior site management actively participate in the decision making process for significant changes in the reload design, core operating strategy, fuel mechanical design, and fuel cost associated with the reload.2.38.Updated Final Safety Analysis Report (UFSAR) -a document that contains all the changes necessary to reflect information and analyses submitted to the NRC per 10 CFR 50.71(e) since the submission of the original FSAR. UFSAR description includes text, tables, diagrams, etc., that provide an understanding of the design bases, safety analyses and facility operation under conditions of normal operation, anticipated operational occurrences, design basis accidents, external events, and


natural phenomena for which the plant is designed to function.
NF-AA-1 00 Revision 13 Page 3 of 54 2.16. End of Rated (EOR) - the cycle exposure at which there is insufficient reactivity to achieve rated thermal power without using cycle extension maneuvers.
NF-AA-100 Revision 13 Page 5 of 54 3.RESPONSIBILITIES NOTE: Other personnel may perform activities allocated to the Core Designer within this procedure as determined by lower-level procedures or as appointed by the applicable NF manager.
2.17. Energy Utilization Plan (EUP) - the approved schedule of core thermal energy requirements (i.e., cycle length, power level, effective full power days, and EOC extension options such as coastdown) and outage dates input to the reload design.
However, the Core Designer retains responsibility for overall cognizance and coordination of the reload process within NF.
2.18. Final Fuel Cycle Design (FFCD) - the approved reload design that will be used for an operating fuel cycle.
3.1.Chemistry RDOT Member - the individual in the site Chemistry department allocated responsibility for ensuring that the historical and projected chemistry parameters used as input into the RDOT are correct, all planned chemistry changes over the time period reviewed by the RDOT are accurately described and the possible impacts are reviewed, and relevant chemistry industry operating experience is appropriately incorporated into the RDOT review.
2.19. Fuel - the assemblies (PWR) or bundles (BWR) of fuel rods assembled in a retaining structure. In this document, references to reload fuel should be taken to encompass both the reload-specific fuel and fuel related components.
3.2.Core Designer - the individuals in the NF organization allocated responsibility for the bundle/core design(s) and oversight of the reload design process. This individual is
2.20. Fuel Change Package (FCP) - a type of EC/ECR used to document, approve and implement core reload and cycle management activities.
2.21. Fuel Related Components - components that are intimately related to a single fuel assembly, i.e., their entire in-core lifetime is usually associated with installation in only one fuel assembly. Therefore their procurement/use is intimately linked to the reload fuel. Examples of fuel related components are (PWR): removable burnable poison rod assemblies (various designs such as Removable Burnable Poison Assemblies (RBPAs), Burnable Poison Rod Assemblies (BPRAs), Vibration Suppressors, or Wet Annular Burnable Absorbers (WABAs)); (BWR): channels, channel fasteners. Design and process changes to Fuel Related Components are evaluated as changes to Fuel Assemblies.
2.22. Lead Test Assembly (LTA) - a fuel assembly or core component with no preexisting in-reactor operating experience.
2.23. Lead Use Assembly (LUA) - a fuel assembly or core component with pre-existing in-reactor experience but may be its first application at an Exelon reactor.
2.24. Like-for-Like Replacement I Identical Replacement Item - the replacement item is considered like-for-like if there is a high level of confidence that there have been no changes in the design, material, or manufacturing process since the procurement of the item being replaced.
2.25. Monitoring Plan - a compensatory action to actively look for Risk Triggers and initiate contingency plans.
2.26. Nuclear Fuels (NF) - the Exelon Nuclear department responsible for fuel related engineering activities.
2.27. Nuclear Safety Review Board (NSRB) - a committee that oversees the management and operating practices of Exelon Nuclear.


responsible for ensuring the Core Reload and Cycle Management activities are documented on the appropriate FCP EC/ECR. This individual communicates Core Reload and Cycle Management related issues to management, appropriate site organizations, fuel vendor(s) and core component vendor(s).
NF-AA-100 Revision 13 Page 4 of 54 2.28. Operational Experience (OE) - nuclear industry information based on plant performance issues, lessons learned, regulatory reports, internal Exelon reports or via industry information exchange forums, such as OPEX from INPO.
3.3.Director of NF - Provides authorization of AWAs that affect the functional design, design criteria of the plant or the 1 OCFR50.59 Evaluation, if applicable.
2.29. Passport Engineering Change / PIMS Engineering Change Request (EC/ECR) -
3.4.EC Coordinator (Passport only) - performs administrative review of the FCP.
approved design packages that govern configuration changes to the plants.
2.30. Plant Operations Review Committee (PORC) - a multi-disciplined committee responsible for review of activities that have potential to affect nuclear safety.
2.31. Preliminary Fuel Cycle Design (PFCD) - the preliminary reload design that is presented to the RDOT for a design technical review and to the RRB for NF management approval prior to finalizing the reload design for licensing.
2.32. Re-insert Fuel - fuel used in the reload cycle that has been previously irradiated and was stored in the spent fuel pool during the previous cycle.
2.33. Re-use Fuel - fuel used in the reload cycle that was operating in the reactor core in the previous cycle.
2.34. Reload- Desi g n Review Criteria - Criteria relating to Core Reload design and operating strategy changes and their impacts on: safety, operations, performance, monitoring, modeling, training, procedures and licensing, against which all reload designs must be reviewed. These criteria are generally vendor/unit specific and are specified in lower-level procedures and/or guidelines (see Attachment 3).
2.35. Risk Management Assessment - the end deliverable from the Risk Management Process: a document that identifies key risk management parameters and associated actions.
2.36. Risk Trigger - A deviation from a nominal range of acceptable values that will activate a contingency plan.
2.37. Senior Management Design Initialization (SMDI) - a meeting held early in the Reload Design process to ensure that senior site management actively participate in the decision making process for significant changes in the reload design, core operating strategy, fuel mechanical design, and fuel cost associated with the reload.
2.38. Updated Final Safety Analysis Report (UFSAR) - a document that contains all the changes necessary to reflect information and analyses submitted to the NRC per 10 CFR 50.71(e) since the submission of the original FSAR. UFSAR description includes text, tables, diagrams, etc., that provide an understanding of the design bases, safety analyses and facility operation under conditions of normal operation, anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant is designed to function.


NF-AA-100 Revision 13 Page 5 of 54
: 3. RESPONSIBILITIES NOTE:      Other personnel may perform activities allocated to the Core Designer within this procedure as determined by lower-level procedures or as appointed by the applicable NF manager.          However, the Core Designer retains responsibility for overall cognizance and coordination of the reload process within NF.
3.1. Chemistry RDOT Member - the individual in the site Chemistry department allocated responsibility for ensuring that the historical and projected chemistry parameters used as input into the RDOT are correct, all planned chemistry changes over the time period reviewed by the RDOT are accurately described and the possible impacts are reviewed, and relevant chemistry industry operating experience is appropriately incorporated into the RDOT review.
3.2. Core Designer - the individuals in the NF organization allocated responsibility for the bundle/core design(s) and oversight of the reload design process. This individual is responsible for ensuring the Core Reload and Cycle Management activities are documented on the appropriate FCP EC/ECR. This individual communicates Core Reload and Cycle Management related issues to management, appropriate site organizations, fuel vendor(s) and core component vendor(s).
3.3. Director of NF - Provides authorization of AWAs that affect the functional design, design criteria of the plant or the 1 OCFR50.59 Evaluation, if applicable.
3.4. EC Coordinator (Passport only) - performs administrative review of the FCP.
Ensures all documents and attachments are in the FCP package. Ensures that the ADL and the install attributes are correctly populated.
Ensures all documents and attachments are in the FCP package. Ensures that the ADL and the install attributes are correctly populated.
3.5.Fuel Buyer - procures nuclear fuel and fuel related components.
3.5. Fuel Buyer - procures nuclear fuel and fuel related components.
3.6.Fuel Reliability Engineer (FRE) - initiates detailed reviews of new or changed fuel assemblies and core component design and fabrication process changes.
3.6. Fuel Reliability Engineer (FRE) - initiates detailed reviews of new or changed fuel assemblies and core component design and fabrication process changes.
Responsible for determining the impact of chemistry changes identified at RDOT meetings on fuel performance.
Responsible for determining the impact of chemistry changes identified at RDOT meetings on fuel performance.
3.7.Manager, BWR/PWR Design - a manager within NF with responsibility for Core Reload and Cycle Management support to the unit under consideration. This individual assigns the Core Designer, and any delegates, as well as necessary reviewers and verifiers. Provides authorization for AWAs that do not affect the functional design, design criteria of the plant or the 1 OCFR50.59 Evaluation, if applicable.
3.7. Manager, BWR/PWR Design - a manager within NF with responsibility for Core Reload and Cycle Management support to the unit under consideration. This individual assigns the Core Designer, and any delegates, as well as necessary reviewers and verifiers. Provides authorization for AWAs that do not affect the functional design, design criteria of the plant or the 1 OCFR50.59 Evaluation, if applicable.
NF-AA-100 Revision 13 Page 6 of 543.8.NF Data Bank (NFDB) Analyst - updates the NFDB for each site based on fuel assemblies discharged and fresh fuel assemblies introduced by the reload3.9.NF Engineer for Fuel Fabrication Oversight - coordinates the monitoring of the nuclear fuel and core component fabrication process.3.10.NF Vice President - The NF VP has final approval authority for the EUP, fuel design changes, and the final fuel cycle design. Chairs the Reload Review Board.3.11.Nuclear Oversight Vendor Auditor - conducts assessments and surveillances of vendors, as appropriate, to evaluate design, licensing, and fabrication activities for fuel and other related components and services.3.12.Outage Services -- the organization that receives and installs nuclear fuel and core components.3.13.Reactor Engineer (RE) - the individuals in the site organization who are the primary interface with the Core Designer during the Core Reload and Cycle Management activities.3.14.Reload Design Overview Team (RDOT) - a multi-discipline team made up of NF and site personnel that provides planning, input, and review of reload activities. It assures both that appropriate boundaries for the reload design are established, and that any changes to established boundaries required by the design are acceptable (see Attachment 1).3.15.Reload Review Board (RRB) - an internal NF board that provides management challenge, oversight and review of all Core Reload Designs. The RRB oversees the reload strategy, fuel cycle economics and business issues, impacts of design decisions, and communication of reload plans outside of NF. The RRB meets as required to assure that each reload receives the appropriate level of oversight.
 
Various steps of this procedure require the NF VP to obtain RRB concurrence of the approach to be adopted (see Attachment 4)3.16.Site Vice President -
NF-AA-100 Revision 13 Page 6 of 54 3.8. NF Data Bank (NFDB) Analyst - updates the NFDB for each site based on fuel assemblies discharged and fresh fuel assemblies introduced by the reload 3.9. NF Engineer for Fuel Fabrication Oversight - coordinates the monitoring of the nuclear fuel and core component fabrication process.
The Site VP provides final site concurrence for core reloads and for significant fuel design changes.3.17.Supp l y Mana g ement Bu y er - procures core components.
3.10. NF Vice President - The NF VP has final approval authority for the EUP, fuel design changes, and the final fuel cycle design. Chairs the Reload Review Board.
NF-AA-100 Revision 13 Page 7 of 54 4.MAIN BODY NOTE: Tasks in this procedure related to FCP preparation and independent review by NF personnel are associated with Certification Guide N-AN-ENG-CERT-NF05. Certification Guide requirements for other tasks are defined in the applicable governing procedures/T&RMs, as necessary.
3.11. Nuclear Oversight Vendor Auditor - conducts assessments and surveillances of vendors, as appropriate, to evaluate design, licensing, and fabrication activities for fuel and other related components and services.
NOTE: The procedure steps in Section 4 include many tasks that may be performed concurrently or in a different sequence.
3.12. Outage Services -- the organization that receives and installs nuclear fuel and core components.
3.13. Reactor Engineer (RE) - the individuals in the site organization who are the primary interface with the Core Designer during the Core Reload and Cycle Management activities.
3.14. Reload Design Overview Team (RDOT) - a multi-discipline team made up of NF and site personnel that provides planning, input, and review of reload activities. It assures both that appropriate boundaries for the reload design are established, and that any changes to established boundaries required by the design are acceptable (see Attachment 1).
3.15. Reload Review Board (RRB) - an internal NF board that provides management challenge, oversight and review of all Core Reload Designs. The RRB oversees the reload strategy, fuel cycle economics and business issues, impacts of design decisions, and communication of reload plans outside of NF. The RRB meets as required to assure that each reload receives the appropriate level of oversight.
Various steps of this procedure require the NF VP to obtain RRB concurrence of the approach to be adopted (see Attachment 4) 3.16. Site Vice President - The Site VP provides final site concurrence for core reloads and for significant fuel design changes.
3.17. Supp ly Management Bu yer - procures core components.
 
NF-AA-100 Revision 13 Page 7 of 54
: 4. MAIN BODY NOTE:     Tasks in this procedure related to FCP preparation and independent review by NF personnel are associated with Certification Guide N-AN-ENG-CERT-NF05. Certification Guide requirements for other tasks are defined in the applicable governing procedures/T&RMs, as necessary.
NOTE:     The procedure steps in Section 4 include many tasks that may be performed concurrently or in a different sequence.
Essential sequencing is explicitly stated where necessary.
Essential sequencing is explicitly stated where necessary.
Detailed schedules are specified in vendor-specific guidelines.
Detailed schedules are specified in vendor-specific guidelines. Refer to NF-AB-100-4000, NF-AB-100-5000, NF-AB-100-6000, NF-AP-100-7000 and NF-AP-100-8000.
Refer to NF-AB-100-4000, NF-AB-100-5000, NF-AB-100-6000, NF-AP-100-7000 and NF-AP-100-8000.
NOTE:       Steps 4.1 and 4.2 deal with long lead-time items that are typically addressed well in advance of the bulk of the reload process.
NOTE: Steps 4.1 and 4.2 deal with long lead-time items that are typically addressed well in advance of the bulk of the reload process.
4.1. Core Component Management and Replacement 4.1.1. TRACK and PROJECT the exposure and/or lifetime history of core components.
4.1.Core Component Management and Replacement 4.1.1.TRACK and PROJECT the exposure and/or lifetime history of core components.
This responsibility is typically divided between NF and Plant / Reactor Engineering as follows:
This responsibility is typically divided between NF and Plant / Reactor Engineering as follows:
BWR control blades; PWR in-core detectors, neutron sources, and control rod assemblies. (Core Designer or Plant
BWR control blades; PWR in-core detectors, neutron sources, and control rod assemblies. (Core Designer or Plant / Reactor Engineering)
/Reactor Engineering)
BWR Local Power Range Monitors (LPRM), Wide Range Neutron Monitors (WRNM), and core support plate plugs. (Plant / Reactor Engineering) 4.1.2. PERFORM or COMMISSION any technical analysis required to determine which core components need replacement for the reload cycle, due to lifetime issues or other constraints. (Core Designer or Plant / Reactor Engineering) 4.1.3. DEVELOP replacement and core shuffling strategies for the above core components. (Core Designer or Plant / Reactor Engineering)
BWR Local Power Range Monitors (LPRM), Wide Range Neutron Monitors (WRNM), and core support plate plugs.(Plant/Reactor Engineering) 4.1.2.PERFORM or COMMISSION any technical analysis required to determine which core components need replacement for the reload cycle, due to lifetime issues or other constraints.(Core Designer or Plant
: 1.     REVIEW any changes in design attributes of the core components.
/Reactor Engineering) 4.1.3.DEVELOP replacement and core shuffling strategies for the above core components.(Core Designer or Plant
: 2.     NOTIFY the Core Designer of any design changes in components that will impact the reload cycle design inputs or requirements.
/Reactor Engineering) 1.REVIEW any changes in design attributes of the core components.
: 3.     PROVIDE replacement requirements to the site organizations responsible for budgeting the items and preparing purchase requisitions.
2.NOTIFY the Core Designer of any design changes in components that will impact the reload cycle design inputs or requirements.
: 4.     ENSURE any core components listed in 4.1.1 that are to be replaced in the upcoming refueling outage are included in the outage work scope. Refer to OU-AA-101.
3.PROVIDE replacement requirements to the site organizations responsible for budgeting the items and preparing purchase requisitions.
 
4.ENSURE any core components listed in 4.1.1 that are to be replaced in the upcoming refueling outage are included in the outage work scope. Refer to OU-AA-101.
NF-AA-100 Revision 13 Page 8 of 54 4.1.4. PREPARE configuration change packages for core component changes. Refer to CC-AA-1 03 and NF-AA-1 01. (FRE or designated individual) 4.1.5. PROCURE any necessary core components from approved suppliers. (Supply Management Buyer)
NF-AA-100 Revision 13 Page 8 of 54 4.1.4.PREPARE configuration change packages for core component changes. Refer to CC-AA-1 03 and NF-AA-1 01.(FRE or designated individual) 4.1.5.PROCURE any necessary core components from approved suppliers.(Supply Management Buyer) 1.REVIEW current fuel and/or core component contract(s) for applicability. A unique contract (including a bid specification, bid evaluation, contract award process, etc.) may be required.
: 1. REVIEW current fuel and/or core component contract(s) for applicability. A unique contract (including a bid specification, bid evaluation, contract award process, etc.) may be required.
4.1.6.INSTALL the core components and VERIFY their adequate operation prior to reload cycle operation.(Outage Services or Site Maintenance) 4.2.Fuel Design Selection NOTE: The following review of fuel designs and fuel related component designs pertains to fuel product lines or hardware options that are mechanically different (e.g., GE14 to GNF2, advanced cladding material, channel material, mixing vanes, etc.) from that used previously for the station in question.
4.1.6. INSTALL the core components and VERIFY their adequate operation prior to reload cycle operation. (Outage Services or Site Maintenance) 4.2. Fuel Design Selection NOTE:     The following review of fuel designs and fuel related component designs pertains to fuel product lines or hardware options that are mechanically different (e.g.,
The review of neutronically different bundle designs is considered part of the Core Reload design process.
GE14 to GNF2, advanced cladding material, channel material, mixing vanes, etc.) from that used previously for the station in question.         The review of neutronically different bundle designs is considered part of the Core Reload design process.
NOTE: For significant fuel design changes, a Special SMDI meeting should be conducted at a time early enough to accommodate changes in strategy.
NOTE:     For significant fuel design changes, a Special SMDI meeting should be conducted at a time early enough to accommodate changes in strategy.
4.2.1.REVIEW available fuel designs and fuel related component designs (e.g., channels, WABAs, BPRAs) to identify designs that have potential to be used in the reload.(Core Designer and FRE)1CONSIDER the following items when reviewing fuel designs:
4.2.1. REVIEW available fuel designs and fuel related component designs (e.g., channels, WABAs, BPRAs) to identify designs that have potential to be used in the reload.
(Core Designer and FRE) 1      CONSIDER the following items when reviewing fuel designs:
suitability of the current fuel design product lines available under current fuel vendor contracts product lines available from other fuel vendors degree of confidence in and experience with available products changes necessitated by Fuel Reliability or operating experience (OE) issues changes necessary to achieve unit objectives such as the EUP, enhanced economics, performance goals, or operational goals applicability of current in-core and out-of-core methods (CM-1)
suitability of the current fuel design product lines available under current fuel vendor contracts product lines available from other fuel vendors degree of confidence in and experience with available products changes necessitated by Fuel Reliability or operating experience (OE) issues changes necessary to achieve unit objectives such as the EUP, enhanced economics, performance goals, or operational goals applicability of current in-core and out-of-core methods (CM-1)
NF-AA-100 Revision 13 Page 9 of 54 significance of any design changes relative to the current fuel design mechanical/nuclear compatibility with dry storage/transport system deployed at the site, as applicable4.2.2.RECOMMEND the fuel design to be used for the Core Reload design.(Manager, BWR/PWR Design) 1.SUPPORT the recommendation with initial evaluations relating to the items listed in 4.2.1.1.
 
Detailed evaluations that are required prior to fuel design implementation may be completed later (see Step 4.2.4).
NF-AA-100 Revision 13 Page 9 of 54 significance of any design changes relative to the current fuel design mechanical/nuclear compatibility with dry storage/transport system deployed at the site, as applicable 4.2.2. RECOMMEND the fuel design to be used for the Core Reload design. (Manager, BWR/PWR Design)
2.RECOMMEND taking part in a new fuel design evaluation or test loading at this stage, if applicable.4.2.3.CLASSIFY the fuel design selected for the Core Reload. Refer to NF-AA-101.(FRE)1.If the fuel design selected is "Like-for-Like" to the fuel design currently in use at the unit, then DOCUMENT this in the Core Reload FCP and SKIP to Step 4.3.Otherwise PROCEED to Step 4.2.4.4.2.4.PREPARE a configuration change package for the new or changed fuel design.
: 1. SUPPORT the recommendation with initial evaluations relating to the items listed in 4.2.1.1. Detailed evaluations that are required prior to fuel design implementation may be completed later (see Step 4.2.4).
Refer to CC-AA-103 and NF-AA-101.(FRE or designated individual) 1.IDENTIFY all new supporting reviews or analyses that are needed based on the results of the NF-AA-1 01 review.
: 2. RECOMMEND taking part in a new fuel design evaluation or test loading at this stage, if applicable.
2.ENSURE that Core Reload design inputs or requirements that are impacted by fuel design changes are addressed in the CDIR (Step 4.3.8).4.2.5.APPROVE the selection of a new/changed fuel design before Core Reload design work begins. This approval shall have the concurrence of the Reload Review Board (Reference Attachment 4 for conduct of RRB meeting).(NF Vice President) 1.If the change includes a change in fuel vendor, then OBTAIN the written concurrence of the President and Chief Nuclear Officer.
4.2.3. CLASSIFY the fuel design selected for the Core Reload. Refer to NF-AA-101. (FRE)
2.If the change is not approved, then DIRECT the Core Designer to review alternate fuel design options (i.e., RETURN to Step 4.2.1).4.2.6.PRESENT significant fuel design changes to senior site management via a special Senior Management Design Initialization (SMDI) meeting (Step 4.4). (Core Designer or FRE)4.2.7.If LTAs/LUAs from a different fuel vendor are being introduced, then CONSIDER expanding the duration of the reload schedule to allow for transfer of analysis results between the two fuel vendors. (Core Designer)
: 1. If the fuel design selected is "Like-for-Like" to the fuel design currently in use at the unit, then DOCUMENT this in the Core Reload FCP and SKIP to Step 4.3. Otherwise PROCEED to Step 4.2.4.
NF-AA-100 Revision 13 Page 10 of 54 NOTE: During the Core Reload design process, numerous design analyses (internal and/or external) will be generated to support the reload.
4.2.4. PREPARE a configuration change package for the new or changed fuel design.
These design analyses shall be prepared, reviewed, approved and retained in accordance with CC-AA-309.Processing of these design analyses shall be done using a Core Reload FCP per this procedure and NF-AA-100-1000.
Refer to CC-AA-103 and NF-AA- 101. (FRE or designated individual)
4.3.Core Reload Design Inputs, Assumptions, and Requirements 4.3.1.INITIATE and PREPARE a Core Reload FCP type EC/ECR to control the reload configuration change. Refer to NF-AA-100-1000.(Core Designer) 4.3.2.PERFORM a review of historic (previous 2 to 3 years) and current industry operating experience information to assure that applicable experiences are considered when preparing or reviewing reload products in order to prevent similar problems. Refer to NF-AA-100-1010.(Core Designer)4.3.3.PERFORM a Technical Pre-Job Brief for the Core Reload design effort. Refer to HU-AA-1212.(Core Designer and Manager, BWR/PWR Design) 1.DISCUSS Nuclear Fuels, Inter-Department and potential Independent Third Party reviews. The Pre-Job Brief is expected to be a face-to-face discussion with the Manager, BWR/PWR Design, taking the appropriate questioning/challenging role.
: 1. IDENTIFY all new supporting reviews or analyses that are needed based on the results of the NF-AA-1 01 review.
4.3.4.IDENTIFY members of the Reload Design Overview Team (RDOT).(Core Designer)1.RDOT members and responsibilities are described in Attachment 1.
: 2. ENSURE that Core Reload design inputs or requirements that are impacted by fuel design changes are addressed in the CDIR (Step 4.3.8).
4.3.5.DEFINE the reload Energy Utilization Plan (EUP). Refer to NF-AA-105-1000.(Core Designer)1.CONSULT the unit business plans and Exelon Planned Outage Schedule to ensure that targets for core energy output (i.e., cycle length, power level, effective full power days, and end-of-cycle extension options such as coastdown) are met; 2.If the plant is generator-limited or if the reactor operating power level changes due to seasonal limitations (e.g., river temperature, grid voltage), then OBTAIN a projected reactor power level load profile for the design cycle from the Station; 3.DETERMINE the expected energy requirements for the reload cycle plus additional forecast cycles as specified in NF-AA-105-1000; NF-AA-100 Revision 13 Page 11 of 54 4.DOCUMENT the EUP in the Core Reload FCP and the appropriate vendor transmittals.
4.2.5. APPROVE the selection of a new/changed fuel design before Core Reload design work begins. This approval shall have the concurrence of the Reload Review Board (Reference Attachment 4 for conduct of RRB meeting). (NF Vice President)
4.3.6.SCHEDULE and CONVENE a Reload Design Kickoff Meeting with the fuel vendor.
: 1. If the change includes a change in fuel vendor, then OBTAIN the written concurrence of the President and Chief Nuclear Officer.
Members and responsibilities are described in Attachment 5.(Manager, BWRIPWR Design)1If required, then ENSURE a tracking item is generated to process/evaluate any identified change(s), such as methodology changes, product line updates, shipping requirements, and manufacturing changes.
: 2. If the change is not approved, then DIRECT the Core Designer to review alternate fuel design options (i.e., RETURN to Step 4.2.1).
2.IDENTIFY any long lead-time items, such as Technical Specification Change Requests or other licensing actions that are expected to be required.
4.2.6. PRESENT significant fuel design changes to senior site management via a special Senior Management Design Initialization (SMDI) meeting (Step 4.4). (Core Designer or FRE) 4.2.7. If LTAs/LUAs from a different fuel vendor are being introduced, then CONSIDER expanding the duration of the reload schedule to allow for transfer of analysis results between the two fuel vendors. (Core Designer)
3.IDENTIFY impacts to the training simulator that may result from fuel design changes or changes in the fuel design codes. Changing fuel design codes may result in the inability to perform a cycle specific simulator core update because the simulator uses output from the design codes.
 
A.If simulator updates are impacted, COORDINATE with Site Training and IT actions necessary to maintain cycle specific simulator core updates.B.BE AWARE that this impact may have budget and long lead-time implications.4.3.7.PRESENT the initial EUP to the Reload Review Board before the Fuel Cycle Design activities begin (Refer to Attachment 4 for conduct of RRB meeting).(Core Designer)1.PREPARE the RRB presentation in accordance with the NF standard template.Refer to NF-AA-100-2000.
NF-AA-100 Revision 13 Page 10 of 54 NOTE:     During the Core Reload design process, numerous design analyses (internal and/or external) will be generated to support the reload. These design analyses shall be prepared, reviewed, approved and retained in accordance with CC-AA-309. Processing of these design analyses shall be done using a Core Reload FCP per this procedure and NF-AA-100-1000.
2.INCORPORATE changes required by the RRB and DOCUMENT RRB concurrence with the EUP, If the EUP includes a change in rated power level or significant change in operating strategy, then OBTAIN the written concurrence of the President and Chief Nuclear Officer.
4.3. Core Reload Design Inputs, Assumptions, and Requirements 4.3.1. INITIATE and PREPARE a Core Reload FCP type EC/ECR to control the reload configuration change. Refer to NF-AA-100-1000. (Core Designer) 4.3.2. PERFORM a review of historic (previous 2 to 3 years) and current industry operating experience information to assure that applicable experiences are considered when preparing or reviewing reload products in order to prevent similar problems. Refer to NF-AA-100-1010. (Core Designer) 4.3.3. PERFORM a Technical Pre-Job Brief for the Core Reload design effort. Refer to HU-AA- 1212. (Core Designer and Manager, BWR/PWR Design)
4.If necessary, then REVISE the EUP with the approval of the appropriate Manager, BWR/PWR Design. For significant changes to the EUP due to cycle design actualities (e.g., when achieving the EUP does not offer optimum economics, when an acceptable core design cannot be achieved, or when a significant change in outage/cycle length occurs), OBTAIN the concurrence of the NF VP.
: 1. DISCUSS Nuclear Fuels, Inter-Department and potential Independent Third Party reviews. The Pre-Job Brief is expected to be a face-to-face discussion with the Manager, BWR/PWR Design, taking the appropriate questioning/challenging role.
N F-AA-1 00 Revision 13 Page 12 of 544.3.8.DEFINE the content of the Cycle Design Inputs and Requirements (CDIR) for the reload cycle. The format of the CDIR may be unit-specific and is described by lower level procedures and/or guidelines. Typical items to be included are listed in . The CDIR may consist of multiple documents that are required at different phases during the reload process (e.g., cycle design phase, reload licensing phase, etc.).(RDOT)(CM-3)1.COORDINATE the preparation of the CDIR. (Core Designer)2.INCLUDE the EUP specified by the Reload Review Board in Step 4.3.7.(Core Designer)3.INCLUDE applicable design considerations and impacts using the Design Attribute Review (DAR). Refer to CC-AA-1 02.(Core Designer)4.PROVIDE a draft copy of the CDIR to RDOT members early enough (approximately 2 weeks) to allow sufficient review time.(Core Designer)4.3.9.PERFORM a Risk Management Assessment of the Reload.(Core Designer) 1.The scope of the assessment must CONSIDER all risks from initial fuel receipt of the reload batch through the end of the first cycle (including the offload and shuffle at the end of the cycle), and INCLUDE risks associated with significant changes to the core-operating regime or fuel design. The assessment scope may also be extended to risks during the design process and subsequent cycles.
4.3.4. IDENTIFY members of the Reload Design Overview Team (RDOT). (Core Designer)
NOTE: The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per regulatory processes such as 50.59, 50.90, 50.46, etc.
: 1. RDOT members and responsibilities are described in Attachment 1.
2.CONSIDER in all risk management assessments any changes to the vendor or Exelon methods that support core operation.
4.3.5. DEFINE the reload Energy Utilization Plan (EUP). Refer to NF-AA-105-1000. (Core Designer)
: 1. CONSULT the unit business plans and Exelon Planned Outage Schedule to ensure that targets for core energy output (i.e., cycle length, power level, effective full power days, and end-of-cycle extension options such as coastdown) are met;
: 2. If the plant is generator-limited or if the reactor operating power level changes due to seasonal limitations (e.g., river temperature, grid voltage), then OBTAIN a projected reactor power level load profile for the design cycle from the Station;
: 3. DETERMINE the expected energy requirements for the reload cycle plus additional forecast cycles as specified in NF-AA-105-1000;
 
NF-AA-100 Revision 13 Page 11 of 54
: 4.       DOCUMENT the EUP in the Core Reload FCP and the appropriate vendor transmittals.
4.3.6. SCHEDULE and CONVENE a Reload Design Kickoff Meeting with the fuel vendor.
Members and responsibilities are described in Attachment 5. (Manager, BWRIPWR Design) 1      If required, then ENSURE a tracking item is generated to process/evaluate any identified change(s), such as methodology changes, product line updates, shipping requirements, and manufacturing changes.
: 2.       IDENTIFY any long lead-time items, such as Technical Specification Change Requests or other licensing actions that are expected to be required.
: 3.       IDENTIFY impacts to the training simulator that may result from fuel design changes or changes in the fuel design codes. Changing fuel design codes may result in the inability to perform a cycle specific simulator core update because the simulator uses output from the design codes.
A.       If simulator updates are impacted, COORDINATE with Site Training and IT actions necessary to maintain cycle specific simulator core updates.
B. BE AWARE that this impact may have budget and long lead-time implications.
4.3.7. PRESENT the initial EUP to the Reload Review Board before the Fuel Cycle Design activities begin (Refer to Attachment 4 for conduct of RRB meeting). (Core Designer)
: 1.     PREPARE the RRB presentation in accordance with the NF standard template. Refer to NF-AA-100-2000.
: 2.     INCORPORATE changes required by the RRB and DOCUMENT RRB concurrence with the EUP, If the EUP includes a change in rated power level or significant change in operating strategy, then OBTAIN the written concurrence of the President and Chief Nuclear Officer.
: 4.     If necessary, then REVISE the EUP with the approval of the appropriate Manager, BWR/PWR Design. For significant changes to the EUP due to cycle design actualities (e.g., when achieving the EUP does not offer optimum economics, when an acceptable core design cannot be achieved, or when a significant change in outage/cycle length occurs), OBTAIN the concurrence of the NF VP.
 
N F-AA-1 00 Revision 13 Page 12 of 54 4.3.8. DEFINE the content of the Cycle Design Inputs and Requirements (CDIR) for the reload cycle. The format of the CDIR may be unit-specific and is described by lower level procedures and/or guidelines. Typical items to be included are listed in Attachment 2. The CDIR may consist of multiple documents that are required at different phases during the reload process (e.g., cycle design phase, reload licensing phase, etc.). (RDOT) (CM-3)
: 1.     COORDINATE the preparation of the CDIR. (Core Designer)
: 2.     INCLUDE the EUP specified by the Reload Review Board in Step 4.3.7.
(Core Designer)
: 3.     INCLUDE applicable design considerations and impacts using the Design Attribute Review (DAR). Refer to CC-AA-1 02. (Core Designer)
: 4. PROVIDE a draft copy of the CDIR to RDOT members early enough (approximately 2 weeks) to allow sufficient review time. (Core Designer) 4.3.9. PERFORM a Risk Management Assessment of the Reload. (Core Designer)
: 1. The scope of the assessment must CONSIDER all risks from initial fuel receipt of the reload batch through the end of the first cycle (including the offload and shuffle at the end of the cycle), and INCLUDE risks associated with significant changes to the core-operating regime or fuel design. The assessment scope may also be extended to risks during the design process and subsequent cycles.
NOTE:     The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per regulatory processes such as 50.59, 50.90, 50.46, etc.
: 2. CONSIDER in all risk management assessments any changes to the vendor or Exelon methods that support core operation.
ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.
ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.
ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.-DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).
ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.
NF-AA-100 Revision 13 Page 13 of 54 3.PERFORM the following required elements in the Risk Management Assessment: (CM-12)-IDENTIFY Risk Triggers
              -       DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).
-DETERMINE Compensatory Actions
 
-IDENTIFY Contingency Plans
NF-AA-100 Revision 13 Page 13 of 54
-IDENTIFY Monitoring Plans 4.DETERMINE if an Independent Third Party Review is required. Refer to HU-AA-1212 and NF-AA-100-1600.
: 3.     PERFORM the following required elements in the Risk Management Assessment: (CM-12)
5.COMMUNICATE the resulting Risk Management Assessment with NF and Station management through the RRB and SMDI processes described in subsequent steps.
                -       IDENTIFY Risk Triggers
4.3.10.REVIEW and CONCUR with completed CDIR documentation and the Risk Management Assessment before they are used in definitive design analyses.(RDOT)1.Appropriate NF and site representatives should PERFORM this review to ensure that the personnel responsible for said items approve. This may include representatives on the RDOT and others as necessary.
                -       DETERMINE Compensatory Actions
4.3.11.APPROVE the Risk Management Assessment before the reload design is released for definitive design analyses.(Manager, BWR/PWR Design)4.4.Senior Management Review and Approval NOTE: The SMDI review of the cycle design may be performed after completion of the Preliminary Fuel Cycle Design (PFCD) in Step 4.5.
                -       IDENTIFY Contingency Plans
4.4.1.PRESENT the key information for the proposed cycle design (e.g., EUP, fuel design, operating strategy, modifications, training requirements, reload schedule, Risk Management Assessment, recent fuel vendor audit/assessment results, near term future audit/assessment goals, etc.) to senior site management via the Senior Management Design Initialization (SMDI) meeting. The scope and focus of the SMDI meeting and the personnel required to attend are discussed in Attachment 4.(Core Designer)(CM-12)1.PREPARE the SMDI presentation in accordance with the NF standard template. Refer to NF-AA-100-2000.
                -       IDENTIFY Monitoring Plans
4.4.2.If the cycle design plans are acceptable, then DOCUMENT SMDI concurrence.(Core Designer)
: 4.     DETERMINE if an Independent Third Party Review is required. Refer to HU-AA-1212 and NF-AA-100-1600.
NF-AA-100 Revision 13 Page 14 of 54 4.4.3.If the cycle design plans are not acceptable, then DEFINE requirements to be satisfied for concurrence.(SMD!)1.ADD the resulting requirements to the reload support work scope, CDIR, and/or schedule.(Core Designer)2.If the SMDI is held after the PFCD is developed in Step 4.
: 5.     COMMUNICATE the resulting Risk Management Assessment with NF and Station management through the RRB and SMDI processes described in subsequent steps.
5 and the SMDI requirements require a significant change to the PFCD and/or its characteristics, then RETURN to earlier analysis and approval Steps of this procedure.(Core Designer)4.5.Preliminary Fuel Cycle Design (PFCD)
4.3.10. REVIEW and CONCUR with completed CDIR documentation and the Risk Management Assessment before they are used in definitive design analyses.
NOTE: Subject to the election of appropriate commercial options in the fuel contract, either the fuel vendor or NF staff may perform the bundle/cycle design activities discussed in this section.4.5.1.DEVELOP Scoping Fuel Bundle/Cycle Design(s) to be considered as the first step leading to the identification of the PFCD. Criteria for developing alternate designs may include operating margin, core limits, economics, EUP, operating strategy, etc.(Core Designer) 1.DETERMINE the range of design options to be considered.(Manager, BWRIPWR Design)
(RDOT)
A.SPECIFY the level of detail required, which need not be that required for formal calculations or analyses, and SPECIFY design comparison criteria.B.SPECIFY the number of future operating cycles to be considered for each design option.
: 1. Appropriate NF and site representatives should PERFORM this review to ensure that the personnel responsible for said items approve. This may include representatives on the RDOT and others as necessary.
2.USE the design inputs specified in Step 4.3 of this procedure to design Scoping Fuel Bundle/Cycle Design(s). Some flexibility in the design inputs may be allowed to fully evaluate options.(Core Designer)A.COLLABORATE with the fuel vendor, as needed, in developing bundle/cycle designs.
4.3.11. APPROVE the Risk Management Assessment before the reload design is released for definitive design analyses. (Manager, BWR/PWR Design) 4.4. Senior Management Review and Approval NOTE:       The SMDI review of the cycle design may be performed after completion of the Preliminary Fuel Cycle Design (PFCD) in Step 4.5.
B.TRANSMIT the EUP and CDIR to the vendor. Refer to CC-AA-310.
4.4.1. PRESENT the key information for the proposed cycle design (e.g., EUP, fuel design, operating strategy, modifications, training requirements, reload schedule, Risk Management Assessment, recent fuel vendor audit/assessment results, near term future audit/assessment goals, etc.) to senior site management via the Senior Management Design Initialization (SMDI) meeting. The scope and focus of the SMDI meeting and the personnel required to attend are discussed in Attachment 4.
(Core Designer) (CM-12)
: 1.     PREPARE the SMDI presentation in accordance with the NF standard template. Refer to NF-AA-100-2000.
4.4.2. If the cycle design plans are acceptable, then DOCUMENT SMDI concurrence.
(Core Designer)
 
NF-AA-100 Revision 13 Page 14 of 54 4.4.3. If the cycle design plans are not acceptable, then DEFINE requirements to be satisfied for concurrence. (SMD!)
: 1.     ADD the resulting requirements to the reload support work scope, CDIR, and/or schedule. (Core Designer)
: 2.     If the SMDI is held after the PFCD is developed in Step 4. 5 and the SMDI requirements require a significant change to the PFCD and/or its characteristics, then RETURN to earlier analysis and approval Steps of this procedure. (Core Designer) 4.5. Preliminary Fuel Cycle Design (PFCD)
NOTE:       Subject to the election of appropriate commercial options in the fuel contract, either the fuel vendor or NF staff may perform the bundle/cycle design activities discussed in this section.
4.5.1. DEVELOP Scoping Fuel Bundle/Cycle Design(s) to be considered as the first step leading to the identification of the PFCD. Criteria for developing alternate designs may include operating margin, core limits, economics, EUP, operating strategy, etc.
(Core Designer)
: 1.     DETERMINE the range of design options to be considered. (Manager, BWRIPWR Design)
A.     SPECIFY the level of detail required, which need not be that required for formal calculations or analyses, and SPECIFY design comparison criteria.
B.     SPECIFY the number of future operating cycles to be considered for each design option.
: 2.     USE the design inputs specified in Step 4.3 of this procedure to design Scoping Fuel Bundle/Cycle Design(s). Some flexibility in the design inputs may be allowed to fully evaluate options. (Core Designer)
A.     COLLABORATE with the fuel vendor, as needed, in developing bundle/cycle designs.
B.     TRANSMIT the EUP and CDIR to the vendor. Refer to CC-AA-310.
This transmittal may be delayed if the vendor is not involved in the PFCD.
This transmittal may be delayed if the vendor is not involved in the PFCD.
NF-AA-1 00 Revision 13 Page 15 of 544.5.2.REVIEW the proposed Scoping Fuel Bundle/Cycle Design alternatives and SELECT the design to be taken forward to the more detailed PFCD design stage.(Manager, BWRIPWR Design) 1.CONSULT the Reload Review Board, if necessary (e.g., if the EUP is revised significantly or a significant change in operating strategy is proposed).(Core Designer)OBTAIN confirmation from the fuel vendor that the selected reload fuel assembly/bundle design(s) can be fabricated, licensed, and shipped under the terms of the current contract.(Core Designer)(CM-2)A.IDENTIFY any potential manufacturing problems or constraints and CONSIDER vendor requests for changes to the design that will assist in the manufacturing campaign.
OBTAIN vendor experience or feedback on any other aspect of the fuel designs and IDENTIFY potential improvements.4.5.3.FINALIZE the Preliminary Fuel Cycle Design (PFCD) based on the Scoping Fuel Cycle Design selected. (Core Designer)(CM-9)1.IMPLEMENT the approved Core Design Inputs and Requirements, including both the approved EUP and the approved Fuel Design, into the PFCD.
2.MODEL the reload cycle plus two additional cycles to provide long-term fuel cycle optimization and economics information. The additional cycles may be modeled with existing or proposed fuel designs.
3.REVIEW the Risk Management Assessment, and MODIFY if necessary.
A.If the Risk Management Assessment (RMA) is significantly modified, COMMUNICATE the new RMA to NF and Station management (SMDI membership), per NF-AA-100-1600, if they have already received the RMA at this point in the reload process. Non-significant/editorial changes do not require senior management review.4.5.4.CONVENE the RDOT to perform a Design Technical Review (DTR) of the PFCD.(Core Designer)(CM-4)1.PRESENT the PFCD and the Risk Management Assessment and EVALUATE the proposed cycle design against the CDIR and the Reload Design Review Criteria specified in lower level procedures and/or guidelines.
Criteria to be considered are listed in Attachment 3.(Assigned NF Engineers)
NF-AA-100 Revision 13 Page 16 of 54 A.ENSURE that the core loading pattern clearly identifies the locations of the various fuel batches in the core (e.g., using batch designators, color coding, etc.)
2.ENSURE that all inputs, requirements and impacts relating to their area of expertise are being addressed.(RDOT)3.RAISE any concerns that may impede successful implementation of the PFCD.(RDOT)A.RESOLVE RDOT concerns or ASSURE that concerns can be resolved by future reload work scope (Core Designer)
B.ADDRESS new design considerations and impacts in the FCP.(Core Designer)1.If there is an unresolved concern that may jeopardize the PFCD, then RETURN to Step 4.5.1, 4.5.2, or 4.5.3, as necessary.
C.Promptly NOTIFY the Manager, BWR/PWR Design, NF VP, and site management of significant issues or concerns.(Core Designer) 4.DOCUMENT the minutes of the DTR. (Core Designer)A.If needed, then IDENTIFY all work scope necessary to support design implementation including any significant design changes or new fuel design implementation requirements.
4.5.5.PRESENT the PFCD and the Risk Management Assessment to the Reload Review Board (Refer to Attachment 4 for conduct of RRB meeting).(Core Designer)1.PREPARE the RRB presentation in accordance with the NF standard template and NOTE changes from the initial RRB meeting. Refer to NF-AA-100-2000.2.INCORPORATE any changes required by the RRB and DOCUMENT RRB concurrence with the PFCD and the Risk Management Assessment.
A.If Significant changes occur and if needed, then RETURN to earlier Steps of this procedure.
NF-AA-100 Revision 13 Page 17 of 54 B.If the Risk Management Assessment (RMA) is significantly modified per the RRB, COMMUNICATE the new RMA, per NF-AA-100-1600, to any NF and Station management (SMDI membership) who have already received the RMA at this point in the reload process. Non-significant/editorial changes do not require senior management review.PROVIDE the multi-cycle PFCD prediction to the Fuel Buyer for the basis of long-term nuclear materials, conversion and enrichment requirements forecasts.(Core Designer)1.If later changes associated with the Final Fuel Cycle Design or the as-loaded core significantly affect this information, then PROVIDE updated data.4.5.7.PROVIDE the multi-cycle PFCD prediction to the fuel accounting organization, to form the basis of fuel accounting and economics forecasts.(Fuel Buyer) 1.If later changes associated with the Final Fuel Cycle Design or the as-loaded core significantly affect this information, then PROVIDE updated data.4.5.8.OBTAIN fuel vendor review of the PFCD, if included in the scope of the applicable fuel contract.(Core Designer) 1.IDENTIFY any potential manufacturing problems or constraints and CONSIDER vendor requests for changes to the design that will assist in the manufacturing campaign.
2.OBTAIN vendor experience or feedback on any other aspect of the design.
3.COLLABORATE with the vendor to identify improvements to the PFCD that can be incorporated into the Final Fuel Cycle Design.
4.ADDRESS new design considerations and impacts in the FCP.
AGREE on the preliminary schedule for final fuel delivery and all other company and vendor deliverables required under the fuel fabrication contract.4.5.9.For certain vendor-specific reload plans, the PFCD may be used as the basis for the fuel order. If the fuel order is placed based on the approved PFCD, then PERFORM Steps 4.7.1 thru 4.7.4 at this point.(Core Designer)
NF-AA-100 Revision 13 Page 18 of 54 4.6.Final Fuel Cycle Design NOTE: Subject to the election of appropriate commercial options in the fuel contract, either the fuel vendor or NF staff may perform the cycle design activities discussed in this section.NOTE: The FFCD need only model the upcoming Core Reload cycle.Multi-cycle modeling is not required.4.6.1.DEVELOP the Final Fuel Cycle Design (FFCD). (Core Designer)1.INCLUDE any changes on which RDOT and/or SMDI concurrence are based.(Core Designer)2.COLLABORATE with the fuel vendor and INCORPORATE any vendor-recommended design changes agreed by the Manager, BWR/PWR Design.(Core Designer)3.VERIFY the material condition of any re-insert fuel (previously irradiated fuel that is resident in the fuel pool during the current operating cycle) utilized in the FFCD. (Core Designer)A.OBTAIN material condition information from Reactor Engineer and/or Fuel Reliability Engineer.(Core Designer)
B.DEVELOP an action plan (e.g., FFCD redesign, fuel inspection, fuel repair, etc.) if the material condition of any re-insert fuel is questionable or needs repair.(Core Designer, Reactor Engineer and NF Technical Support Engineer) 4.ASSESS the impact of any changes between the PFCD and FFCD, versus the CDIR, the Risk Management Assessment, and the Reload Design Review Criteria.Significant changes or additional information not available during the PFCD phase may require a second design technical review per Step 4.5.4.(RDOT)A.If required by Manager, BWR/PWR Design, then REQUEST vendor review of the FFCD.(Core Designer) 5.Promptly NOTIFY the Manager, BWR/PWR Design, Fuel Buyer and the RDOT; plus the NF VP and site management as necessary, of significant changes. Also, NOTIFY these and other stakeholders of minor changes at the earliest convenient opportunity.(Core Designer)A.If significant changes occur, then OBTAIN RRB concurrence.
NF-AA-100 Revision 13 Page 19 of 54 6.VERIFY the design per applicable procedures and DOCUMENT the FFCD in the FCP.(Core Designer)4.6.2.PROVIDE the affected departments (typically the RDOT members) with a summary of the FFCD for use in early identification of procedure and training requirements for the reload.(Core Designer)1.IDENTIFY anticipated procedure changes, deletions, and additions. (RDOT) 2.IDENTIFY training changes, deletions, and additions, and FORWARD to the Training department to support identification and completion of training requirements. (RDOT)
NOTE: The PFCD may have been used as the basis for the fuel order instead of the FFCD per Step 4.5.9.
4.6.3.FREEZE the FFCD. The FFCD is to be used as the basis for the fuel order, licensing analyses and reload preparations.(Core Designer) 1.If needed, then EXECUTE small changes to the Reload Core design and new fuel requirements may be made subsequent to the FFCD, to better optimize the reload. Such changes must be acceptable to the fuel vendor under the terms of the fuel contract and of sufficiently minor impact not to invalidate any reload licensing work in progress.
2.If more significant perturbations occur, then CONSIDER these changes as a redesign and HANDLE as described in Step 4.9.5.
4.7.Fuel Procurement, Fabrication and Receipt 4.7.1.VERIFY that the following prerequisites/inputs have been addressed.(Core Designer)-All necessary prior RRB approvals plus President and Chief Nuclear Officer concurrence, when necessary- SMDI and RDOT concurrence
-FFCD (or PFCD) has been reviewed by the fuel vendor
-FFCD (or PFCD) has been reviewed against the CDIR 4.7.2.PROVIDE (or CONFIRM vendor provides) the finalized FFCD (or PFCD) enriched material requirements to the Fuel Buyer for procurement processing.(Core Designer)
NF-AA-1 00 Revision 13 Page 20 of 54 1.CONFIRM the enriched material delivery date.(Fuel Buyer)4.7.3.ORDER fuel and fuel-related components based on the requirements of the approved FFCD (or PFCD).(Fuel Buyer) 1.VERIFY significant design changes/new fuel designs have been evaluated sufficiently and that unverified portions of changes are controlled. Refer to


CC-AA-1 03 and NF-AA-1 01.
NF-AA-1 00 Revision 13 Page 15 of 54 4.5.2. REVIEW the proposed Scoping Fuel Bundle/Cycle Design alternatives and SELECT the design to be taken forward to the more detailed PFCD design stage. (Manager, BWRIPWR Design)
2.OBTAIN the necessary order information (e.g., quantities, designs, specifications, etc.) from the Core Designer.
: 1. CONSULT the Reload Review Board, if necessary (e.g., if the EUP is revised significantly or a significant change in operating strategy is proposed). (Core Designer)
A.ADJUST the Fuel order to account for inventory items to be either supplied or utilized.
OBTAIN confirmation from the fuel vendor that the selected reload fuel assembly/bundle design(s) can be fabricated, licensed, and shipped under the terms of the current contract. (Core Designer) (CM-2)
3.OBTAIN required delivery schedules from the Site, Outage Services or per the relevant contract.
A.     IDENTIFY any potential manufacturing problems or constraints and CONSIDER vendor requests for changes to the design that will assist in the manufacturing campaign.
ORDER fuel and fuel-related components in compliance with all conditions (such as lead-time) of the relevant contract(s).4.7.4.COORDINATE monitoring of the fabrication process.(NF Engineer for Fuel Fabrication Oversight) 1.MONITOR the progress of the fabrication campaign.(NF Engineer for Fuel Fabrication Oversight) 2.MONITOR generic issues associated with the fuel vendor and its fabrication facility.(NF Engineer for Fuel Fabrication Oversight) 3.CONDUCT assessments and surveillances of vendors, as appropriate, to evaluate design, licensing, and fabrication activities for fuel and other related components and services.(Nuclear Oversight Vendor Auditor)(CM-14)NF shall SUPPLY technical specialists, as requested, to assist in the assessments of the fuel design, licensing, and fabrication activities.
OBTAIN vendor experience or feedback on any other aspect of the fuel designs and IDENTIFY potential improvements.
NF will PROVIDE a summary report of the results of the most recent fuel vendor audits/assessments and actions for future SMDI meetings.4.7.5.ESTABLISH final fuel and component delivery schedules and associated shipping and security instructions with the vendor(s).(Fuel Buyer and/or Site)4.7.6.CONFIRM the reload design has been determined to be acceptable by the successful completion of all previous steps of this procedure. (Core Designer)1.If required, then INCLUDE review of applicable CC-AA-103 and NF-AA-101 activities/analyses.
4.5.3. FINALIZE the Preliminary Fuel Cycle Design (PFCD) based on the Scoping Fuel Cycle Design selected. (Core Designer) (CM-9)
NF-AA-100 Revision 13 Page 21 of 54 2.If any previous Step remains incomplete just prior to scheduled Fuel Receipt, then NF shall NOTIFY both the Site and Fuel Vendor of any restrictions which must be applied pending that Step's successful completion.
: 1. IMPLEMENT the approved Core Design Inputs and Requirements, including both the approved EUP and the approved Fuel Design, into the PFCD.
3.If required, then ENSURE that all design analyses required to receive and store new fuel designs on-site have been completed and documented in the applicable DCP per NF-AA-101.
: 2. MODEL the reload cycle plus two additional cycles to provide long-term fuel cycle optimization and economics information. The additional cycles may be modeled with existing or proposed fuel designs.
4.VERIFY that the reactivity requirements to receive and store the fresh reload batch on-site have been met.
: 3. REVIEW the Risk Management Assessment, and MODIFY if necessary.
4.7.7.RECEIVE fuel according to procedures that require a physical inspection of the fuel, plus a review of supporting vendor certifications and as-built data.(Outage Services)1.ENSURE that the delivered fuel meets pre-defined acceptance criteria.
A.      If the Risk Management Assessment (RMA) is significantly modified, COMMUNICATE the new RMA to NF and Station management (SMDI membership), per NF-AA-100-1600, if they have already received the RMA at this point in the reload process. Non-significant/editorial changes do not require senior management review.
2.Any fuel that fails these inspections shall either:
4.5.4. CONVENE the RDOT to perform a Design Technical Review (DTR) of the PFCD.
be RECTIFIED or REPLACED by the fuel vendor; or be subject to a full assessment to determine its suitability for use.
(Core Designer) (CM-4)
4.8.Reload Licensing and Core Loading Plan 4.8.1.If the PFCD was used for fuel procurement, then DEVELOP the FFCD per step 4.6.1 prior to Reload Licensing and Core Loading Plan activities.(Core Designer) 4.8.2.INITIATE the scope of work required for the successful licensing of the FFCD.(Core Designer)1.DOCUMENT the Licensing and Core Loading information in the Core Reload FCP.2.COLLABORATE with all stakeholders and the fuel vendor to agree on the associated responsibilities, acceptance criteria, specific deliverables, and schedule.3.ENSURE the resolution of all remaining concerns associated with the reload Design Technical Review are included in the work scope including any design/licensing basis changes required by significant design changes or new fuel design. Refer to NF-AA-101.
: 1. PRESENT the PFCD and the Risk Management Assessment and EVALUATE the proposed cycle design against the CDIR and the Reload Design Review Criteria specified in lower level procedures and/or guidelines.
4.ENSURE all analysis inputs, assumptions, methods, outputs and acceptance criteria for the reload analysis are either specified or accepted prior to the start of work.(CM-2)
Criteria to be considered are listed in Attachment 3. (Assigned NF Engineers)
NF-AA-100 Revision 13 Page 22 of 544.8.3.If required, then ENSURE a 50.59 Review is performed for any identified changes, such as methodology changes, product line updates, shipping requirements and manufacturing changes, which support the Core Reload FCP. (Refer to LS-AA-104)(Core Designer)4.8.4.INITIATE any licensing actions required for the reload, for instance changes to the unit Technical Specifications.(Licensing and Regulatory Affairs) 1.INITIATE licensing actions as soon as practicable to ensure sufficient lead-time for regulatory submittal and review.
 
2.Where licensing changes or threats are identified later in the reload process, promptly NOTIFY all stakeholders so that action plans can be developed.4.8.5.MANAGE completion of reload analysis and licensing support. (Core Designer)1.COORDINATE the scheduling, monitoring, completion and review of all reload support work.
NF-AA-100 Revision 13 Page 16 of 54 A. ENSURE that the core loading pattern clearly identifies the locations of the various fuel batches in the core (e.g., using batch designators, color coding, etc.)
2.ENSURE that all identified reload design and licensing scope is being addressed by scheduled work activities.
: 2. ENSURE that all inputs, requirements and impacts relating to their area of expertise are being addressed. (RDOT)
3.ENSURE that incomplete analyses are expected to be resolved in a timely manner, or have sufficient contingencies.
: 3. RAISE any concerns that may impede successful implementation of the PFCD. (RDOT)
4.ADAPT the reload support work scope and schedule, as agreed with key stakeholders, to account for any analysis problems.
A. RESOLVE RDOT concerns or ASSURE that concerns can be resolved by future reload work scope (Core Designer)
5.COORDINATE with site to ensure that the outage work program includes all fuel-related activities.
B. ADDRESS new design considerations and impacts in the FCP.
6.PROVIDE reload core design data to the Operations Training Department for simulator testing.
(Core Designer)
Refer to TQ-AA-303. The ability to supply this data may be impacted by the actions identified in Step 4.3.6.3.
: 1. If there is an unresolved concern that may jeopardize the PFCD, then RETURN to Step 4.5.1, 4.5.2, or 4.5.3, as necessary.
7.ENSURE that all source documentation is produced by appropriate parties (fuel vendor, contractors, NF, etc.) in a timely manner, reviewed, then distributed to appropriate recipients at site.
C. Promptly NOTIFY the Manager, BWR/PWR Design, NF VP, and site management of significant issues or concerns. (Core Designer)
A.DOCUMENT the review and acceptance of fuel vendor and contractor documentation in the Core Reload FCP.
: 4. DOCUMENT the minutes of the DTR. (Core Designer)
8.DOCUMENT the applicable design analyses, UFSAR updates, ADL changes, etc. in the Core Reload FCP.
A. If needed, then IDENTIFY all work scope necessary to support design implementation including any significant design changes or new fuel design implementation requirements.
NF-AA-100 Revision 13 Page 23 of 54 9.If significant changes are made to the inputs, assumptions, or methods used in reload analysis and/or licensing, WRITE an IR to document the changes and DISCUSS the changes during future Reload Kickoff meetings, per .4.8.6.PROVIDE details of any "new" or "non-standard" testing requirements and acceptance test criteria for the reload in the FCP. (Core Designer)1.Promptly NOTIFY Reactor Engineering of the"new" or "non-standard" testing once its existence becomes known.4.8.7.PREPARE the reload COLR, based on the results collated at Step 4.8.5.(Core Designer)4.8.8.PREPARE the reload IOCFR50.59 screening/
4.5.5. PRESENT the PFCD and the Risk Management Assessment to the Reload Review Board (Refer to Attachment 4 for conduct of RRB meeting). (Core Designer)
evaluation per LS-AA-104.(Core Designer)4.8.9.PROVIDE information to Site Training Department to support development of training materials for the Core Reload design.(Core Designer)1.SOLICIT Reactor Engineering and Training input on content of information.
: 1. PREPARE the RRB presentation in accordance with the NF standard template and NOTE changes from the initial RRB meeting. Refer to NF-AA-100-2000.
: 2. INCORPORATE any changes required by the RRB and DOCUMENT RRB concurrence with the PFCD and the Risk Management Assessment.
A. If Significant changes occur and if needed, then RETURN to earlier Steps of this procedure.
 
NF-AA-100 Revision 13 Page 17 of 54 B.      If the Risk Management Assessment (RMA) is significantly modified per the RRB, COMMUNICATE the new RMA, per NF-AA-100-1600, to any NF and Station management (SMDI membership) who have already received the RMA at this point in the reload process. Non-significant/editorial changes do not require senior management review.
4.5.6. PROVIDE the multi-cycle PFCD prediction to the Fuel Buyer for the basis of long-term nuclear materials, conversion and enrichment requirements forecasts. (Core Designer)
: 1. If later changes associated with the Final Fuel Cycle Design or the as-loaded core significantly affect this information, then PROVIDE updated data.
4.5.7. PROVIDE the multi-cycle PFCD prediction to the fuel accounting organization, to form the basis of fuel accounting and economics forecasts. (Fuel Buyer)
: 1.      If later changes associated with the Final Fuel Cycle Design or the as-loaded core significantly affect this information, then PROVIDE updated data.
4.5.8. OBTAIN fuel vendor review of the PFCD, if included in the scope of the applicable fuel contract. (Core Designer)
: 1.      IDENTIFY any potential manufacturing problems or constraints and CONSIDER vendor requests for changes to the design that will assist in the manufacturing campaign.
: 2. OBTAIN vendor experience or feedback on any other aspect of the design.
: 3. COLLABORATE with the vendor to identify improvements to the PFCD that can be incorporated into the Final Fuel Cycle Design.
: 4. ADDRESS new design considerations and impacts in the FCP.
AGREE on the preliminary schedule for final fuel delivery and all other company and vendor deliverables required under the fuel fabrication contract.
4.5.9. For certain vendor-specific reload plans, the PFCD may be used as the basis for the fuel order. If the fuel order is placed based on the approved PFCD, then PERFORM Steps 4.7.1 thru 4.7.4 at this point. (Core Designer)
 
NF-AA-100 Revision 13 Page 18 of 54 4.6. Final Fuel Cycle Design NOTE:      Subject to the election of appropriate commercial options in the fuel contract, either the fuel vendor or NF staff may perform the cycle design activities discussed in this section.
NOTE:      The FFCD need only model the upcoming Core Reload cycle. Multi-cycle modeling is not required.
4.6.1. DEVELOP the Final Fuel Cycle Design (FFCD). (Core Designer)
: 1.      INCLUDE any changes on which RDOT and/or SMDI concurrence are based.
(Core Designer)
: 2.      COLLABORATE with the fuel vendor and INCORPORATE any vendor-recommended design changes agreed by the Manager, BWR/PWR Design.
(Core Designer)
: 3. VERIFY the material condition of any re-insert fuel (previously irradiated fuel that is resident in the fuel pool during the current operating cycle) utilized in the FFCD. (Core Designer)
A.      OBTAIN material condition information from Reactor Engineer and/or Fuel Reliability Engineer. (Core Designer)
B.      DEVELOP an action plan (e.g., FFCD redesign, fuel inspection, fuel repair, etc.) if the material condition of any re-insert fuel is questionable or needs repair. (Core Designer, Reactor Engineer and NF Technical Support Engineer)
: 4. ASSESS the impact of any changes between the PFCD and FFCD, versus the CDIR, the Risk Management Assessment, and the Reload Design Review Criteria. Significant changes or additional information not available during the PFCD phase may require a second design technical review per Step 4.5.4. (RDOT)
A.      If required by Manager, BWR/PWR Design, then REQUEST vendor review of the FFCD. (Core Designer)
: 5. Promptly NOTIFY the Manager, BWR/PWR Design, Fuel Buyer and the RDOT; plus the NF VP and site management as necessary, of significant changes. Also, NOTIFY these and other stakeholders of minor changes at the earliest convenient opportunity. (Core Designer)
A.      If significant changes occur, then OBTAIN RRB concurrence.
 
NF-AA-100 Revision 13 Page 19 of 54
: 6.      VERIFY the design per applicable procedures and DOCUMENT the FFCD in the FCP. (Core Designer) 4.6.2. PROVIDE the affected departments (typically the RDOT members) with a summary of the FFCD for use in early identification of procedure and training requirements for the reload. (Core Designer)
: 1.      IDENTIFY anticipated procedure changes, deletions, and additions. (RDOT)
: 2.      IDENTIFY training changes, deletions, and additions, and FORWARD to the Training department to support identification and completion of training requirements. (RDOT)
NOTE:      The PFCD may have been used as the basis for the fuel order instead of the FFCD per Step 4.5.9.
4.6.3. FREEZE the FFCD. The FFCD is to be used as the basis for the fuel order, licensing analyses and reload preparations. (Core Designer)
: 1. If needed, then EXECUTE small changes to the Reload Core design and new fuel requirements may be made subsequent to the FFCD, to better optimize the reload. Such changes must be acceptable to the fuel vendor under the terms of the fuel contract and of sufficiently minor impact not to invalidate any reload licensing work in progress.
: 2.      If more significant perturbations occur, then CONSIDER these changes as a redesign and HANDLE as described in Step 4.9.5.
4.7. Fuel Procurement, Fabrication and Receipt 4.7.1. VERIFY that the following prerequisites/inputs have been addressed. (Core Designer)
      -  All necessary prior RRB approvals plus President and Chief Nuclear Officer concurrence, when necessary
      - SMDI and RDOT concurrence
      -    FFCD (or PFCD) has been reviewed by the fuel vendor
      - FFCD (or PFCD) has been reviewed against the CDIR 4.7.2. PROVIDE (or CONFIRM vendor provides) the finalized FFCD (or PFCD) enriched material requirements to the Fuel Buyer for procurement processing. (Core Designer)
 
NF-AA-1 00 Revision 13 Page 20 of 54
: 1. CONFIRM the enriched material delivery date. (Fuel Buyer) 4.7.3. ORDER fuel and fuel-related components based on the requirements of the approved FFCD (or PFCD). (Fuel Buyer)
: 1. VERIFY significant design changes/new fuel designs have been evaluated sufficiently and that unverified portions of changes are controlled. Refer to CC-AA-1 03 and NF-AA-1 01.
: 2. OBTAIN the necessary order information (e.g., quantities, designs, specifications, etc.) from the Core Designer.
A.      ADJUST the Fuel order to account for inventory items to be either supplied or utilized.
: 3. OBTAIN required delivery schedules from the Site, Outage Services or per the relevant contract.
ORDER fuel and fuel-related components in compliance with all conditions (such as lead-time) of the relevant contract(s).
4.7.4. COORDINATE monitoring of the fabrication process. (NF Engineer for Fuel Fabrication Oversight)
: 1. MONITOR the progress of the fabrication campaign. (NF Engineer for Fuel Fabrication Oversight)
: 2. MONITOR generic issues associated with the fuel vendor and its fabrication facility. (NF Engineer for Fuel Fabrication Oversight)
: 3. CONDUCT assessments and surveillances of vendors, as appropriate, to evaluate design, licensing, and fabrication activities for fuel and other related components and services. (Nuclear Oversight Vendor Auditor) (CM-14)
NF shall SUPPLY technical specialists, as requested, to assist in the assessments of the fuel design, licensing, and fabrication activities.
NF will PROVIDE a summary report of the results of the most recent fuel vendor audits/assessments and actions for future SMDI meetings.
4.7.5. ESTABLISH final fuel and component delivery schedules and associated shipping and security instructions with the vendor(s). (Fuel Buyer and/or Site) 4.7.6. CONFIRM the reload design has been determined to be acceptable by the successful completion of all previous steps of this procedure. (Core Designer)
: 1. If required, then INCLUDE review of applicable CC-AA-103 and NF-AA-101 activities/analyses.
 
NF-AA-100 Revision 13 Page 21 of 54
: 2.      If any previous Step remains incomplete just prior to scheduled Fuel Receipt, then NF shall NOTIFY both the Site and Fuel Vendor of any restrictions which must be applied pending that Step's successful completion.
: 3.      If required, then ENSURE that all design analyses required to receive and store new fuel designs on-site have been completed and documented in the applicable DCP per NF-AA-101.
: 4.      VERIFY that the reactivity requirements to receive and store the fresh reload batch on-site have been met.
4.7.7. RECEIVE fuel according to procedures that require a physical inspection of the fuel, plus a review of supporting vendor certifications and as-built data. (Outage Services)
: 1.      ENSURE that the delivered fuel meets pre-defined acceptance criteria.
: 2.      Any fuel that fails these inspections shall either: be RECTIFIED or REPLACED by the fuel vendor; or be subject to a full assessment to determine its suitability for use.
4.8. Reload Licensing and Core Loading Plan 4.8.1. If the PFCD was used for fuel procurement, then DEVELOP the FFCD per step 4.6.1 prior to Reload Licensing and Core Loading Plan activities. (Core Designer) 4.8.2. INITIATE the scope of work required for the successful licensing of the FFCD. (Core Designer)
: 1.      DOCUMENT the Licensing and Core Loading information in the Core Reload FCP.
: 2.      COLLABORATE with all stakeholders and the fuel vendor to agree on the associated responsibilities, acceptance criteria, specific deliverables, and schedule.
: 3.      ENSURE the resolution of all remaining concerns associated with the reload Design Technical Review are included in the work scope including any design/licensing basis changes required by significant design changes or new fuel design. Refer to NF-AA-101.
: 4.      ENSURE all analysis inputs, assumptions, methods, outputs and acceptance criteria for the reload analysis are either specified or accepted prior to the start of work. (CM-2)
 
NF-AA-100 Revision 13 Page 22 of 54 4.8.3. If required, then ENSURE a 50.59 Review is performed for any identified changes, such as methodology changes, product line updates, shipping requirements and manufacturing changes, which support the Core Reload FCP. (Refer to LS-AA-104)
(Core Designer) 4.8.4. INITIATE any licensing actions required for the reload, for instance changes to the unit Technical Specifications. (Licensing and Regulatory Affairs)
: 1.     INITIATE licensing actions as soon as practicable to ensure sufficient lead-time for regulatory submittal and review.
: 2.     Where licensing changes or threats are identified later in the reload process, promptly NOTIFY all stakeholders so that action plans can be developed.
4.8.5. MANAGE completion of reload analysis and licensing support. (Core Designer)
: 1.     COORDINATE the scheduling, monitoring, completion and review of all reload support work.
: 2.     ENSURE that all identified reload design and licensing scope is being addressed by scheduled work activities.
: 3.     ENSURE that incomplete analyses are expected to be resolved in a timely manner, or have sufficient contingencies.
: 4.      ADAPT the reload support work scope and schedule, as agreed with key stakeholders, to account for any analysis problems.
: 5.     COORDINATE with site to ensure that the outage work program includes all fuel-related activities.
: 6.     PROVIDE reload core design data to the Operations Training Department for simulator testing. Refer to TQ-AA-303. The ability to supply this data may be impacted by the actions identified in Step 4.3.6.3.
: 7.     ENSURE that all source documentation is produced by appropriate parties (fuel vendor, contractors, NF, etc.) in a timely manner, reviewed, then distributed to appropriate recipients at site.
A.     DOCUMENT the review and acceptance of fuel vendor and contractor documentation in the Core Reload FCP.
: 8.     DOCUMENT the applicable design analyses, UFSAR updates, ADL changes, etc. in the Core Reload FCP.
 
NF-AA-100 Revision 13 Page 23 of 54
: 9.      If significant changes are made to the inputs, assumptions, or methods used in reload analysis and/or licensing, WRITE an IR to document the changes and DISCUSS the changes during future Reload Kickoff meetings, per Attachment 5.
4.8.6. PROVIDE details of any "new" or "non-standard" testing requirements and acceptance test criteria for the reload in the FCP. (Core Designer)
: 1.     Promptly NOTIFY Reactor Engineering of the "new" or "non-standard" testing once its existence becomes known.
4.8.7. PREPARE the reload COLR, based on the results collated at Step 4.8.5. (Core Designer) 4.8.8. PREPARE the reload IOCFR50.59 screening/ evaluation per LS-AA-104. (Core Designer) 4.8.9. PROVIDE information to Site Training Department to support development of training materials for the Core Reload design. (Core Designer)
: 1.     SOLICIT Reactor Engineering and Training input on content of information.
Items to be considered are listed in Attachment 3.
Items to be considered are listed in Attachment 3.
2.PROVIDE information early enough to support completion of personnel training before aspects of the new reload affect duties.
: 2.     PROVIDE information early enough to support completion of personnel training before aspects of the new reload affect duties.
3.IDENTIFY continuing training needs during the cycle if the design results in significant changes during varying stages of core life.
: 3.     IDENTIFY continuing training needs during the cycle if the design results in significant changes during varying stages of core life.
4.DOCUMENT any "new" or "non-standard" training requirements in the Core Reload FCP.
: 4.     DOCUMENT any "new" or "non-standard" training requirements in the Core Reload FCP.
NOTE: A Core Loading Plan consistent with the expected Core Reload design may be provided earlier in the reload process to allow early development of the fuel shuffle sequence.However, the final version used to load andverify the as-loaded core must be consistent with the final core following any redesign.
NOTE:       A Core Loading Plan consistent with the expected Core Reload design may be provided earlier in the reload process to allow early development of the fuel shuffle sequence. However, the final version used to load and verify the as-loaded core must be consistent with the final core following any redesign.
4.8.10.PROVIDE a full-core Core Loading Plan to Reactor Engineering that indicates core-loading locations of all the reload cycle's assemblies/bundles using their vendor-supplied identifiers.(Core Designer) 1.PROVIDE core locations of affected control rod assemblies/control blades, burnable poison assemblies and neutron sources using their vendor-supplied identifiers.
4.8.10. PROVIDE a full-core Core Loading Plan to Reactor Engineering that indicates core-loading locations of all the reload cycle's assemblies/bundles using their vendor-supplied identifiers. (Core Designer)
NF-AA-100 Revision 13 Page 24 of 54 2.OBTAIN documentation from the fuel vendor detailing the as-built characteristics of the reload fuel versus the new assembly/bundle identifiers.
: 1.     PROVIDE core locations of affected control rod assemblies/control blades, burnable poison assemblies and neutron sources using their vendor-supplied identifiers.
A.PROVIDE this information, as necessary, to appropriate personnel to update SNM and fuel movement software.
 
3.ENSURE the Core Loading Plan is consistent with the final core design.
NF-AA-100 Revision 13 Page 24 of 54
4.VERIFY that the Core Loading Plan satisfies any optimization targets specified (e.g., at Step 4.3.8), for instance on as-built new fuel variations; component lifetime management; or fuel move optimization, 5.REVIEW the Core Loading Plan and as-built data prior to implementation.
: 2. OBTAIN documentation from the fuel vendor detailing the as-built characteristics of the reload fuel versus the new assembly/bundle identifiers.
6.TRANSMIT the Core Loading Plan via a TODI. Refer to CC-AA-310.
A.     PROVIDE this information, as necessary, to appropriate personnel to update SNM and fuel movement software.
4.8.11.PROVIDE the following information to the NFDB analyst:
: 3. ENSURE the Core Loading Plan is consistent with the final core design.
: 4. VERIFY that the Core Loading Plan satisfies any optimization targets specified (e.g., at Step 4.3.8), for instance on as-built new fuel variations; component lifetime management; or fuel move optimization,
: 5. REVIEW the Core Loading Plan and as-built data prior to implementation.
: 6. TRANSMIT the Core Loading Plan via a TODI. Refer to CC-AA-310.
4.8.11. PROVIDE the following information to the NFDB analyst:
For BWRs, nominal enrichment for each reload bundle type For PWRs, nominal enrichment for the high enrichment central region of the fuel, excluding axial blankets, for each reload sub-batch. Include U-234 and U-236 enrichment, if applicable.
For BWRs, nominal enrichment for each reload bundle type For PWRs, nominal enrichment for the high enrichment central region of the fuel, excluding axial blankets, for each reload sub-batch. Include U-234 and U-236 enrichment, if applicable.
List of serial numbers for fuel to be discharged List of serial numbers for fuel to be re-inserted from the spent fuel pool, if applicable Rod type map(s) when introducing a new fuel type into the unit, if applicable.(e.g., GE14 to GNF2, LTAs, etc.)
List of serial numbers for fuel to be discharged List of serial numbers for fuel to be re-inserted from the spent fuel pool, if applicable Rod type map(s) when introducing a new fuel type into the unit, if applicable.
Rod type maps for all reload sub-batches containing IFBA rods 4.8.12.PROVIDE the computer files required by the core shuffle simulator (e.g., SHUFFLEWORKS) to Reactor Engineering.(Core Designer) 4.8.13.PROVIDE the FCP review package, reload 50.59 documentation, COLR, and vendor documentation summarizing the acceptability of the reload design to the appropriate members of the RDOT and CONVENE the RDOT, if necessary.(Core Designer)4.8,14.PROVIDE the affected departments (typically the RDOT members) with an FCP review package for use in performing the interfacing review of, and input to the FCP.(Core Designer)1.OBTAIN affected Inter-Department reviews of the FCP. Refer to CC-AA-102 Attachment 9 and 10 Review Checklists. At a minimum, impact reviews should be obtained from all "Required" RDOT members (see Attachment 1) in NF-AA-100 Revision 13 Page 25 of 54 departments other than Nuclear Fuels, as well as from departments that may be affected by unique items in a given reload.
(e.g., GE14 to GNF2, LTAs, etc.)
2.IDENTIFY the procedure changes, deletions, and additions. In Passport, enter affected procedures on the ADL. In PIMS, list the affected procedures in the ECR text or Attachments. If a procedure is required to perform calibration or testing for the FCP, but is only partially implemented, then ASSURE that the procedure changes required are separate or interim procedure revisions. Refer to CC-AA-102, Attachment 9.
Rod type maps for all reload sub-batches containing IFBA rods 4.8.12. PROVIDE the computer files required by the core shuffle simulator (e.g.,
: 3. training changes, deletions, and additions required by Operations in the Special Instructions, and FORWARD to the Training department to support identification and completion of training requirements. Refer to CC-AA-102, Attachment 9.
SHUFFLEWORKS) to Reactor Engineering. (Core Designer) 4.8.13. PROVIDE the FCP review package, reload 50.59 documentation, COLR, and vendor documentation summarizing the acceptability of the reload design to the appropriate members of the RDOT and CONVENE the RDOT, if necessary. (Core Designer) 4.8,14. PROVIDE the affected departments (typically the RDOT members) with an FCP review package for use in performing the interfacing review of, and input to the FCP.
4.8.15.After the receipt of the CC-AA-102 Attachment 9 and 10 Review Checklists, ASSEMBLE the contents of the FCP, SIGN the FCP and PROVIDE it to the assigned reviewer.(Core Designer) 4.8.16.REVIEW and APPROVE the Core Reload FCP. Refer to NF-AA-100-1000. This initial revision of the FCP will be the basis of the fuel load and approval is required prior to reloading the core for the reload cycle.(NF Engineer and Manager, BWRIPWR Design) 4.8.17.PROCESS the FCP package. (Core Designer)1. TRANSMIT the Approved FCP, including FCP Attachments, to the appropriate site Records Management.
(Core Designer)
2.For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-102 Attachment forms, PROVIDE a copy to the appropriate Site EC Coordinator for review, and RETAIN for future transmittal to Records Management. Only the final FCP package including all revisions is transmitted to Records Management.
: 1. OBTAIN affected Inter-Department reviews of the FCP. Refer to CC-AA-102 Attachment 9 and 10 Review Checklists. At a minimum, impact reviews should be obtained from all "Required" RDOT members (see Attachment 1) in
4.8.18.CONFIRM that required activities are done, then BRIEF the Operations Representative on the Core Reload FCP scope and status of the package with respect to its readiness for Ops Acceptance. Required activities include the completion of Work Orders related to the FCP revision, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance.(Reactor Engineer and Site EC Coordinator) 1.In Passport, CHANGE the status of the FCP to MODIFIED.
 
NF-AA-100 Revision 13 Page 25 of 54 departments other than Nuclear Fuels, as well as from departments that may be affected by unique items in a given reload.
: 2.       IDENTIFY the procedure changes, deletions, and additions. In Passport, enter affected procedures on the ADL. In PIMS, list the affected procedures in the ECR text or Attachments. If a procedure is required to perform calibration or testing for the FCP, but is only partially implemented, then ASSURE that the procedure changes required are separate or interim procedure revisions. Refer to CC-AA-102, Attachment 9.
IDENTIFY    3.                   training changes, deletions, and additions required by Operations in the Special Instructions, and FORWARD to the Training department to support identification and completion of training requirements. Refer to CC-AA-102, Attachment 9.
4.8.15. After the receipt of the CC-AA-102 Attachment 9 and 10 Review Checklists, ASSEMBLE the contents of the FCP, SIGN the FCP and PROVIDE it to the assigned reviewer. (Core Designer) 4.8.16. REVIEW and APPROVE the Core Reload FCP. Refer to NF-AA-100-1000. This initial revision of the FCP will be the basis of the fuel load and approval is required prior to reloading the core for the reload cycle. (NF Engineer and Manager, BWRIPWR Design) 4.8.17. PROCESS the FCP package. (Core Designer)
For PIMS    1. ECRs, TRANSMIT the Approved FCP, including FCP Attachments, to the appropriate site Records Management.
: 2.       For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-102 Attachment forms, PROVIDE a copy to the appropriate Site EC Coordinator for review, and RETAIN for future transmittal to Records Management. Only the final FCP package including all revisions is transmitted to Records Management.
4.8.18. CONFIRM that required activities are done, then BRIEF the Operations Representative on the Core Reload FCP scope and status of the package with respect to its readiness for Ops Acceptance. Required activities include the completion of Work Orders related to the FCP revision, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance. (Reactor Engineer and Site EC Coordinator)
: 1.     In Passport, CHANGE the status of the FCP to MODIFIED.
2.
2.
the Cycle Management FCP to CLOSED -
Do not CHANGE              the status of the Cycle Management FCP to CLOSED -
Passport, COMPLT - PIMS.
Passport, COMPLT - PIMS.
NF-AA-100 Revision 13 Page 26 of 54 4.8.19.PROCESS the Core Reload COLR (Refer to AD-AA-102, Station Qualified Reviewer, and NF-AB-120-3600, COLR Generation - BWR Units) and 50.59 documentation (Refer to LS-AA-104), as required. The Core Designer should ensure the following actions are completed.
 
1.TRANSMIT the COLR to Station Reactor Engineering (Core Designer)2.PERFORM SQR review of the COLR (Station Reactor Engineering) 3.APPROVE the COLR (SEAM)4.TRANSMIT the COLR to Station Document Services (Station Reactor Engineering) 5.DISTRIBUTE the COLR to appropriate station personnel upon Station Reactor Engineering notification of issuance prior to startup (Station Document Services) 6.TRANSMIT the COLR to the NRC (Station Regulatory Assurance) 4.8.20.ENSURE all appropriate analyses/activities supporting the fuel load are complete prior to the loading of fuel with significant design changes or new fuel designs.(Core Designer)1.If Refueling and Core Verification activities are required prior to the Reload Licensing and/or Core Loading Plan approval in the Core Reload FCP, then PROCESS an Advance Work Authorization (AWA) prior to the activities.
NF-AA-100 Revision 13 Page 26 of 54 4.8.19. PROCESS the Core Reload COLR (Refer to AD-AA-102, Station Qualified Reviewer, and NF-AB-120-3600, COLR Generation - BWR Units) and 50.59 documentation (Refer to LS-AA-104), as required. The Core Designer should ensure the following actions are completed.
: 1. TRANSMIT the COLR to Station Reactor Engineering (Core Designer)
: 2.     PERFORM SQR review of the COLR (Station Reactor Engineering)
: 3.     APPROVE the COLR (SEAM)
: 4.     TRANSMIT the COLR to Station Document Services (Station Reactor Engineering)
: 5.     DISTRIBUTE the COLR to appropriate station personnel upon Station Reactor Engineering notification of issuance prior to startup (Station Document Services)
: 6.     TRANSMIT the COLR to the NRC (Station Regulatory Assurance) 4.8.20. ENSURE all appropriate analyses/activities supporting the fuel load are complete prior to the loading of fuel with significant design changes or new fuel designs.
(Core Designer)
: 1.     If Refueling and Core Verification activities are required prior to the Reload Licensing and/or Core Loading Plan approval in the Core Reload FCP, then PROCESS an Advance Work Authorization (AWA) prior to the activities.
Refer to NF-AA-1 00-1000.
Refer to NF-AA-1 00-1000.
4.9.Refueling and Core Verification 4.9.1.PERFORM all fuel-specific work items, as agreed in the outage work program or necessitated by emergent issues, according to identified procedures.(Site)1.PREPARE procedures for fuel activities including offload, repair, movement, reload, component shuffling, etc. to assure that Fuel Reliability, Reactivity Management, vendor handling and SNM tracking requirements are met.
4.9. Refueling and Core Verification 4.9.1. PERFORM all fuel-specific work items, as agreed in the outage work program or necessitated by emergent issues, according to identified procedures. (Site)
2.Procedures shall ASSURE that no changes to fuel or core configurations shall be performed until appropriate prerequisites are satisfied.
: 1.     PREPARE procedures for fuel activities including offload, repair, movement, reload, component shuffling, etc. to assure that Fuel Reliability, Reactivity Management, vendor handling and SNM tracking requirements are met.
4.9.2.INITIATE any surveillance required to identify fuel failures or diagnose other fuel
: 2.     Procedures shall ASSURE that no changes to fuel or core configurations shall be performed until appropriate prerequisites are satisfied.
4.9.2. INITIATE any surveillance required to identify fuel failures or diagnose other fuel performance issues (e.g., foreign material, cladding performance, spacer grid damage, channel bow, control rod/blade condition, etc.) in re-use, re-insert or other offloaded fuel. (Site)
 
NF-AA-100 Revision 13 Page 27 of 54 PERFORM additional surveillance to either justify the re-insertion of particular (re-use or re-insert) fuel, or support company assessments on the suitability of the fuel design, as required.
4.9.3. LOAD the core according to the final Core Loading Plan (including core component loading requirements). (Site andlor Outage Services)
: 1.      PERFORM an independent visual verification of the loaded locations/orientations and correct physical seating of all fuel and control components as prescribed by fuel handling/SNM procedures.
4.9.4. INITIATE a revision to the Core Reload FCP to incorporate final configuration changes (Cycle Specific Physics Data, Core Monitoring Database, Cycle Management Report, etc). (Core Designer)
NOTE:      Prioritization of objectives for a redesign and significant changes resulting from a redesign shall be subject to the approvals consistent with the earlier steps of this Procedure. The amount of re-work shall depend on the perturbation from the FFCD, which should be minimized. If possible, redesigns should be done such that the initial FFCD licensing analyses remain bounding.
4.9.5. If a core redesign is necessary due to unanticipated reasons (e.g., a large shift in the outage schedule, fuel failure/damage, etc.), then INITIATE redesign activities.
(Core Designer)
: 1.      DEVELOP a redesign plan and schedule.
: 2.      ENSURE that the implemented reload design satisfies the requirements of this Procedure and other applicable procedures, prior to releasing redesign information to site.
: 3.      Following the completion of all prior Steps of the redesign work, DOCUMENT the changes in the Core Reload FCP and REPEAT the appropriate Steps of this procedure affected by the redesign as a consistent set.
4.9.6. Prior to start- up, REVIEW the applicability of all assumptions and analyses on which the acceptability of the reload core is based. (Core Designer and Reactor Engineer)
: 1. CHECK both that the plant state is as assumed in the reload analyses and that the as-loaded core is consistent with all analyses.
: 2. VERIFY all activities required by the CDIR or by NF-AA-101, if required, have been completed.


performance issues (e.g., foreign material, cladding performance, spacer grid damage, channel bow, control rod/blade condition, etc.) in re-use, re-insert or other offloaded fuel.(Site)
NF-AA-100 Revision 13 Page 28 of 54
NF-AA-100 Revision 13 Page 27 of 54 PERFORM additional surveillance to either justify the re-insertion of particular (re-use or re-insert) fuel, or support company assessments on the suitability of the fuel design, as required.4.9.3.LOAD the core according to the final Core Loading Plan (including core component loading requirements).(Site andlor O utage Services) 1.PERFORM an independent visual verification of the loaded locations/orientations and correct physical seating of all fuel and control components as prescribed by fuel handling/SNM procedures.4.9.4.INITIATE a revision to the Core Reload FCP to incorporate final configuration changes (Cycle Specific Physics Data, Core Monitoring Database, Cycle Management Report, etc). (Core Designer)NOTE: Prioritization of objectives for a redesign and significant changes resulting from a redesign shall be subject to the approvals consistent with the earlier steps of this Procedure.
: 3. CONVENE review meetings as necessary to complete this step.
The amount of re-work shall depend on the perturbation from the FFCD, which should be minimized. Ifpossible, redesigns should be done such that the initial FFCD licensing analyses remain bounding.4.9.5.If a core redesign is necessary due to unanticipated reasons (e.g., a large shift in the outage schedule, fuel failure/damage, etc.), then INITIATE redesign activities.(Core Designer)1.DEVELOP a redesign plan and schedule.
: 4. If a redesign has taken place, then ASSURE that all necessary steps of this procedure have been repeated.
2.ENSURE that the implemented reload design satisfies the requirements of this Procedure and other applicable procedures, prior to releasing redesign information to site.
: 5. CONFIRM that the 10CFR50.59 screening/evaluation has been completed, and that reload information has been presented to and reviewed by the site PORC and NSRB as required by applicable procedures. Refer to LS-AA-1 04 and LS-AA-106.
3.Following the completion of all prior Steps of the redesign work, DOCUMENT the changes in the Core Reload FCP and REPEAT the appropriate Steps of this procedure affected by the redesign as a consistent set.4.9.6.Prior to start-up, REVIEW the applicability of all assumptions and analyses on which the acceptability of the reload core is based.(Core Designer and Reactor Engineer)1.CHECK both that the plant state is as assumed in the reload analyses and that the as-loaded core is consistent with all analyses.
: 6. REVIEW the Risk Management Assessment, and MODIFY if necessary. The RDOT, RRB, and SMDI shall approve of significant changes to the Risk Management Assessment as directed by NF-AA-1 00-1600.
2.VERIFY all activities required by the CDIR or by NF-AA-101, if required, have been completed.
4.9.7. PROVIDE the departments affected by the FCP Revision with an FCP review package for use in performing the interfacing review of, and input to the FCP Revision. (Core Designer)
NF-AA-100 Revision 13 Page 28 of 54 3.CONVENE review meetings as necessary to complete this step.
: 1. OBTAIN affected Inter-Department reviews of the FCP by utilizing the CC-AA-102 Attachment 10 Review Checklists. At a minimum, impact reviews should be obtained from Reactor Engineering and Operations.
4.If a redesign has taken place, then ASSURE that all necessary steps of this procedure have been repeated.
4.9.8. After the receipt of the CC-AA-1 02 Attachment 10 Review Checklists, ASSEMBLE the contents of the FCP, SIGN the FCP and PROVIDE it to the assigned reviewer.
5.CONFIRM that the 10CFR50.59 screening/evaluation has been completed, and that reload information has been presented to and reviewed by the site PORC and NSRB as required by applicable procedures. Refer to LS-AA-1 04 and LS-AA-106.
(Core Designer) 4.9.9. REVIEW and APPROVE the Core Reload FCP Revision. Refer to NF-AA-1 00-1000. This revision is intended to be the final revision of the FCP and approval is required prior to cycle start- up. (NF Engineer and Manager, BWRIPWR Design) 4.9.10. PROCESS the FCP package. (Core Designer)
6.REVIEW the Risk Management Assessment, and MODIFY if necessary. The RDOT, RRB, and SMDI shall approve of significant changes to the Risk Management Assessment as directed by NF-AA-1 00-1600.
: 1. For PIMS ECRs, TRANSMIT the Approved FCP Revision, including FCP Attachments, to the appropriate site Records Management.
4.9.7.PROVIDE the departments affected by the FCP Revision with an FCP review package for use in performing the interfacing review of, and input to the FCP Revision.(Core Designer) 1.OBTAIN affected Inter-Department reviews of the FCP by utilizing the CC-AA-102 Attachment 10 Review Checklists. At a minimum, impact reviews should be obtained from Reactor Engineering and Operations.
: 2. For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-1 02 Attachment forms, for this revision, PROVIDE a copy to the appropriate Site EC Coordinator for review, and PROVIDE the final FCP package including printouts of all revisions (i.e.,
4.9.8.After the receipt of the CC-AA-1 02 Attachment 10 Review Checklists, ASSEMBLE the contents of the FCP, SIGN the FCP and PROVIDE it to the assigned reviewer.(Core Designer)4.9.9.REVIEW and APPROVE the Core Reload FCP Revision. Refer to NF-AA-1 00-1000. This revision is intended to be the final revision of the FCP and approval is required prior to cycle start-up.(NF Engineer and Manager, BWRIPWR Design) 4.9.10.PROCESS the FCP package.(Core Designer)1.For PIMS ECRs, TRANSMIT the Approved FCP Revision, including FCP Attachments, to the appropriate site Records Management.
dispositions, Passport screen printouts, FCP Attachments, CC-AA-1 02 Attachments, etc. for each revision) to the Site EC Coordinator for processing.
2.For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-1 02 Attachment forms, for this revision, PROVIDE a copy to the appropriate Site EC Coordinator for review, and PROVIDE the final FCP package including printouts of all revisions (i.e., dispositions, Passport screen printouts, FCP Attachments, CC-AA-1 02 Attachments, etc. for each revision) to the Site EC Coordinator for processing.
4.9.11. VERIFY that all necessary documentation has been supplied and that updates are prepared, ready for start-up. (Site) This includes:
4.9.11.VERIFY that all necessary documentation has been supplied and that updates are prepared, ready for start-up. (Site) This includes:
All site manuals and procedures (including TS and COLR)
All site manuals and procedures (including TS and COLR)
NF-AA-100 Revision 13 Page 29 of 54 Incorporation of cycle-specific physics data into site procedures or databook Incorporation of reload design acceptance criteria into site startup and core management procedures The Core Monitoring System Database The training simulator and other training materials Design Basis Database (e.g., ATLAS, DBdb, etc.)
NF-AA-100 Revision 13 Page 29 of 54 Incorporation of cycle-specific physics data into site procedures or databook Incorporation of reload design acceptance criteria into site startup and core management procedures The Core Monitoring System Database The training simulator and other training materials Design Basis Database (e.g., ATLAS, DBdb, etc.)
4.9.12.IMPLEMENT the updates of site material (documentation, procedures, computer databases, etc.) including updates necessary for any redesign impact.(Site)4.9.13.PROVIDE all necessary pre-startup information to USNRC.(Licensing and Regulatory Affairs)4.9.14.CONFIRM that required activities are done, then BRIEF the Operations Representative on the Core Reload FCP scope, completion, and final status of the package with respect to its readiness for Ops Acceptance. Required activities include the completion of Work Orders related to the FCP, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance.(Reactor Engineer and Site EC Coordinator) 4.9.15.If Operations has been satisfied that the Core Reload FCP scope is acceptable, then CHANGE the status of the Core Reload FCP to reflect that all changes and associated fuel moves are complete (CLOSED -- Passport, COMPLT - RIMS). (Core Designer or Site EC Coordinator) 4.9.16.PERFORM a Post-Job Brief for the Core Reload after the start-up of the refueled unit.Refer to HU-AA-1212.(Core Designer and Manager, BWRIPWR Design) 1.ENSURE participation by appropriate RDOT members and the vendor.
4.9.12. IMPLEMENT the updates of site material (documentation, procedures, computer databases, etc.) including updates necessary for any redesign impact. (Site) 4.9.13. PROVIDE all necessary pre-startup information to USNRC. (Licensing and Regulatory Affairs) 4.9.14. CONFIRM that required activities are done, then BRIEF the Operations Representative on the Core Reload FCP scope, completion, and final status of the package with respect to its readiness for Ops Acceptance. Required activities include the completion of Work Orders related to the FCP, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance. (Reactor Engineer and Site EC Coordinator) 4.9.15. If Operations has been satisfied that the Core Reload FCP scope is acceptable, then CHANGE the status of the Core Reload FCP to reflect that all changes and associated fuel moves are complete (CLOSED -- Passport, COMPLT - RIMS). (Core Designer or Site EC Coordinator) 4.9.16. PERFORM a Post-Job Brief for the Core Reload after the start-up of the refueled unit. Refer to HU-AA- 1212. (Core Designer and Manager, BWRIPWR Design)
4.10.Core Operation and Monitoring NOTE: When significant changes to the core design have been made, reload-specific tests or surveillances may be specified.
: 1.     ENSURE participation by appropriate RDOT members and the vendor.
4.10.1.PERFORM any BOC tests or surveillances required to demonstrate that the Core Reload behaves as predicted (and therefore that the reload analyses are applicable to the as-loaded core).(Site)
4.10. Core Operation and Monitoring NOTE:     When significant changes to the core design have been made, reload-specific tests or surveillances may be specified.
NF-AA-100 Revision 13 Page 30 of 54 1.If any test or surveillance criteria are violated, then REFER to the applicable procedures, which shall clearly prescribe the actions to be taken and define whether continued operation is acceptable.
4.10.1. PERFORM any BOC tests or surveillances required to demonstrate that the Core Reload behaves as predicted (and therefore that the reload analyses are applicable to the as-loaded core). (Site)
4.10.2.PROVIDE all necessary post-start-up reload documentation to the USNRC, for example: COLR (if not provided at Step 4.9.13); start-up report, 10CFR50.46 documentation, etc.(Licensing and Regulatory Affairs) 4.10.3.If required by station procedures, then INITIATE an independent offsite review of the reload design and 10CFR50.59 screening/evaluation following the start-up of the cycle.(NF)4.10.4.OPERATE the core according to procedures that ensure that NF core management guidance, required to preserve analysis assumptions or to support fuel performance objectives, is followed.(Site)(CM-11)1.MONITOR the core performance against defined acceptance criteria.
 
NF-AA-100 Revision 13 Page 30 of 54
: 1.     If any test or surveillance criteria are violated, then REFER to the applicable procedures, which shall clearly prescribe the actions to be taken and define whether continued operation is acceptable.
4.10.2. PROVIDE all necessary post-start-up reload documentation to the USNRC, for example: COLR (if not provided at Step 4.9.13); start-up report, 10CFR50.46 documentation, etc. (Licensing and Regulatory Affairs) 4.10.3. If required by station procedures, then INITIATE an independent offsite review of the reload design and 10CFR50.59 screening/evaluation following the start-up of the cycle. (NF) 4.10.4. OPERATE the core according to procedures that ensure that NF core management guidance, required to preserve analysis assumptions or to support fuel performance objectives, is followed. (Site) (CM-11)
: 1.     MONITOR the core performance against defined acceptance criteria.
Monitored information should include but not be limited to: (Site and/or NF)
Monitored information should include but not be limited to: (Site and/or NF)
Power distributions versus prediction and evaluation limits
Power distributions versus prediction and evaluation limits
-Core reactivity versus prediction and evaluation limits
                -       Core reactivity versus prediction and evaluation limits
-Fuel exposure versus prediction and evaluation limits
                -       Fuel exposure versus prediction and evaluation limits
-Margin to thermal limits (e.g., LHGR, APLHGR, MCPR, etc.) and operational limits (setpoints, rod limits, etc.)
                -       Margin to thermal limits (e.g., LHGR, APLHGR, MCPR, etc.) and operational limits (setpoints, rod limits, etc.)
-Accuracy of maneuver predictions
                -       Accuracy of maneuver predictions
-Coolant chemistry control
                -       Coolant chemistry control
-Fuel integrity/coolant activity
                -       Fuel integrity/coolant activity
-Reactivity management items
                -       Reactivity management items
-Trends which erode margin to operating goals
                -       Trends which erode margin to operating goals
-Anomalies-Any additional cycle-specific items determined under this process 2.DOCUMENT core monitoring results according to unit procedures.(Reactor Engineering)
                -       Anomalies
A.HIGHLIGHT all anomalous conditions and trends that threaten operating goals or safety margins and COMMUNICATE to key stakeholders.
                -       Any additional cycle-specific items determined under this process
Refer to LS-AA-125.
: 2.     DOCUMENT core monitoring results according to unit procedures. (Reactor Engineering)
3.DOCUMENT significant problems identified according to the corporate corrective action process.(Site and Core Designer)
A.       HIGHLIGHT all anomalous conditions and trends that threaten operating goals or safety margins and COMMUNICATE to key stakeholders. Refer to LS-AA-125.
NF-AA-100 Revision 13 Page 31 of 54 A.Where appropriate, SHARE anomaly data and experience with the fuel vendor and DISSEMINATE as industry Operating Experience (CM-16)4.10.5.USE relevant results of the monitoring program as inputs to the following cycle Reload design.(Core Designer)NOTE: It is not necessary to open a Cycle Management FCP for a given operating cycle unless a configuration change is to be incorporated.
: 3.     DOCUMENT significant problems identified according to the corporate corrective action process. (Site and Core Designer)
4.10.6.DOCUMENT Cycle Management configuration changes (revisions to COLR, 50.59 screening/evaluation, Cycle Management Report, etc.) on a Cycle Management FCP.(Core Designer) 1.INITIATE an FCP type EC/ECR to control the Cycle Management configuration changes. Refer to NF-AA-100-1000. (Core Designer)2.PERFORM a Technical Pre-Job Brief for the Cycle Management design effort.Refer to HU-AA-1212. (Core Designer and Manager, BWR/PWR Design)A.DISCUSS NF, Inter-Department and potential Independent Third Party reviews. The Pre-Job Brief is expected to be a face
 
-to-face discussion with the Manager, BWRIPWR Design, taking the appropriate questioning
NF-AA-100 Revision 13 Page 31 of 54 A.     Where appropriate, SHARE anomaly data and experience with the fuel vendor and DISSEMINATE as industry Operating Experience (CM-16) 4.10.5. USE relevant results of the monitoring program as inputs to the following cycle Reload design. (Core Designer)
/challenging role.
NOTE:     It is not necessary to open a Cycle Management FCP for a given operating cycle unless a configuration change is to be incorporated.
3.INCLUDE applicable design considerations and impacts using the Design Attribute Review (DAR). Refer to CC-AA-1 02 Attachment 1. (Core Designer)NOTE: The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per a regulatory processes such as 50.59, 50.90, 50.46, etc.
4.10.6. DOCUMENT Cycle Management configuration changes (revisions to COLR, 50.59 screening/evaluation, Cycle Management Report, etc.) on a Cycle Management FCP. (Core Designer)
4.REVIEW the Risk Management Assessment for the cycle of interest and update as necessary based on the Cycle Management changes at hand.(Core Designer)CONSIDER in the assessment any changes to the vendor or Exelon methods that support core operation.
: 1. INITIATE an FCP type EC/ECR to control the Cycle Management configuration changes. Refer to NF-AA-100-1000. (Core Designer)
: 2. PERFORM a Technical Pre-Job Brief for the Cycle Management design effort. Refer to HU-AA-1212. (Core Designer and Manager, BWR/PWR Design)
A.       DISCUSS NF, Inter-Department and potential Independent Third Party reviews. The Pre-Job Brief is expected to be a face -to-face discussion with the Manager, BWRIPWR Design, taking the appropriate questioning/challenging role.
: 3. INCLUDE applicable design considerations and impacts using the Design Attribute Review (DAR). Refer to CC-AA-1 02 Attachment 1. (Core Designer)
NOTE:     The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per a regulatory processes such as 50.59, 50.90, 50.46, etc.
: 4. REVIEW the Risk Management Assessment for the cycle of interest and update as necessary based on the Cycle Management changes at hand.
(Core Designer)
CONSIDER in the assessment any changes to the vendor or Exelon methods that support core operation.
ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.
ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.
ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.


ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.
NF-AA-100 Revision 13 Page 32 of 54
NF-AA-100 Revision 13 Page 32 of 54
-DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).5.If issues arise during cycle operation which cause significant changes to the inputs, assumptions, or methods used in reload analysis and/or licensing, WRITE an IR to document the changes and DISCUSS the changes during future Reload Kickoff meetings, per Attachment 5. (Core Designer)REVISE necessary documentation, as applicable, and PERFORM the appropriate steps defined in NF and other procedures for the Cycle Management changes at hand.(Core Designer)7.PROVIDE the departments affected by the Cycle Management FCP Revision with an FCP review package for use in performing the interfacing review of, and input to the FCP Revision. (Core Designer)A.IDENTIFY the procedure changes, deletions, and additions. In Passport, enter affected procedures on the ADL. In PIMS, list the affected procedures in the ECR text or Attachments. Refer to CC-AA-102, Attachment 9.
    -       DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).
B.IDENTIFY training changes, deletions, and additions required by Operations in the Special Instructions, and FORWARD to the Training department to support identification and completion of training requirements. Refer to CC-AA-1 02, Attachment 9.
: 5. If issues arise during cycle operation which cause significant changes to the inputs, assumptions, or methods used in reload analysis and/or licensing, WRITE an IR to document the changes and DISCUSS the changes during future Reload Kickoff meetings, per Attachment 5. (Core Designer)
C.OBTAIN affected Inter-Department reviews of the Cycle Management FCP. Refer to CC-AA-102 Attachment 10 Review Checklists. At a minimum, impact reviews should be obtained from Reactor Engineering and Operations.8.After the receipt of the CC-AA-1 02 Attachment 9 and 10 Review Checklists, ASSEMBLE the contents of the Cycle Management FCP, SIGN the FCP and PROVIDE it to the assigned reviewer.(Core Designer)9.REVIEW and APPROVE the Cycle Management FCP. Refer to NF-AA-1 00
REVISE necessary documentation, as applicable, and PERFORM the appropriate steps defined in NF and other procedures for the Cycle Management changes at hand. (Core Designer)
-1000. The FCP will be the basis for the current Cycle Management configuration changes.(NF Engineer and Manager, BWRIPWR Design) 10.PROCESS the FCP package.(Core Designer)A.For PIMS ECRs, TRANSMIT the Approved FCP, including FCP Attachments, to the appropriate site Records Management.
: 7. PROVIDE the departments affected by the Cycle Management FCP Revision with an FCP review package for use in performing the interfacing review of, and input to the FCP Revision. (Core Designer)
B.For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-102 Attachment forms, PROVIDE a copy to the appropriate Site EC Coordinator for NF-AA-100 Revision 13 Page 33 of 54 review, and RETAIN for transmittal of the final FCP package to Records Management.
A.     IDENTIFY the procedure changes, deletions, and additions. In Passport, enter affected procedures on the ADL. In PIMS, list the affected procedures in the ECR text or Attachments. Refer to CC-AA-102, Attachment 9.
C.For Passport ECs, TRANSMIT the final FCP package including printouts of all revisions to the appropriate Site EC Coordinator for processing.
B.     IDENTIFY training changes, deletions, and additions required by Operations in the Special Instructions, and FORWARD to the Training department to support identification and completion of training requirements. Refer to CC-AA-1 02, Attachment 9.
NOTE: Cycle Management FCPs can consist of documentation changes (e.g., core model changes, 50.46 reports, UFSAR changes), operating procedure changes (e.g., COLR changes) and physical changes to the plant (e.g., mid-cycle re-design, core monitoring system software/database change).Ops Acceptance of the Cycle Management FCP is only required when operating procedure changes or physical changes to the plant are made. Required activities include the completion of Work Orders related to the FCP revision, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance.
C.     OBTAIN affected Inter-Department reviews of the Cycle Management FCP. Refer to CC-AA-102 Attachment 10 Review Checklists. At a minimum, impact reviews should be obtained from Reactor Engineering and Operations.
11.If Ops Acceptance is required for the FCP revision, CONFIRM that required activities are done, then BRIEF the Operations Representative on the Cycle Management FCP scope and status of the package with respect to its readiness for Ops Acceptance. (Reactor Engineer and Site EC Coordinator)
: 8. After the receipt of the CC-AA-1 02 Attachment 9 and 10 Review Checklists, ASSEMBLE the contents of the Cycle Management FCP, SIGN the FCP and PROVIDE it to the assigned reviewer. (Core Designer)
A.In Passport, CHANGE the status of the FCP to MODIFIED.
: 9. REVIEW and APPROVE the Cycle Management FCP. Refer to NF-AA-1 00 -
B.Do not CHANGE the status of the Cycle Management FCP to CLOSED - Passport, COMPLT - PIMS.
1000. The FCP will be the basis for the current Cycle Management configuration changes. (NF Engineer and Manager, BWRIPWR Design)
12.PERFORM a Post-Job Brief for the Cycle Management changes after the changes have been implemented. Refer to HU-AA-1212.(Core Designer and Manager, BWRIPWR Design)13.REVISE the Cycle Management FCP, as required, for subsequent changes throughout the balance of the cycle. (Core Designer)14.At the end of the operating cycle, CHANGE the status of the Cycle Management FCP to reflect that all Cycle Management configuration changes are complete (CLOSED - Passport, COMPLT - PIMS). Ops Acceptance is not required to close out the Cycle Management FCP at the end of the operating cycle.(Core Designer or Site EC Coordinator)
: 10. PROCESS the FCP package. (Core Designer)
NF-AA-100 Revision 13 Page 34 of 54 4.11.FCP Revisions 4.11.1.INITIATE a revision to the FCP in Passport or PIMS, as appropriate.(Core Designer)NOTE: NF Director authorization is required for an Advanced Work Authorization (AWA) when the functional design, design criteria of the plant, or the 10CFR50.59 Evaluation are affected by the FCP revision.
A.     For PIMS ECRs, TRANSMIT the Approved FCP, including FCP Attachments, to the appropriate site Records Management.
An NF Manager can authorize other AWAs.
B.     For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-102 Attachment forms, PROVIDE a copy to the appropriate Site EC Coordinator for
4.11.2.If Advanced Work Authorization (AWA) to proceed at risk with installation work activities in the field without issuance of a completely approved configuration change is required and the responsible NF Manager is sure that the advance work does not affect any in-service equipment and the equipment will not be placed in service and will not be declared operable, and will have no impact on operating or in-service equipment until the configuration change is approved, then: (Core Designer)1.DOCUMENT the following information in the Revision Summary area of the FCP (Topic Notes Panel - Passport, Page 2 - PIMS)
 
A statement that an AWA is being used Scope of Authorized Work, including markups as needed Justification the advance work does not affect any in-service equipment and the equipment will not be placed in service and will not be declared operable, and will have no impact on operating or in
NF-AA-100 Revision 13 Page 33 of 54 review, and RETAIN for transmittal of the final FCP package to Records Management.
-service equipment until the configuration change is approved Special Instructions Name of Core Designer that is giving the AWA Name of the Recipient (Reactor Engineer, etc.) of the AWA Name of NF Management personnel that is authorizing the AWA Date of the AWA 2.INDICATE in the FCP that AWA is going to be used:
C.       For Passport ECs, TRANSMIT the final FCP package including printouts of all revisions to the appropriate Site EC Coordinator for processing.
A.ADD an AWA Milestone for Passport or ADD a Y in the AWA field at the top of Page 2 for PIMS.
NOTE:     Cycle Management FCPs can consist of documentation changes (e.g., core model changes, 50.46 reports, UFSAR changes), operating procedure changes (e.g., COLR changes) and physical changes to the plant (e.g., mid-cycle re-design, core monitoring system software/database change).
NF-AA-100 Revision 13 Page 35 of 54 4.11.3.SUMMARIZE revision changes in the "REVISION  
Ops Acceptance of the Cycle Management FCP is only required when operating procedure changes or physical changes to the plant are made. Required activities include the completion of Work Orders related to the FCP revision, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance.
: 11. If Ops Acceptance is required for the FCP revision, CONFIRM that required activities are done, then BRIEF the Operations Representative on the Cycle Management FCP scope and status of the package with respect to its readiness for Ops Acceptance. (Reactor Engineer and Site EC Coordinator)
A.       In Passport, CHANGE the status of the FCP to MODIFIED.
B.       Do not CHANGE the status of the Cycle Management FCP to CLOSED - Passport, COMPLT - PIMS.
: 12. PERFORM a Post-Job Brief for the Cycle Management changes after the changes have been implemented. Refer to HU-AA-1212. (Core Designer and Manager, BWRIPWR Design)
: 13. REVISE the Cycle Management FCP, as required, for subsequent changes throughout the balance of the cycle. (Core Designer)
: 14. At the end of the operating cycle, CHANGE the status of the Cycle Management FCP to reflect that all Cycle Management configuration changes are complete (CLOSED - Passport, COMPLT - PIMS). Ops Acceptance is not required to close out the Cycle Management FCP at the end of the operating cycle. (Core Designer or Site EC Coordinator)
 
NF-AA-100 Revision 13 Page 34 of 54 4.11. FCP Revisions 4.11.1. INITIATE a revision to the FCP in Passport or PIMS, as appropriate. (Core Designer)
NOTE:       NF Director authorization is required for an Advanced Work Authorization (AWA) when the functional design, design criteria of the plant, or the 10CFR50.59 Evaluation are affected by the FCP revision. An NF Manager can authorize other AWAs.
4.11.2. If Advanced Work Authorization (AWA) to proceed at risk with installation work activities in the field without issuance of a completely approved configuration change is required and the responsible NF Manager is sure that the advance work does not affect any in-service equipment and the equipment will not be placed in service and will not be declared operable, and will have no impact on operating or in-service equipment until the configuration change is approved, then: (Core Designer)
: 1.     DOCUMENT the following information in the Revision Summary area of the FCP (Topic Notes Panel - Passport, Page 2 - PIMS)
A statement that an AWA is being used Scope of Authorized Work, including markups as needed Justification the advance work does not affect any in-service equipment and the equipment will not be placed in service and will not be declared operable, and will have no impact on operating or in -
service equipment until the configuration change is approved Special Instructions Name of Core Designer that is giving the AWA Name of the Recipient (Reactor Engineer, etc.) of the AWA Name of NF Management personnel that is authorizing the AWA Date of the AWA
: 2.     INDICATE in the FCP that AWA is going to be used:
A.     ADD an AWA Milestone for Passport or ADD a Y in the AWA field at the top of Page 2 for PIMS.
 
NF-AA-100 Revision 13 Page 35 of 54 4.11.3. SUMMARIZE revision changes in the "REVISION  


==SUMMARY==
==SUMMARY==
".(Core Designer) 1.IDENTIFY any affected documents (i.e., documents on the ADL) in the"REVISION  
". (Core Designer)
: 1.     IDENTIFY any affected documents (i.e., documents on the ADL) in the "REVISION  


==SUMMARY==
==SUMMARY==
" for the respective revision in which they were changed.4.11.4.REVISE the portions of the FCP that are affected, including disposition notes, package attributes (from use of CC-AA-102), package milestones, ADL, etc. as needed to reflect the change.(Core Designer) 1.DISCUSS change with the interfacing departments that may be affected. The DAR (CC-AA-102 Attachment 1) is used to determine which groups are affected.4.11.5.If affected, then DOCUMENT the review and acceptability of the revision by providing CC-AA-102 Attachment 10 Forms or electronically signing the FCP revision. (Affected Interfacing Departments)
" for the respective revision in which they were changed.
NOTE: The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per a regulatory processes such as 50.59, 50.90, 50.46, etc.
4.11.4. REVISE the portions of the FCP that are affected, including disposition notes, package attributes (from use of CC-AA-102), package milestones, ADL, etc. as needed to reflect the change. (Core Designer)
4.11.6.REVIEW the Risk Management Assessment for the cycle of interest and update as necessary based on the changes at hand.(Core Designer)
: 1.     DISCUSS change with the interfacing departments that may be affected. The DAR (CC-AA-102 Attachment 1) is used to determine which groups are affected.
-CONSIDER in the assessment any changes to the vendor or Exelon methods that support core operation.
4.11.5. If affected, then DOCUMENT the review and acceptability of the revision by providing CC-AA-102 Attachment 10 Forms or electronically signing the FCP revision. (Affected Interfacing Departments)
-ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.
NOTE:       The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per a regulatory processes such as 50.59, 50.90, 50.46, etc.
4.11.6. REVIEW the Risk Management Assessment for the cycle of interest and update as necessary based on the changes at hand. (Core Designer)
        -       CONSIDER in the assessment any changes to the vendor or Exelon methods that support core operation.
        -       ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.
ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.
ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.
-DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).
        -       DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).
4.11.7.REVIEW, APPROVE and PROCESS the FCP in accordance with appropriate sections above for the revision scope. This may require different reviews than those required in previous FCP revisions.(Core Designer)
4.11.7. REVIEW, APPROVE and PROCESS the FCP in accordance with appropriate sections above for the revision scope. This may require different reviews than those required in previous FCP revisions. (Core Designer)
NF-AA-100 Revision 13 Page 36 of 54 4.12.Reload And Cycle Management Decision Making NOTE: The Nuclear Fuels Department many times participates in decision making processes that have been initiated by outside sources, especially the nuclear stations.
 
When this is the case, Step 4.12.1 may be credited as complete via NF participation in the larger process.
NF-AA-100 Revision 13 Page 36 of 54 4.12. Reload And Cycle Management Decision Making NOTE:       The Nuclear Fuels Department many times participates in decision making processes that have been initiated by outside sources, especially the nuclear stations. When this is the case, Step 4.12.1 may be credited as complete via NF participation in the larger process.
4.12.1.APPLY the appropriate decision making process when a technical decision that impacts plant operation is to be made during the reload or cycle management processes.
4.12.1. APPLY the appropriate decision making process when a technical decision that impacts plant operation is to be made during the reload or cycle management processes.
1USE a graded approach to decision making that is dependent on the decision's: (Core Designer/Manager, BWRIPWR Design)
1      USE a graded approach to decision making that is dependent on the decision's: (Core Designer / Manager, BWRIPWR Design)
Risk - possible consequences to the plant or company Number and Level of the stakeholders Distribution of departments represented by the stakeholders (e.g.
Risk - possible consequences to the plant or company Number and Level of the stakeholders Distribution of departments represented by the stakeholders (e.g.
station operations vs. corporate accounting) 2.SCREEN-OUT lower level decisions which should not require a formal decision making process to ensure resources are used as optimally as possible.(Core Designer/Manager, BWRIPWR Design) 3.When available, UTILIZE an existing decision making process such as (but not limited to): (Core Designer)-OTDM - Operational technical decision making process as documented in OP-AA-106-101-1006.
station operations vs. corporate accounting)
-A technical evaluation as documented in accordance with CC-AA-309-101.
: 2. SCREEN -OUT lower level decisions which should not require a formal decision making process to ensure resources are used as optimally as possible. (Core Designer / Manager, BWRIPWR Design)
-The Kepner-Tregoe problem solving and decision making process (trained facilitators and individual users available within Exelon).
: 3. When available, UTILIZE an existing decision making process such as (but not limited to): (Core Designer)
4.12.2.DOCUMENT the use of all formal decision making processes in the reload and cycle management processes in the associated FCP(s) This does not include those items screened-out of the process in Step 4.12.1.2 - though they may need to be included in the FCP per a different procedural requirement.(Core Designer) 4.12.3.ENSURE that all decisions made and documented in the reload and cycle management FCPs have received the proper level of communication and review commensurate with the importance of the decision.(Manager, BWR/PWR Design) 4.12.4.REVIEW all technical decisions made using one of the formal methods above for possible inclusion in the proper reload and/or cycle management risk management assessment(s) per NF-AA-1 00-1600.(Core Designer)
              -       OTDM - Operational technical decision making process as documented in OP-AA-106-101-1006.
NF-AA-100 Revision 13 Page 37 of 54 5.DOCUMENTATION NOTE: Detailed documentation requirements for documents that are attached to or referenced by Core Reload and CycleManagement FCPs are contained in lower level procedures and T&RMs.
              -       A technical evaluation as documented in accordance with CC-AA-309-101.
5.1.TRANSMIT the Core Reload and Cycle Management FCP EC/ECR to Records Management. (Reference RM-AA-101) SRRS# 3B.
              -       The Kepner-Tregoe problem solving and decision making process (trained facilitators and individual users available within Exelon).
107 (Core Designer)6.REFERENCES 6.1.Writer References 6.1.1.1 OCFR50 Appendix B, Criterion III, Design Control 6.1.2.ASME NQA-1, Quality Assurance Program Requirements for Nuclear Facilities 6.1.3.INPO 90-009 (Guideline), March 1990, Guidelines for the Conduct of Design Engineering 6.1.4.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores 6.1.5.INPO SOER 03-02, Managing Core. Design Changes 6.1.6.NF-AA-1, Exelon Nuclear Policy, Fuel Management 6.1.7.NF-AA-3, Exelon Nuclear Policy, Fuel Reliability 6.1.8.NF-AA-10, Reload Control Process Description 6.2.User References 6.2.1.AD-AA-102, Station Qualified Review 6.2.2.CC-AA-102, Design Input and Configuration Change Impact Screening 6.2.3.CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes 6.2.4.CC-AA-309, Control of Design Analyses 6.2.5.CC-AA-310, Transmittal of Design Information 6.2.6.HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Brief 6.2.7.LS-AA-104, Exelon 50.59 Review Process NF-AA-100 Revision 13 Page 38 of 54 6.2.8.LS-AA-106, Plant Operations Review Committee 6.2.9.LS-AA-125, Corrective Action Program (CAP) Procedure 6.2.10.NF-AA-101, Nuclear Fuel Assembly and Core Component Design and Fabrication Process Changes 6.2.11.NF-AA-100-1000, Core Reload and Cycle Management Configuration Changes 6.2.12.NF-AA-100-1010, Use of Operating Experience Information for Nuclear Fuels Work Products 6.2.13.NF-AA-100-1600, Reload and Cycle Operation Risk Management Assessment Instructions 6.2.14.NF-AA-100-2000, Standard RRB and SMDI Presentation Templates 6.2.15.NF-AA-105-1000, Energy Utilization Plan Development 6.2.16.NF-AB-100-4000, GNF Reload Control Implementation 6.2.17.NF-AB-100-5000, Westinghouse Reload Control Implementation for BWRs 6.2.18.NF-AB-100-6000, Framatome Advanced Nuclear Fuel (FANP) Reload Control For BWRs 6.2.19.NF-AB-120-3600, Core Operating Limits Report Generation - BWR Units 6.2.20.NF-AP-100-7000, Westinghouse NSSS Reload Design Control Implementation 6.2.21.NF-AP-100-8000, AREVA PWR Reload Control Implementation 6.2.22.OU-AA-101, Refuel Outage Management 6.2.23.RM-AA-101, Records Management Program 6.2.24.TQ-AA-303, Controlling Simulator Core Updates and Thermal-Hydraulic Model Updates 6.3.Station Commitments 6.3.1.These commitments apply to all Exelon nuclear generating stations 1.CM-1INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03980, Passport 00353329-01) (Step 4.2.1, Attachment 3) 2.CM-2INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03977, Passport 00353329-11) (Step 4.5.2, 4.8.2)
4.12.2. DOCUMENT the use of all formal decision making processes in the reload and cycle management processes in the associated FCP(s) This does not include those items screened-out of the process in Step 4.12.1.2 - though they may need to be included in the FCP per a different procedural requirement. (Core Designer) 4.12.3. ENSURE that all decisions made and documented in the reload and cycle management FCPs have received the proper level of communication and review commensurate with the importance of the decision. (Manager, BWR/PWR Design) 4.12.4. REVIEW all technical decisions made using one of the formal methods above for possible inclusion in the proper reload and/or cycle management risk management assessment(s) per NF-AA-1 00-1600. (Core Designer)
NF-AA-100 Revision 13 Page 39 of 54 3.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03976, Passport 00353329-10) (Step 4.3.8) 4.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03975, Passport 00353329-09) (Step 4.5.4, Attachment 3) 5.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03972, Passport 00353329-07) (Attachment 3) 6.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03971, Passport 00353329-06) (Attachment 3) 7.CM-7INPO SOER 96-02, Design and Operating Considerations for
 
.Reactor Cores (PIMS T03970, Passport 00353329-05) (Attachment 3) 8.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03969, Passport 00353329-04) (Attachment 3) 9.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03964, Passport 00353329-03) (Step 4.5.3, Attachment 2)
NF-AA-100 Revision 13 Page 37 of 54
: 10. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03973, Passport 00353329-08) (Attachment 3) 11.INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03981, Passport 00353329-02) (Step 4.10.4) 12.CM-12 INPO SOER 03-02, Managing Core Design Changes (PIMS T04588, Passport 00353329-12) (Step 4.3.9, 4.4.1) 13.CM-13 INPO SOER 03-02, Managing Core Design Changes (PIMS T04590, Passport 00353329-14) (Attachment 1 and Attachment 3) 14.CM-14 INPO SOER 03-02, Managing Core Design Changes (PIMS T04591, Passport 00353329-15) (Step 4.7.4 and Attachment 3) 15.CM-15 INPO SOER 03-02, Managing Core Design Changes (PIMS T04592, Passport 00353329-16) (Attachment 3) 16.CM-16 INPO SOER 03-02, Managing Core Design Changes (PIMS T04589, Passport 00353329-13) (Step 4.10.4)
: 5. DOCUMENTATION NOTE:     Detailed documentation requirements for documents that are attached to or referenced by Core Reload and Cycle Management FCPs are contained in lower level procedures and T&RMs.
NF-AA-100 Revision 13 Page 40 of 54 7.ATTACHMENTS 7.1.Attachment 1 - Conduct of Reload Design Overview Team (RDOT)
5.1. TRANSMIT the Core Reload and Cycle Management FCP EC/ECR to Records Management. (Reference RM-AA-101) SRRS# 3B. 107 (Core Designer)
Meetings And Team Responsibilities 7.2.Attachment 2 - Cycle Design Inputs and Requirements 7.3.Attachment 3 - Reload Design Review Criteria 7.4.Attachment 4 - Conduct of Reload Review Board and SMDI Meetings 7.5.Attachment 5 - Conduct of Reload Design Kickoff Meetings And Team Responsibilities NF-AA-100 Revision 13 Page 41 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT)
: 6. REFERENCES 6.1. Writer References 6.1.1. 1 OCFR50 Appendix B, Criterion III, Design Control 6.1.2. ASME NQA-1, Quality Assurance Program Requirements for Nuclear Facilities 6.1.3. INPO 90-009 (Guideline), March 1990, Guidelines for the Conduct of Design Engineering 6.1.4. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores 6.1.5. INPO SOER 03-02, Managing Core. Design Changes 6.1.6. NF-AA-1, Exelon Nuclear Policy, Fuel Management 6.1.7. NF-AA-3, Exelon Nuclear Policy, Fuel Reliability 6.1.8. NF-AA-10, Reload Control Process Description 6.2. User References 6.2.1. AD-AA-102, Station Qualified Review 6.2.2. CC-AA-102, Design Input and Configuration Change Impact Screening 6.2.3. CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes 6.2.4. CC-AA-309, Control of Design Analyses 6.2.5. CC-AA-310, Transmittal of Design Information 6.2.6. HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Brief 6.2.7. LS-AA-104, Exelon 50.59 Review Process
Meetings And Team Responsibilities Page 1 of 3 Reload Design Overview Team (RDOT)
 
NF-AA-100 Revision 13 Page 38 of 54 6.2.8. LS-AA-106, Plant Operations Review Committee 6.2.9. LS-AA-125, Corrective Action Program (CAP) Procedure 6.2.10. NF-AA-101, Nuclear Fuel Assembly and Core Component Design and Fabrication Process Changes 6.2.11. NF-AA-100-1000, Core Reload and Cycle Management Configuration Changes 6.2.12. NF-AA-100-1010, Use of Operating Experience Information for Nuclear Fuels Work Products 6.2.13. NF-AA-100-1600, Reload and Cycle Operation Risk Management Assessment Instructions 6.2.14. NF-AA-100-2000, Standard RRB and SMDI Presentation Templates 6.2.15. NF-AA-105-1000, Energy Utilization Plan Development 6.2.16. NF-AB-100-4000, GNF Reload Control Implementation 6.2.17. NF-AB-100-5000, Westinghouse Reload Control Implementation for BWRs 6.2.18. NF-AB-100-6000, Framatome Advanced Nuclear Fuel (FANP) Reload Control For BWRs 6.2.19. NF-AB-120-3600, Core Operating Limits Report Generation - BWR Units 6.2.20. NF-AP-100-7000, Westinghouse NSSS Reload Design Control Implementation 6.2.21. NF-AP-100-8000, AREVA PWR Reload Control Implementation 6.2.22. OU-AA-101, Refuel Outage Management 6.2.23. RM-AA-101, Records Management Program 6.2.24. TQ-AA-303, Controlling Simulator Core Updates and Thermal-Hydraulic Model Updates 6.3. Station Commitments 6.3.1. These commitments apply to all Exelon nuclear generating stations
: 1. CM-1    INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03980, Passport 00353329-01) (Step 4.2.1, Attachment 3)
: 2. CM-2    INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03977, Passport 00353329-11) (Step 4.5.2, 4.8.2)
 
NF-AA-100 Revision 13 Page 39 of 54 CM-3
: 3.            INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03976, Passport 00353329-10) (Step 4.3.8)
CM-4
: 4.             INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03975, Passport 00353329-09) (Step 4.5.4, Attachment 3)
CM-5
: 5.             INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03972, Passport 00353329-07) (Attachment 3)
CM-6
: 6.             INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03971, Passport 00353329-06) (Attachment 3)
: 7. CM-7    INPO SOER 96-02, Design and Operating Considerations for
      . Reactor Cores (PIMS T03970, Passport 00353329-05) (Attachment 3)
CM-8
: 8.             INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03969, Passport 00353329-04) (Attachment 3)
CM-9
: 9.             INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03964, Passport 00353329-03) (Step 4.5.3, Attachment 2)
CM-10
: 10.           INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03973, Passport 00353329-08) (Attachment 3)
CM-11
: 11.           INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03981, Passport 00353329-02) (Step 4.10.4)
: 12. CM-12   INPO SOER 03-02, Managing Core Design Changes (PIMS T04588, Passport 00353329-12) (Step 4.3.9, 4.4.1)
: 13. CM-13   INPO SOER 03-02, Managing Core Design Changes (PIMS T04590, Passport 00353329-14) (Attachment 1 and Attachment 3)
: 14. CM-14   INPO SOER 03-02, Managing Core Design Changes (PIMS T04591, Passport 00353329-15) (Step 4.7.4 and Attachment 3)
: 15. CM-15   INPO SOER 03-02, Managing Core Design Changes (PIMS T04592, Passport 00353329-16) (Attachment 3)
: 16. CM-16   INPO SOER 03-02, Managing Core Design Changes (PIMS T04589, Passport 00353329-13) (Step 4.10.4)
 
NF-AA-100 Revision 13 Page 40 of 54
: 7. ATTACHMENTS 7.1. Attachment 1 - Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities 7.2. Attachment 2 - Cycle Design Inputs and Requirements 7.3. Attachment 3 - Reload Design Review Criteria 7.4. Attachment 4 - Conduct of Reload Review Board and SMDI Meetings 7.5. Attachment 5 - Conduct of Reload Design Kickoff Meetings And Team Responsibilities
 
NF-AA-100 Revision 13 Page 41 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities Page 1 of 3 Reload Design Overview Team (RDOT)
The RDOT is a multi-discipline team made up of NF and site personnel that provides planning, input, and review of reload activities. It assures both that appropriate boundaries for the reload design are established, and that any changes to established boundaries required by the design are acceptable.
The RDOT is a multi-discipline team made up of NF and site personnel that provides planning, input, and review of reload activities. It assures both that appropriate boundaries for the reload design are established, and that any changes to established boundaries required by the design are acceptable.
RDOT members consist of one representative from each of the following departments.
RDOT members consist of one representative from each of the following departments.
Attendance by these members is required for the RDOT meetings in which the CDIR is defined, and the Design Technical Review of the reload is performed. Additional required attendees are identified on a case-by-case basis, considering functional areas that may be affected by significant changes proposed for a specific reload. Depending on the agenda, supplemental RDOT meetings may not require full RDOT attendance. The Core Designer and PWR/BWR Design Manager are responsible for making this determination.
Attendance by these members is required for the RDOT meetings in which the CDIR is defined, and the Design Technical Review of the reload is performed. Additional required attendees are identified on a case-by-case basis, considering functional areas that may be affected by significant changes proposed for a specific reload. Depending on the agenda, supplemental RDOT meetings may not require full RDOT attendance. The Core Designer and PWR/BWR Design Manager are responsible for making this determination.
-NF (the unit Core Designer, who shall chair) [required]
                  -     NF (the unit Core Designer, who shall chair) [required]
-Fuel Reliability Engineer [required]
                  -     Fuel Reliability Engineer [required]
-Engineering Safety Analysis [required]
                  -     Engineering Safety Analysis [required]
-Reactor Engineering (site) [required]
                  -     Reactor Engineering (site) [required]
-Chemistry (site) (CM-13) [required]
                  -     Chemistry (site) (CM-13) [required]
-Radiation Protection (site) [required]
                  -     Radiation Protection (site) [required]
-Operations (site) [required]
                  -     Operations (site) [required]
Training (site) [required]
Training (site) [required]
Other contributors (e.g., Plant Engineering, Plant Maintenance, Licensing and Regulatory Affairs, Outage Management, Outage Services, Emergency Procedure Guidelines owner, vendor representative) shall participate as required.
Other contributors (e.g., Plant Engineering, Plant Maintenance, Licensing and Regulatory Affairs, Outage Management, Outage Services, Emergency Procedure Guidelines owner, vendor representative) shall participate as required.
NF-AA-100 Revision 13 Page 42 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT)
 
Meetings And Team Responsibilities Page 2 of 3 The conduct of RDOT meetings is as follows:
NF-AA-100 Revision 13 Page 42 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities Page 2 of 3 The conduct of RDOT meetings is as follows:
1.ORGANIZE RDOT meetings required by this procedure. (Core Designer)2.PREPARE the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers) 3.SCHEDULE RDOT meetings required by this procedure and ENSURE adequate representation at each RDOT meeting commensurate with the items to be discussed.(RDOT Chair) a.IDENTIFY required attendees in meeting invitation.
: 1. ORGANIZE RDOT meetings required by this procedure. (Core Designer)
b.IDENTIFY optional attendees in meeting invitation.
: 2. PREPARE the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers)
c.ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
: 3. SCHEDULE RDOT meetings required by this procedure and ENSURE adequate representation at each RDOT meeting commensurate with the items to be discussed.
4.CONVENE RDOT meetings and PRESENT prepared information.(Core Designer and other responsible NF engineers) a.If meeting is cancelled prior to scheduled meeting time then RESCHEDULE (Core Designer)b.If meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (RDOT Chairman)
(RDOT Chair)
-If meeting is cancelled then RESCHEDULE.(Core Designer)-If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s) and SOLICIT comments (Core Designer).
: a. IDENTIFY required attendees in meeting invitation.
-WRITE an Issue Report documenting inadequate attendance (RDOT members not in attendance, RDOT Chairman or designee).5.DOCUMENT required RDOT concurrence in minutes to the RDOT meetings. (Core Designer)a. If concurrence is not obtained, then ESCALATE differences of opinion to Senior Managers from Nuclear Fuels and other areas, as applicable.(Core Designer)6.DOCUMENT action items in minutes to the RDOT meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)7.DOCUMENT meeting attendance in minutes to the RDOT meetings.(Core Designer)
: b. IDENTIFY optional attendees in meeting invitation.
NF-AA-100 Revision 13 Page 43 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT)
: c. ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
Meetings And Team Responsibilities Page 3 of 3 The RDOT shall:
: 4. CONVENE RDOT meetings and PRESENT prepared information. (Core Designer and other responsible NF engineers)
: 1. and AGREE upon the Cycle Design Inputs and Requirements (as defined by this and other guidelines/procedures);
: a. If meeting is cancelled prior to scheduled meeting time then RESCHEDULE (Core Designer)
: 2. that all design inputs are approved by both NF and unit representatives before they are used in definitive analyses; 3.
: b. If meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (RDOT Chairman)
4.
                - If meeting is cancelled then RESCHEDULE. (Core Designer)
design implementation;
                - If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s) and SOLICIT comments (Core Designer).
: 5. on the reload design and licensing schedule and responsibilities; 6.
                - WRITE an Issue Report documenting inadequate attendance (RDOT members not in attendance, RDOT Chairman or designee).
schedule, and DETERMINE corrective actions;
: 5. DOCUMENT required RDOT concurrence in minutes to the RDOT meetings. (Core Designer)
: 7. the applicability of the reload analyses prior to criticality;
: a. If concurrence is not obtained, then ESCALATE differences of opinion to Senior Managers from Nuclear Fuels and other areas, as applicable. (Core Designer)
: 8. as necessary to fulfill the above goals.
: 6. DOCUMENT action items in minutes to the RDOT meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
: 7. DOCUMENT meeting attendance in minutes to the RDOT meetings. (Core Designer)
 
NF-AA-100 Revision 13 Page 43 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities Page 3 of 3 The RDOT shall:
EVALUATE  1.                   and AGREE upon the Cycle Design Inputs and Requirements (as defined by this and other guidelines/procedures);
ASSURE    2.               that all design inputs are approved by both NF and unit representatives before they are used in definitive analyses; SUPPORT NF 3. technical reviews of proposed fuel and core designs; ASSURE the4.completeness of the work scope scheduled to support reload design implementation; AGREE      5.             on the reload design and licensing schedule and responsibilities; EVALUATE 6.changes to, or problems with, the reload design and/or licensing schedule, and DETERMINE corrective actions; CONFIRM    7.                 the applicability of the reload analyses prior to criticality; MEET      8.           as necessary to fulfill the above goals.
To support the above requirements, each RDOT representative shall:
To support the above requirements, each RDOT representative shall:
1.PARTICIPATE as required in RDOT meetings or activities;
: 1. PARTICIPATE as required in RDOT meetings or activities; ENSURE    2.               that the Cycle Design Inputs and Requirements reflect all relevant plant and procedure changes in their functional area; REVIEW    3.               historic and current, unit and industry, operational experience and assure that applicable experience is considered; PROVIDE    4. or CONCUR with inputs for design and licensing analyses; REVIEW    5.               proposed fuel and reload designs for potential impacts related to their functional area; 6.
: 2. that the Cycle Design Inputs and Requirements reflect all relevant plant and procedure changes in their functional area;
IDENTIFY changes    to programs, procedures, UFSAR or training in their functional area necessitated by the reload design; 7.
: 3. historic and current, unit and industry, operational experience and assure that applicable experience is considered; 4.
IDENTIFY performance      monitoring needs relating to fuel, core, or operating strategies for which the station has little experience.
with inputs for design and licensing analyses;
 
: 5. proposed fuel and reload designs for potential impacts related to their functional area; 6.
functional area necessitated by the reload design; 7.
strategies for which the station has little experience.
NF-AA-100 Revision 13 Page 44 of 54 ATTACHMENT 2 Cycle Design Inputs and Requirements Page 1 of I The CDIR should assure clear and consistent understanding between the company, fuel vendor and other organizations performing reload work. The items to be documented may include:
NF-AA-100 Revision 13 Page 44 of 54 ATTACHMENT 2 Cycle Design Inputs and Requirements Page 1 of I The CDIR should assure clear and consistent understanding between the company, fuel vendor and other organizations performing reload work. The items to be documented may include:
1.Cycle design inputs such as: the EUP; the reload fuel design; the cycle operating regime; estimated previous cycle EOC; (CM-9) 2.Other cycle design objectives such as component lifetimes; fuel utilization targets; reload batch size; fuel shuffle constraints; specific use of particular fuel or components (reinserts; LTAs);3.Constraints or objectives for the BOC and EOC refueling outages; 4.Vendor fuel design limits such as maximum enrichment, burnup capability; incore residence; maneuvering rates; chemistry; LHGR; 5.Safety analysis inputs such as RCS characteristics, reactivity parameters, boron concentrations; etc.;
: 1. Cycle design inputs such as: the EUP; the reload fuel design; the cycle operating regime; estimated previous cycle EOC; (CM-9)
6.Limiting values for safety parameters, e.g., moderator temperature coefficients; peaking factors; shutdown margin; source term; 7.Bounding operating envelopes and/or conditions; 8.Other plant characteristics (and particularly changes) which affect the assumed bases for the reload design; 9.Requirements for particular analyses or methodologies; 10.Targets for ease
: 2. Other cycle design objectives such as component lifetimes; fuel utilization targets; reload batch size; fuel shuffle constraints; specific use of particular fuel or components (reinserts; LTAs);
/flexibility of operation; margin to setpoints; etc.;
: 3. Constraints or objectives for the BOC and EOC refueling outages;
11.Relevant inputs from: dialogue with the fuel vendor or contractors; Operating Experience; unit core monitoring programs; 12.An initial schedule of reload work, deliverables and responsibilities NF-AA-100 Revision 13 Page 45 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 1 of 3 The following list of items should be considered when developing specific Reload Design Review Criteria for lower level procedures or guidelines.
: 4. Vendor fuel design limits such as maximum enrichment, burnup capability; incore residence; maneuvering rates; chemistry; LHGR;
: 5. Safety analysis inputs such as RCS characteristics, reactivity parameters, boron concentrations; etc.;
: 6. Limiting values for safety parameters, e.g., moderator temperature coefficients; peaking factors; shutdown margin; source term;
: 7. Bounding operating envelopes and/or conditions;
: 8. Other plant characteristics (and particularly changes) which affect the assumed bases for the reload design;
: 9. Requirements for particular analyses or methodologies;
: 10. Targets for ease/flexibility of operation; margin to setpoints; etc.;
: 11. Relevant inputs from: dialogue with the fuel vendor or contractors; Operating Experience; unit core monitoring programs;
: 12. An initial schedule of reload work, deliverables and responsibilities
 
NF-AA-100 Revision 13 Page 45 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 1 of 3 The following list of items should be considered when developing specific Reload Design Review Criteria for lower level procedures or guidelines.
Evaluations of new core designs and operating strategies should include (but are not limited to) the considerations tabulated below. The effects of each on: limiting safety analysis envelopes, actual steady-state plant operation and `routine' maneuvers and transients should all be considered.
Evaluations of new core designs and operating strategies should include (but are not limited to) the considerations tabulated below. The effects of each on: limiting safety analysis envelopes, actual steady-state plant operation and `routine' maneuvers and transients should all be considered.
1Operational Strategy. Will the operational strategy be different to previous cycles? Are changes to o eratinrocedures required?
1  Operational Strategy. Will the operational strategy be different to previous cycles? Are changes to o eratin      rocedures required?
2 Operational considerations:
2   Operational considerations:
-Can the core be operated at the intended power level for the full cycle with sufficient flexibility and maneuverability?
      - Can the core be operated at the intended power level for the full cycle with sufficient flexibility and maneuverability?
-Is there any change in margin to heat rate and other safety limits? Will this significantly affect required operator actions in steady-state or transient operation?
      - Is there any change in margin to heat rate and other safety limits? Will this significantly affect required operator actions in steady-state or transient operation?
-Will core response or controllability be affected (changes in rod worth, reactor period, temperature coefficients of reactivity, shutdown margin, potential for xenon oscillations, power oscillations or instability, shutdown boron requirements, etc.)?
      - Will core response or controllability be affected (changes in rod worth, reactor period, temperature coefficients of reactivity, shutdown margin, potential for xenon oscillations, power oscillations or instability, shutdown boron requirements, etc.)?
-Equipment Out-of-Service (EOOS). Is the scope of the licensed EOOS analyses sufficient to address current plant operational issues? Are additional EOOS analyses required/warranted?
      - Equipment Out-of-Service (EOOS). Is the scope of the licensed EOOS analyses sufficient to address current plant operational issues? Are additional EOOS analyses required/warranted?
3 Reactivity Management.
3   Reactivity Management. Are the reactivity control mechanisms during fuel handling or for the loaded core different from previous cycles? Are there any changes to reactivity envelopes or control assumptions?
Are the reactivity control mechanisms during fuel handling or for the loaded core different from previous cycles? Are there any changes to reactivity envelopes or control assumptions?
4   Core Monitoring. Could core monitoring and instrument indications be affected by significant changes in loading pattern, power distribution, core response, core pressure drop, shielding, etc.?
4 Core Monitoring. Could core monitoring and instrument indications be affected by significant changes in loading pattern, power distribution, core response, core pressure drop, shielding, etc.?
5   Changes in Key Parameters. Are significant changes expected at any time in cycle for:
5 Changes in Key Parameters. Are significant changes expected at any time in cycle for:
power distribution; reactivity coefficients; reactivity worths; reactor response? Will this affect the values assumed in safety analyses? How will this be a pparent to the operators?
power distribution; reactivity coefficients; reactivity worths; reactor response? Will this affect the values assumed in safet y analyses? How will this be a pp arent to the operators?
6   Does the cycle design si nificantl affect instrument response or calibration, etc.?
6 Does the cycle design si nificantlaffect instrument response or calibration, etc.?
7   Does the cycle design affect other items of plant, e.g., flows or temperatures to Steam Generators, reactor vessel fluence, etc.? Have these been discussed with cognizant 1 en ineers? CM-4)
7 Does the cycle design affect other items of plant, e.g., flows or temperatures to Steam Generators, reactor vessel fluence, etc.? Have these been discussed with cognizant 1 en ineers?
 
CM-4)
NF-AA-1 00 Revision 13 Page 46 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 2 of 3 Chemistry. Will coolant chemistry be affected by core design changes or will the chemistry regime be changed? Will planned chemistry changes affect core or fuel performance (e.g., zinc, hydrogen, noble metals, pH control, etc.)? Will known abnormal chemistry conditions affect core or fuel performance (e.g., condenser tube leaks, chemical intrusions, etc.)? Are empirical failed fuel indications expected to differ from p revious cycles? CM-13 4
NF-AA-1 00 Revision 13 Page 46 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 2 of 3 Chemistry. Will coolant chemistry be affected by core design changes or will the chemistry regime be changed? Will planned chemistry changes affect core or fuel performance (e.g., zinc, hydrogen, noble metals, pH control, etc.)? Will known abnormal chemistry conditions affect core or fuel performance (e.g., condenser tube leaks, chemical intrusions, etc.)? Are empirical failed fuel indications expected to differ from p revious c y cles?CM-13 4 9 ! Modeling. Does the core design involve novel characteristics that may affect modeling accuracy vs. Previous cycles? How will this be accounted for in the design and testing acceptance criteria? Should procedural guidance or training be provided to address any limitations associated with the predictive capability of core performance prediction tools?
9 ! Modeling. Does the core design involve novel characteristics that may affect modeling accuracy vs. Previous cycles? How will this be accounted for in the design and testing acceptance criteria? Should procedural guidance or training be provided to address any limitations associated with the predictive capability of core performance prediction tools?
Start-Up. Are any special considerations or testing necessary for initial start-up, due to use of markedly different fuel or core desians?
Is an independent, backup code required or recommended? (CM-7) (CM-15 0  Methodologies. Are reload methodologies changing? Are new methodologies approved by NRC? What is the comparative experience between new and old methodologies?
12 Fuel, Fuel Related-and Core- Component Loading. Are fuel enrichments, loading pattern, poison loadings, burnable absorber design/loading, core component design/composition, etc. all within previous experience or proven envelopes?
Are increased uncertainties or extra conservatism required? (CM-1)
13 Coolant Conditions. Will coolant flows and/or temperatures be affected by changes in fuel desi n, physical modifications to the RCS or changes in operating strategy 14 Vendor Precautions. Are any changes to the operating regime necessary to comply with vendor-provided cautions or limitations for the operation of fuel, fuel related- and core-components?
NOTE: A change in methodology may require a 50.59 Review (Refer to LS-AA-1 04 Start-Up. Are any special considerations or testing necessary for initial start-up, due to use of markedly different fuel or core desians?
12   Fuel, Fuel Related- and Core- Component Loading. Are fuel enrichments, loading pattern, poison loadings, burnable absorber design/loading, core component design/composition, etc. all within previous experience or proven envelopes?
13   Coolant Conditions. Will coolant flows and/or temperatures be affected by changes in fuel desi n, physical modifications to the RCS or changes in operating strategy 14   Vendor Precautions. Are any changes to the operating regime necessary to comply with vendor-provided cautions or limitations for the operation of fuel, fuel related- and core-components?
Reload. Does the reload batch size, fuel design, outage offload or reload sequence, etc.
Reload. Does the reload batch size, fuel design, outage offload or reload sequence, etc.
place any new requirements on fuel storage, handling or monitoring, both in- and ex-core? (examples: heat load, criticality, shielding, storage capacity, etc) (CM 5) (CM-6 Cumulative Changes. Each new core should be evaluated to ensure that cumulative small changes over multiple cycles do not constitute a significant unreviewed change.
place any new requirements on fuel storage, handling or monitoring, both in- and ex-core? (examples: heat load, criticality, shielding, storage capacity, etc) (CM 5) (CM-6 Cumulative Changes. Each new core should be evaluated to ensure that cumulative small changes over multiple cycles do not constitute a significant unreviewed change.
Examples: mix of fuel designs; erosion of operating margin; increase in poison concentrations; accuracy of core monitoringg.__
Examples: mix of fuel designs; erosion of operating margin; increase in poison concentrations; accuracy of core monitoringg.__
0 Methodologies. Are reload methodologies changing? Are new methodologies approved by NRC? What is the comparative experience between new and old methodologies?
 
Are increased uncertainties or extra conservatism required? (CM-1)
NF-AA-1 00 Revision 13 Page 47 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 3 of 3 Core monitoring. Does the new cycle introduce any features which are worthy of 17 increased monitoring? Are changes to core monitoring procedures required? Are any new tests necessary?
NOTE: A change in methodology may require a 50.59 Review (Refer to LS-AA-1 04 Is an independent, backup code required or recommended?(CM-7) (CM-15 NF-AA-1 00 Revision 13 Page 47 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 3 of 3 Core monitoring. Does the new cycle introduce any features which are worthy of 17 increased monitoring? Are changes to core monitoring procedures required? Are any new tests necessary?
Design Complexity. Does either the fuel cycle design or individual assembly designs 18 (loading variations, etc.) introduce unnecessary complexity to the manufacturing and/or reload process? Are increased vendor and/or reload QA measures warranted? (CM-14)-
18 Design Complexity. Does either the fuel cycle design or individual assembly designs (loading variations, etc.) introduce unnecessary complexity to the manufacturing and/or reload process? Are increased vendor and/or reload QA measures warranted?(CM-14)-NOTE: Acceptance criteria for simulator fidelity may be more limiting if the simulator is used to qualify operators on reactivity manipulations.
NOTE: Acceptance criteria for simulator fidelity may be more limiting if the simulator is used to qualify operators on reactivity manipulations.
Training. Is operator training and/or a replica simulator update required? (CM-10) Doesthe reload increase PCI (pellet cladding interaction) vulnerability that requires additional 19 operator awareness training?
Training. Is operator training and/or a replica simulator update required? (CM-10) Does the reload increase PCI (pellet cladding interaction) vulnerability that requires additional 19   operator awareness training?
Is the training simulator impacted as a result of fuel design changes or changes in the fuel design codes? Changing fuel design codes may result in the inability to perform a cycle specific simulator core update because the simulator uses output from the design codes.Plant Conditions. Have any changes in plant conditions invalidated the assumptions made in the fuel c cle deli nand reload licensin ?
Is the training simulator impacted as a result of fuel design changes or changes in the fuel design codes? Changing fuel design codes may result in the inability to perform a cycle specific simulator core update because the simulator uses output from the design codes.
Plant Conditions. Have any changes in plant conditions invalidated the assumptions 1 20  made in the fuel c cle deli nand reload licensin ?
Procedures. Do changes associated with the new core design require changes to plant or fuel suDoort organization procedures?
Procedures. Do changes associated with the new core design require changes to plant or fuel suDoort organization procedures?
LTAs.Does the cycle contain an LTA program? Have all necessary plant and licensing 22' issues regardin t ghe LTAs been identified?
LTAs. Does the cycle contain an LTA program? Have all necessary plant and licensing 22' issues regardinghe  t LTAs been identified?
Cycle Flexibility
Cycle Flexibility/Contingency. Does the cycle support sufficient contingency for a change in operating strategy, increased cycle energy or a BOG redesign due to either: (indicated or potential) fuel failures; handling damage; or previous cycle under /overrun? Does the cycle support sensible options for future c Iles?
/Contingency. Does the cycle support sufficient contingency for a change in operating strategy, increased cycle energy or a BOG redesign due to either: (indicated or potential) fuel failures; handling damage; or previous cycle under
/overrun? Does the cycle support sensible options for future c Iles?
Cycle design objectives. Does the design meet the defined cycle design objectives and requirements. Is the balance between economics and operational flexibility appropriate?
Cycle design objectives. Does the design meet the defined cycle design objectives and requirements. Is the balance between economics and operational flexibility appropriate?
Regulatory compliance. Are any changes to TS, UFSAR or other configuration or 2 5 licensing documentation required? Are changes to Design basis documents or databases needed?
Regulatory compliance. Are any changes to TS, UFSAR or other configuration or 25  licensing documentation required? Are changes to Design basis documents or databases needed?
Are there any mixed core issues to address?(CM-8)Source Term.Is the source term used for radiological analyses still appropriate? Is an 27 update to the radiolo g ical anal sis re aired?
Are there any mixed core issues to address? (CM-8)
1 20 NF-AA-100 Revision 13 Page 48 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page I of 4 RELOAD REVIEW BOARD MEETINGS The Reload Review Board (RRB) is an internal Nuclear Fuels (NF) board that provides oversight and review of all Reload Designs. The RRB oversees the reload strategy, fuel cycle economics and business issues, impacts of design decisions, and communication of reload plans outside of NF. The RRB meets as required by this procedure to assure that each reload receives the appropriate level of oversight.
Source Term. Is the source term used for radiological analyses still appropriate? Is an 27 update to the radiolo g ical anal sis re aired?
 
NF-AA-100 Revision 13 Page 48 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page I of 4 RELOAD REVIEW BOARD MEETINGS The Reload Review Board (RRB) is an internal Nuclear Fuels (NF) board that provides oversight and review of all Reload Designs. The RRB oversees the reload strategy, fuel cycle economics and business issues, impacts of design decisions, and communication of reload plans outside of NF. The RRB meets as required by this procedure to assure that each reload receives the appropriate level of oversight.
The RRB is made up of the following members, or their nominated delegates:
The RRB is made up of the following members, or their nominated delegates:
Vice President, NF (Chairperson)
Vice President, NF (Chairperson)
-NF Director (Vice-Chair)
      -     NF Director (Vice-Chair)
-Manager, BWR/PWR Design
      -     Manager, BWR/PWR Design
-Manager, Engineering Safety Analysis
      -     Manager, Engineering Safety Analysis
-Manager, Fuel Reliability and Projects
      -     Manager, Fuel Reliability and Projects
-Fuel Supply Director
      -     Fuel Supply Director
-Spent Fuel Director Fuel Vendor Representative The conduct of RRB meetings is as follows:
      -     Spent Fuel Director Fuel Vendor Representative The conduct of RRB meetings is as follows:
NOTIFY RRB members and ORGANIZE RRB meetings required by this procedure.(Core Designer)2.PREPARE and PRESENT the information required by applicable steps in this procedure.(Core Designer and other responsible NF engineers) a.ENSURE that commercially sensitive data is not distributed to individuals outside of Exelon.
NOTIFY RRB members and ORGANIZE RRB meetings required by this procedure. (Core Designer)
3.ENSURE adequate representation at each RRB meeting commensurate with the items to be discussed.(NF VP or designee) 4.DOCUMENT required RRB concurrence in minutes to the RRB meetings. (Core Designer)5.DOCUMENT action items in minutes to the RRB meetings and VERIFY that action items are added to the appropriate task tracking system.(Core Designer) 6.DOCUMENT meeting attendance in minutes to the RRB meetings.(Core Designer)
: 2. PREPARE and PRESENT the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers)
NF-AA-100 Revision 13 Page 49 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 2 of 4 Senior Management Design Initialization (SMDI) Meetings A meeting held early in the Reload Design process to ensure that senior site management actively participate in the decision making process for significant changes in Reload Design process, core operating strategy, fuel mechanical design, and fuel cost associated with the reload. The purpose of the meeting is to inform and gain concurrence from site senior management on the reload plans (including cycle operating capacity factor and operating margins) and design options before the reload design is fully initialized. A summary of the results of the most recent fuel vendor audits/assessments, as well as near term future audit/assessment goals, shall also be provided to the SMDI.For major changes with broad impacts (e.g., the proposed use of a new fuel design), a Special SMDI meeting should be conducted to obtain written concurrence from site senior management at a time early enough to accommodate changes in strategy.
: a. ENSURE that commercially sensitive data is not distributed to individuals outside of Exelon.
: 3. ENSURE adequate representation at each RRB meeting commensurate with the items to be discussed. (NF VP or designee)
: 4. DOCUMENT required RRB concurrence in minutes to the RRB meetings. (Core Designer)
: 5. DOCUMENT action items in minutes to the RRB meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
: 6. DOCUMENT meeting attendance in minutes to the RRB meetings. (Core Designer)
 
NF-AA-100 Revision 13 Page 49 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 2 of 4 Senior Management Design Initialization (SMDI) Meetings A meeting held early in the Reload Design process to ensure that senior site management actively participate in the decision making process for significant changes in Reload Design process, core operating strategy, fuel mechanical design, and fuel cost associated with the reload. The purpose of the meeting is to inform and gain concurrence from site senior management on the reload plans (including cycle operating capacity factor and operating margins) and design options before the reload design is fully initialized. A summary of the results of the most recent fuel vendor audits/assessments, as well as near term future audit/assessment goals, shall also be provided to the SMDI.
For major changes with broad impacts (e.g., the proposed use of a new fuel design), a Special SMDI meeting should be conducted to obtain written concurrence from site senior management at a time early enough to accommodate changes in strategy.
The following personnel, or their nominated delegates, are recommended for attendance at SMDI meetings. Attendees required for a quorum are identified below. Identification of additional required attendees shall be made on a case-by-case basis, depending on the functional areas that may be affected by significant changes proposed for a specific reload.
The following personnel, or their nominated delegates, are recommended for attendance at SMDI meetings. Attendees required for a quorum are identified below. Identification of additional required attendees shall be made on a case-by-case basis, depending on the functional areas that may be affected by significant changes proposed for a specific reload.
Optional attendees, while not required for a quorum, should be encouraged to attend.
Optional attendees, while not required for a quorum, should be encouraged to attend.
Required SMDI Meeting Attendees
Required SMDI Meeting Attendees
-NF Vice President (Chairperson)
      -       NF Vice President (Chairperson)
-Site Vice President (Vice-Chair)
      -       Site Vice President (Vice-Chair)
-Station Manager
      -       Station Manager
-Engineering Director
      -       Engineering Director
-Operations Director
      -       Operations Director
-Training Director
      -       Training Director
-Reactor Engineering Manager
      -       Reactor Engineering Manager
-Site Chemistry Manager
        -     Site Chemistry Manager
-Site Radiation Protection Manager
        -     Site Radiation Protection Manager
--NF Core Designer
        --     NF Core Designer
---Manager, Engineering Safety Analysis N F-AA-1 00 Revision 13 Page 50 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 3 of 4 Optional SMDI Meeting Attendees Outage Director
        ---     Manager, Engineering Safety Analysis
-Sr Manager of Design Engineering Sr Manager of System Engineering
 
-Licensing and Regulatory Affairs Manager Other Site Managers (e.g., Engineering, Plant Maintenance, Health Physics, Outage Management)
N F-AA-1 00 Revision 13 Page 50 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 3 of 4 Optional SMDI Meeting Attendees Outage Director
-Reactor Services representative (particularly for fuel design changes or lead assemblies)
-     Sr Manager of Design Engineering Sr Manager of System Engineering
-NF Director NF representatives (Manager PWR/BWR Design, FRE, other individual contributors)
-     Licensing and Regulatory Affairs Manager Other Site Managers (e.g., Engineering, Plant Maintenance, Health Physics, Outage Management)
-     Reactor Services representative (particularly for fuel design changes or lead assemblies)
-     NF Director NF representatives (Manager PWR/BWR Design, FRE, other individual contributors)
Fuel Vendor representative
Fuel Vendor representative
-RDOT representatives
-     RDOT representatives
-Corporate Operations Fleet Outage Scheduler NF-AA-100 Revision 13 Page 51 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 4 of 4 The conduct of SMDI meetings is as follows:
-     Corporate Operations Fleet Outage Scheduler
1.ORGANIZE SMDI meetings required by this procedure.(Core Designer)2.PREPARE the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers) 3.SCHEDULE SMDI meetings required by this procedure and ENSURE adequate representation at each SMDI meeting commensurate with the items to be discussed.(Core Designer) a.INVITE all required and optional members listed above.
 
b.IDENTIFY required attendees in meeting invitation.
NF-AA-100 Revision 13 Page 51 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 4 of 4 The conduct of SMDI meetings is as follows:
c.IDENTIFY optional attendees in meeting invitation.
: 1. ORGANIZE SMDI meetings required by this procedure. (Core Designer)
d.ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
: 2. PREPARE the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers)
4.CONVENE SMDI meetings and PRESENT prepared information. (Core Designer and other responsible NF engineers) a.DISCUSS the procedural purpose and expectations of the SMDI meeting including active participation by all required organizations.(Core Designer)b.CONFIRM all required procedural requirements are satisfied (Core Designer)c.If meeting is cancelled prior to scheduled meeting time then RESCHEDULE (Core Designer)d.If meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting.(NF VP or designee)
: 3. SCHEDULE SMDI meetings required by this procedure and ENSURE adequate representation at each SMDI meeting commensurate with the items to be discussed.
-If meeting is cancelled then RESCHEDULE.(Core Designer)-If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s)and SOLICIT comments (Core Designer).
(Core Designer)
-WRITE an Issue Report documenting inadequate attendance.(SMDI members not in attendance, NF VP or designee).
: a. INVITE all required and optional members listed above.
5.DOCUMENT required Site VP concurrence in minutes to the SMDI meetings. (Core Designer)6.DOCUMENT action items in minutes to the SMDI meetings and VERIFY that action items are added to the appropriate task tracking system.(Core Designer)7.DOCUMENT meeting attendance in minutes to the SMDI meetings.(Core Designer)
: b. IDENTIFY required attendees in meeting invitation.
: c. IDENTIFY optional attendees in meeting invitation.
: d. ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
: 4. CONVENE SMDI meetings and PRESENT prepared information. (Core Designer and other responsible NF engineers)
: a. DISCUSS the procedural purpose and expectations of the SMDI meeting including active participation by all required organizations. (Core Designer)
: b. CONFIRM all required procedural requirements are satisfied (Core Designer)
: c. If meeting is cancelled prior to scheduled meeting time then RESCHEDULE (Core Designer)
: d. If meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (NF VP or designee)
                -   If meeting is cancelled then RESCHEDULE. (Core Designer)
                -   If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member( s) and SOLICIT comments (Core Designer).
                -   WRITE an Issue Report documenting inadequate attendance. (SMDI members not in attendance, NF VP or designee).
: 5. DOCUMENT required Site VP concurrence in minutes to the SMDI meetings. (Core Designer)
: 6. DOCUMENT action items in minutes to the SMDI meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
: 7. DOCUMENT meeting attendance in minutes to the SMDI meetings. (Core Designer)
 
N F-AA-100 Revision 13 Page 52 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 1 of 3 Reload Design Kickoff Meeting The Reload Design Kickoff consists of a team of NF and Fuel Vendor personnel that provides up front planning, input, and review of reload activities. It identifies responsible individuals for various reload design and licensing tasks, potential non-routine activities that may necessitate additional resources, and an overview of expected reload scope and schedule. This meeting provides an opportunity for the fuel vendor to seek clarification and understanding of the reload expectations.
N F-AA-100 Revision 13 Page 52 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 1 of 3 Reload Design Kickoff Meeting The Reload Design Kickoff consists of a team of NF and Fuel Vendor personnel that provides up front planning, input, and review of reload activities. It identifies responsible individuals for various reload design and licensing tasks, potential non-routine activities that may necessitate additional resources, and an overview of expected reload scope and schedule. This meeting provides an opportunity for the fuel vendor to seek clarification and understanding of the reload expectations.
The Reload Design Kickoff meeting should be attended by the following personnel or their nominated delegates. Additional required attendees are identified on a case-by-case basis, considering functional areas that may be affected by significant changes proposed for a specific reload. The Core Designer and BWR/PWR Design Manager are responsible for making this determination.
The Reload Design Kickoff meeting should be attended by the following personnel or their nominated delegates. Additional required attendees are identified on a case-by-case basis, considering functional areas that may be affected by significant changes proposed for a specific reload. The Core Designer and BWR/PWR Design Manager are responsible for making this determination.
-NF PWR/BWR Design Manager (Chairperson - required)
                  - NF PWR/BWR Design Manager (Chairperson - required)
-NF Unit Core Designer (required)- Engineering Safety Analysis Engineer (required)- Fuel Vendor Project Manager (required)
                  - NF Unit Core Designer (required)
-Fuel Vendor Fuel/Core representative (required)- Fuel Vendor Safety Analysis representative (required)- Fuel Reliability Engineer (optional)
                  - Engineering Safety Analysis Engineer (required)
-NF Vendor Oversight Engineer (optional)
                  - Fuel Vendor Project Manager (required)
-Manager, Engineering Safety Analysis (optional)
                  - Fuel Vendor Fuel/Core representative (required)
-NF Fuel Supply Representative (optional)
                  - Fuel Vendor Safety Analysis representative (required)
-Other Fuel Vendor Representatives (e.g., Licensing representatives, Manufacturing representatives, etc.) (as needed)
                  - Fuel Reliability Engineer (optional)
-Other contributors (e.g., Reactor Engineering) shall participate as required.
                  - NF Vendor Oversight Engineer (optional)
                  - Manager, Engineering Safety Analysis (optional)
                  -   NF Fuel Supply Representative (optional)
                  - Other Fuel Vendor Representatives (e.g., Licensing representatives, Manufacturing representatives, etc.) (as needed)
                  - Other contributors (e.g., Reactor Engineering) shall participate as required.
 
NF-AA-100 Revision 13 Page 53 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 2 of 3 The conduct of Reload Design Kickoff meetings is as follows:
NF-AA-100 Revision 13 Page 53 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 2 of 3 The conduct of Reload Design Kickoff meetings is as follows:
1.ORGANIZE and PREPARE for the meeting.(Core Designer and other responsible NF engineers, and/or Fuel Vendor representatives) 2.SCHEDULE the meeting and ENSURE adequate representation for the items to be discussed.(Core Designer and Fuel Vendor representative) a.INVITE all required and optional members listed above.
: 1. ORGANIZE and PREPARE for the meeting. (Core Designer and other responsible NF engineers, and/or Fuel Vendor representatives)
b.IDENTIFY required attendees in meeting invitation.
: 2. SCHEDULE the meeting and ENSURE adequate representation for the items to be discussed. (Core Designer and Fuel Vendor representative)
c.IDENTIFY optional attendees in meeting invitation.
: a. INVITE all required and optional members listed above.
d.ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
: b. IDENTIFY required attendees in meeting invitation.
3.CONVENE the meeting and DISCUSS required topics (Core Designer and other responsible NF engineers, and/or Fuel Vendor representatives).
: c. IDENTIFY optional attendees in meeting invitation.
Topics that should be considered include the following items:
: d. ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
-fuel product lines and optional features
: 3. CONVENE the meeting and DISCUSS required topics (Core Designer and other responsible NF engineers, and/or Fuel Vendor representatives). Topics that should be considered include the following items:
-core design approach (ILLCD, CCC, low leakage design, etc.)
                - fuel product lines and optional features
-fuel channel designs
                - core design approach (ILLCD, CCC, low leakage design, etc.)
-control blades/control assemblies
                -   fuel channel designs
-burnable absorbers
                -   control blades/control assemblies
-operating domains plant modifications industry OPEX vendor and/or Exelon reload process/procedure changes reload methodology changes shipping containers reload schedule manufacturing schedule manufacturing product/process/procedure changes lessons learned from previous reload post job briefs collaboration kickoff meeting action items significant changes that were made to the inputs, assumptions, or methods used for reload analysis and/or licensing for prior reloads during either the reload analysis and licensing phase or during cycle operation NF-AA-100 Revision 13 Page 54 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 3 of 3 a.If the meeting is cancelled prior to scheduled meeting time then RESCHEDULE.(Core Designer)b.If the meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting.(Chair)-If meeting is cancelled then RESCHEDULE.(Core Designer)-If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s) and SOLICIT comments. (Core Designer)-WRITE an Issue Report documenting inadequate attendance (Chair or designee).
                -   burnable absorbers
4.DOCUMENT action items in meeting minutes and VERIFY that action items are added to the appropriate task tracking system.(Core Designer)a.DOCUMENT meeting attendance in meeting minutes to the Reload Design Kickoff meetings. (Core Designer)b.INCLUDE assignments to track resolution of any impacts to simulator cycle specific core update ability as referenced in Step 4.3.6. (Core Designer)
                -   operating domains plant modifications industry OPEX vendor and/or Exelon reload process/procedure changes reload methodology changes shipping containers reload schedule manufacturing schedule manufacturing product/process/procedure changes lessons learned from previous reload post job briefs collaboration kickoff meeting action items significant changes that were made to the inputs, assumptions, or methods used for reload analysis and/or licensing for prior reloads during either the reload analysis and licensing phase or during cycle operation
 
NF-AA-100 Revision 13 Page 54 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 3 of 3
: a. If the meeting is cancelled prior to scheduled meeting time then RESCHEDULE.
(Core Designer)
: b. If the meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (Chair)
          - If meeting is cancelled then RESCHEDULE. (Core Designer)
          - If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s) and SOLICIT comments. (Core Designer)
          - WRITE an Issue Report documenting inadequate attendance (Chair or designee).
: 4. DOCUMENT action items in meeting minutes and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
: a. DOCUMENT meeting attendance in meeting minutes to the Reload Design Kickoff meetings. (Core Designer)
: b. INCLUDE assignments to track resolution of any impacts to simulator cycle specific core update ability as referenced in Step 4.3.6. (Core Designer)
 
ATTACHMENT 6 EGC Procedure NF-AB-1 10, "Bundle and Core Design (BWR)"
ATTACHMENT 6 EGC Procedure NF-AB-1 10, "Bundle and Core Design (BWR)"
SM Exekn Nuclear NF-AB-110 Revision 10 Page 1 of 13 Level 3 - Information Use BUNDLE AND CORE DESIGN (BWR) 1.PURPOSE 1.1.This procedure describes and outlines the processes and specific analyses required to generate reload bundle nuclear design(s) and core designs for Exelon BWR reactors. This includes specification of both the preliminary and final bundle/core designs, if needed, based upon the specific vendor/Exelon process map.
 
1.2.The bundle and core design processes are dependent upon each other. A reload bundle design must be shown to work within a specific core design in order to be acceptable for use.(CM-1)2.TERMS AND DEFINITIONS 2.1.Bundle and Core Design -The bundle and core design specifies the type and the number of fuel bundles to be ordered for Design Cycle, It also specifies the core loading pattern for use in the reload licensing analysis.
Exekn                SM NF-AB-110 Revision 10 Page 1 of 13 Nuclear                                        Level 3 - Information Use BUNDLE AND CORE DESIGN (BWR)
2.2.Cycle Design Inputs and Requirements (CDIR) -- A formal document(s) containing cycle design targets and the inputs and acceptable outputs that define the boundaries of the cycle design and licensing analyses.
: 1. PURPOSE 1.1. This procedure describes and outlines the processes and specific analyses require d
2.3.Current Cycle - The current operating cycle.
to generate reload bundle nuclear design(s) and core designs for Exelon BWR reactors. This includes specification of both the preliminary and final bundle/core designs, if needed, based upon the specific vendor/Exelon process map.
2.4.Design Cycle - The cycle in which the fuel assembly designs, core reload design and reload-licensing work is being performed.
1.2. The bundle and core design processes are dependent upon each other. A reload bundle design must be shown to work within a specific core design in order to be acceptable for use. (CM-1)
2.5.Energy Utilization Plan (EUP) -
: 2. TERMS AND DEFINITIONS 2.1. Bundle and Core Design - The bundle and core design specifies the type and the number of fuel bundles to be ordered for Design Cycle, It also specifies the core loading pattern for use in the reload licensing analysis.
The approved schedule of core thermal energy requirements (i.e., cycle length, power level, EFPD and end of cycle extension options such as coastdown) and outage dates input to the reload design.
2.2. Cycle Design Inputs and Requirements (CDIR) -- A formal document(s) containing cycle design targets and the inputs and acceptable outputs that define the boundaries of the cycle design and licensing analyses.
2.6.Fuel Change Package (FCP) - A type of EC/ECR used to document, approve and implement core reload and cycle management activities.
2.3. Current Cycle - The current operating cycle.
NF-AB-110 Revision 10 Page 2 of 13 3.RESPONSIBILITIES 3.1.Core Designer - The Nuclear Fuels individual who is qualified in the reload bundle and core design processes and assigned by the Manager BWR Design to complete the work in the reload bundle and core design schedule. The Core Designer also provides guidance to the Verifier regarding the scope of verification and satisfactorily resolves the Verifiers comments.
2.4. Design Cycle - The cycle in which the fuel assembly designs, core reload design and reload-licensing work is being performed.
4.MAIN BODY NOTE: Tasks in this Procedure have no associated Certification Guides except those called out in the referenced T&RMs.
2.5. Energy Utilization Plan (EUP) - The approved schedule of core thermal energy requirements (i.e., cycle length, power level, EFPD and end of cycle extension options such as coastdown) and outage dates input to the reload design.
NOTE: The process of determining the bundle and core design is iterative in nature. Each design step does not need to be performed in a given iteration (i.e. unacceptable results at one step negate the need to perform the remaining steps).
2.6. Fuel Change Package (FCP) - A type of EC/ECR used to document, approve and implement core reload and cycle management activities.
 
NF-AB-110 Revision 10 Page 2 of 13
: 3. RESPONSIBILITIES 3.1. Core Designer - The Nuclear Fuels individual who is qualified in the reload bundle and core design processes and assigned by the Manager BWR Design to complete the work in the reload bundle and core design schedule. The Core Designer also provides guidance to the Verifier regarding the scope of verification and satisfactorily resolves the Verifiers comments.
: 4. MAIN BODY NOTE:       Tasks in this Procedure have no associated Certification Guides except those called out in the referenced T&RMs.
NOTE:       The process of determining the bundle and core design is iterative in nature. Each design step does not need to be performed in a given iteration (i.e. unacceptable results at one step negate the need to perform the remaining steps).
The steps in this procedure may be performed in an order different from that listed below.
The steps in this procedure may be performed in an order different from that listed below.
NOTE: All of the tasks in this T&RM are the responsibility of the Core Designer.
NOTE:     All of the tasks in this T&RM are the responsibility of the Core Designer.
NOTE: The Technical Task Pre-job briefing may be performed by the Core Designer as a reverse pre job brief, per HU-AA-1212, at the discretion of the Manager, BWR Design.
NOTE:     The Technical Task Pre-job briefing may be performed by the Core Designer as a reverse pre job brief, per HU-AA-1212, at the discretion of the Manager, BWR Design.
4.1.PERFORM a Technical Task Pre job briefing and risk assessment for the bundle and core design activities per HU-AA-1212.
4.1. PERFORM a Technical Task Pre job briefing and risk assessment for the bundle and core design activities per HU-AA-1212.
4.2.ENSURE that the Design Cycle EUP and CDIR are available for use, as developed per NF-AA-105-1000 and NF-AB-105, respectively, and that the CDIR contains the Design Cycle hot and cold critical eigenvalue targets, as developed per NF-AB-1 10-3020 and NF-AB-1 10-3025, respectively.
4.2. ENSURE that the Design Cycle EUP and CDIR are available for use, as developed per NF-AA-105-1000 and NF-AB-105, respectively, and that the CDIR contains the Design Cycle hot and cold critical eigenvalue targets, as developed per NF-AB-1 10-3020 and NF-AB-1 10-3025, respectively.
NOTE: NF-AA-101 allows the bundle specification to be issued prior to full completion of the new fuel product line evaluation.
NOTE:     NF-AA-101 allows the bundle specification to be issued prior to full completion of the new fuel product line evaluation.
4.3.If a new fuel product line is being introduced for the Design Cycle, then INTERACT with the Fuel Reliability Engineer who is working with NF-AA-101 either prior to or in parallel with this procedure.(CM-1)
4.3. If a new fuel product line is being introduced for the Design Cycle, then INTERACT with the Fuel Reliability Engineer who is working with NF-AA-101 either prior to or in parallel with this procedure. (CM-1)
NF-AB-1 10 Revision 10 Page 3 of 13 4.4.DETERMINE the bundle designs and core design for the Design Cycle.
 
4.4.1.OBTAIN the cell friction model database inputs or parameters (from the fuel vendor if necessary) to support model validation for the Design Cycle core design. The inputs include, but are not limited to, inch-days, exposure, channel types (thick/thin etc.), channel material (Zr-2, Zr-4, NSF, etc.).
NF-AB-1 10 Revision 10 Page 3 of 13 4.4. DETERMINE the bundle designs and core design for the Design Cycle.
4.4.2.REVIEW the bundle and core design constraints and design goals described in the CDIR, and CONSIDER these criteria when determining appropriate candidate reload bundle nuclear design(s) for the Design Cycle.
4.4.1. OBTAIN the cell friction model database inputs or parameters (from the fuel vendor if necessary) to support model validation for the Design Cycle core design. The inputs include, but are not limited to, inch-days, exposure, channel types (thick/thin etc.), channel material (Zr-2, Zr-4, NSF, etc.).
NOTE: Consider the use of previous bundle designs for use in the Design Cycle.
4.4.2. REVIEW the bundle and core design constraints and design goals described in the CDIR, and CONSIDER these criteria when determining appropriate candidate reload bundle nuclear design(s) for the Design Cycle.
4.4.3.DEVELOP a candidate reload bundle nuclear design(s) for the Design Cycle using either NF-AB-1 10-2000 or NF-AB-1 10-3005 to generate the nuclear physics cross section data for the reload bundle design.
NOTE:       Consider the use of previous bundle designs for use in the Design Cycle.
4.4.4.DEVELOP a candidate core loading pattern for the Design Cycle per NF-AB-1 10-2210 using the candidate reload bundle nuclear designs from Step 4.4.3. (CM-3) 4.4.5.REVISE the bundle and/or core designs to meet the CDIR design criteria.
4.4.3. DEVELOP a candidate reload bundle nuclear design(s) for the Design Cycle using either NF-AB-1 10-2000 or NF-AB-1 10-3005 to generate the nuclear physics cross section data for the reload bundle design.
4.4.6.ENSURE that the core design is within the cell friction database. If the core design does not fall within the bounds of the database, then DETERMINE additional actions to be taken (e.g., monitoring, re-channeling, additional design margin).
4.4.4. DEVELOP a candidate core loading pattern for the Design Cycle per NF-AB-1 10-2210 using the candidate reload bundle nuclear designs from Step 4.4.3. (CM-3) 4.4.5. REVISE the bundle and/or core designs to meet the CDIR design criteria.
NOTE: The Core Maneuvering Characteristics determination is normally performed based on the preliminary fuel cycle design (PFCD) as described in NF-AA-100.
4.4.6. ENSURE that the core design is within the cell friction database. If the core design does not fall within the bounds of the database, then DETERMINE additional actions to be taken (e.g., monitoring, re-channeling, additional design margin).
4.4.7.EVALUATE the Design Cycle Core Maneuvering Characteristics determination per NF-AB-110-4010.
NOTE:       The Core Maneuvering Characteristics determination is normally performed based on the preliminary fuel cycle design (PFCD) as described in NF-AA-100.
4.4.8.PERFORM a preliminary fuel storage reactivity evaluation and fuel shipping reactivity evaluation for the reload bundle nuclear design(s) using NF-AB-1 10-2070 as guidance and based on the acceptance criteria listed in the CDIR. Transmittal of the preliminary results is not required.(CM-2)If the fuel storage reactivity criteria are not met, either RETURN to Step DEVELOP a new reload bundle design(s), or CONTACT the fuel vendor to obtain a revised criticality evaluation for the fuel bundle design(s) being considered.
4.4.7. EVALUATE the Design Cycle Core Maneuvering Characteristics determination per NF-AB-110-4010.
4.4.9.ENSURE that the final bundle and core designs are acceptable and that they produce satisfactory results and comply with all design constraints and design goals listed in the CDIR.(CM-7)
4.4.8. PERFORM a preliminary fuel storage reactivity evaluation and fuel shipping reactivity evaluation for the reload bundle nuclear design(s) using NF-AB-1 10-2070 as guidance and based on the acceptance criteria listed in the CDIR. Transmittal of the preliminary results is not required. (CM-2)
NF-AB-1 10Revision 10 Page 4 of 13 1.CONFIRM that the bundle designs meet the manufacturing, fuel storage/fuel shipment criticality and other Exelon or vendor-specific constraints and design criteria.(CM-5)2.CONFIRM that the core design meets the hot excess reactivity, shutdown margin, margin to thermal limits, margin to fuel exposure limits, cycle energy requirements, axial power shape, fuel shuffling criteria, fuel channel distortion related requirements, control rod inventory requirements, control rod sequence exchanges and any other design criteria.(CM-4)(CM-6) 3.For Westinghouse Optima2 fuel bundles, CHECK the sub-bundle R-factors contained in the Westinghouse fuel bundle design report (s)and ENSURE that the sub-bundle R-factors are within the valid range of the Westinghouse critical power ratio (CPR) correlation as defined in the CDIR.
If the fuel storage reactivity criteria are not met, either RETURN to Step 4.4.3        and DEVELOP a new reload bundle design(s), or CONTACT the fuel vendor to obtain a revised criticality evaluation for the fuel bundle design(s) being considered.
4.If necessary, RETURN to the appropriate previous steps of this procedure and ADJUST the core loading pattern, depletion rod patterns, and/or the bundle designs to produce satisfactory results and to comply with the design constraints and goals listed in the CDIR.
4.4.9. ENSURE that the final bundle and core designs are acceptable and that they produce satisfactory results and comply with all design constraints and design goals listed in the CDIR. (CM-7)
5.CONFIRM that the final bundle and core design complies with each line item in the CDIR.
 
A.ASSIGN responsibility for confirming each line item in the CDIR to the appropriate personnel. Individual CDIR line items may be assigned to the NF Core Design engineer, Engineering Safety Analysis engineer, fuel vendor engineer, or station engineer, as appropriate.
NF-AB-1 10 Revision 10 Page 4 of 13
B.OBTAIN confirmation for each line item in the CDIR from the assigned personnel.
: 1. CONFIRM that the bundle designs meet the manufacturing, fuel storage/fuel shipment criticality and other Exelon or vendor-specific constraints and design criteria. (CM-5)
C.DOCUMENT confirmation of each line item in the CDIR as part of the owner acceptance review of the appropriate design documents.
: 2. CONFIRM that the core design meets the hot excess reactivity, shutdown margin, margin to thermal limits, margin to fuel exposure limits, cycle energy requirements, axial power shape, fuel shuffling criteria, fuel channel distortion related requirements, control rod inventory requirements, control rod sequence exchanges and any other design criteria. (CM-4)(CM-6)
5.DOCUMENTATION 5.1.TRANSMIT the final bundle designs and quantities to the fuel vendor and to Records Management with a copy to Fuel Supply via a TODI. (Reference CC-AA-310) (SRRS#3A.130) This transmittal should contain a picture of the bundles/lattices with pin-by-pin designations.
: 3. For Westinghouse Optima2 fuel bundles, CHECK the sub-bundle R-factors contained in the Westinghouse fuel bundle design report (s) and ENSURE that the sub-bundle R-factors are within the valid range of the Westinghouse critical power ratio (CPR) correlation as defined in the CDIR.
5.2.OBTAIN the Design Cycle fuel bundle design reports from the fuel vendor.
: 4. If necessary, RETURN to the appropriate previous steps of this procedure and ADJUST the core loading pattern, depletion rod patterns, and/or the bundle designs to produce satisfactory results and to comply with the design constraints and goals listed in the CDIR.
NOTE: The unique identifier of a fuel bundle design report is typically the name and number of the report as assigned by the fuel vendor.
: 5. CONFIRM that the final bundle and core design complies with each line item in the CDIR.
NF-AB-110 Page 5 of 13 5.2.1.SPECIFY the unique identifier of each of the Design Cycle fuel bundle design reports in the FCP and DOCUMENT the fuel bundle design reports in the disposition section of the appropriate "BWR Core Reload" FCP for the unit/cycle of interest.
A.     ASSIGN responsibility for confirming each line item in the CDIR to the appropriate personnel. Individual CDIR line items may be assigned to the NF Core Design engineer, Engineering Safety Analysis engineer, fuel vendor engineer, or station engineer, as appropriate.
B.     OBTAIN confirmation for each line item in the CDIR from the assigned personnel.
C.     DOCUMENT confirmation of each line item in the CDIR as part of the owner acceptance review of the appropriate design documents.
: 5. DOCUMENTATION 5.1. TRANSMIT the final bundle designs and quantities to the fuel vendor and to Records Management with a copy to Fuel Supply via a TODI. (Reference CC-AA-310) (SRRS#3A.130) This transmittal should contain a picture of the bundles/lattices with pin-by-pin designations.
5.2. OBTAIN the Design Cycle fuel bundle design reports from the fuel vendor.
NOTE:     The unique identifier of a fuel bundle design report is typically the name and number of the report as assigned by the fuel vendor.
 
NF-AB-110 Revision                                                                10 Page 5 of 13 5.2.1. SPECIFY the unique identifier of each of the Design Cycle fuel bundle design reports in the FCP and DOCUMENT the fuel bundle design reports in the disposition section of the appropriate "BWR Core Reload" FCP for the unit/cycle of interest.
Per NF-AA-100-1000, the disposition section of the FCP is entitled "Design Analyses, UFSAR, Documents, and Meetings".
Per NF-AA-100-1000, the disposition section of the FCP is entitled "Design Analyses, UFSAR, Documents, and Meetings".
5.2.2.PERFORM an acceptance review of the fuel bundle design reports obtained from the fuel vendor.
5.2.2. PERFORM an acceptance review of the fuel bundle design reports obtained from the fuel vendor.
1.PERFORM a technical task pre-job brief for the acceptance review in accordance with HU-AA-1212.
: 1.     PERFORM a technical task pre-job brief for the acceptance review in accordance with HU-AA-1212.
2.DOCUMENT the acceptance review in an action tracking item.
: 2.       DOCUMENT the acceptance review in an action tracking item.
3.REFERENCE the action tracking item number in the appropriate section of the FCP. Per NF-AA-1 00-1000, this section of the FCP is entitled"Supporting Documents Preparation and/or Review".
: 3.       REFERENCE the action tracking item number in the appropriate section of the FCP. Per NF-AA-1 00-1000, this section of the FCP is entitled "Supporting Documents Preparation and/or Review".
5.23.PERFORM a fuel storage reactivity evaluation and fuel shipping reactivity evaluation for the reload bundle nuclear design(s) using NF-AB-110-2070 and based on the acceptance criteria listed in the CDIR.(CM-2)5.2.4.DOCUMENT the Design Cycle fuel bundle design reports in the applicable FCP by performing either of the following steps:
5.23. PERFORM a fuel storage reactivity evaluation and fuel shipping reactivity evaluation for the reload bundle nuclear design(s) using NF-AB-110-2070 and based on the acceptance criteria listed in the CDIR. (CM-2) 5.2.4. DOCUMENT the Design Cycle fuel bundle design reports in the applicable FCP by performing either of the following steps:
1.ATTACH the fuel bundle design reports to the FCP, or 2.TRANSMIT the fuel bundle design reports to the station and to Records Management via a TODI (Reference CC-AA-310 and SRRS# 3A.130) or as a Record (SRRS# 3B.107).
: 1.       ATTACH the fuel bundle design reports to the FCP, or
5.3.OBTAIN a Multi-Cycle Analysis report from the fuel vendor. This analysis should utilize the final bundle designs documented in Steps 5.1 and 5.2.
: 2.       TRANSMIT the fuel bundle design reports to the station and to Records Management via a TODI (Reference CC-AA-310 and SRRS# 3A.130) or as a Record (SRRS# 3B.107).
NF-AB-110Revision 10 Page 6 of 13 5.3.1.PERFORM an acceptance review of the Multi-Cycle Analysis report obtained from the fuel vendor.
5.3. OBTAIN a Multi-Cycle Analysis report from the fuel vendor. This analysis should utilize the final bundle designs documented in Steps 5.1 and 5.2.
1.PERFORM a technical task pre-job brief for the acceptance review in accordance with HU-AA-1212.
 
2.COMPLETE Attachment 1, Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review.
NF-AB-110 Revision 10 Page 6 of 13 5.3.1. PERFORM an acceptance review of the Multi-Cycle Analysis report obtained from the fuel vendor.
3.REFERENCE the completed Attachment 1 DVG and the Multi-Cycle Analysis report in the appropriate section of the FCP.
: 1.       PERFORM a technical task pre-job brief for the acceptance review in accordance with HU-AA-1212.
5.3.2.DOCUMENT the Multi-Cycle Analysis report in the applicable FCP by performing either of the following steps:
: 2.       COMPLETE Attachment 1, Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review.
1.ATTACH the Multi-Cycle Analysis report and completed Attachment 1 DVG to the FCP, or 2.ENTER the Multi-Cycle Analysis report and completed Attachment 1 DVG in ERMS (SRRS## 3B.107).
: 3.     REFERENCE the completed Attachment 1 DVG and the Multi-Cycle Analysis report in the appropriate section of the FCP.
5.4.CONSIDER documenting in the FCP (or in other less formal ways) any additional information that describes the development of the Design Cycle bundle and core designs, if that information will provide valuable reference material in the future for individuals who were not directly involved with the design.
5.3.2. DOCUMENT the Multi-Cycle Analysis report in the applicable FCP by perform ing either of the following steps:
6.REFERENCES 6.1.Station Commitments 6.1.1Limerick and Peach Bottom 1.CM-1 T03969: SOER 96-02 Recommendation 1; "Evaluate Mixed Cores During Design" (Steps 1.2, 4.3) 2.CM-2 T03971: SOER 96-02 Recommendation 1; "Fuel Storage Reactivity" (Steps 4.4.8, 5.2.3) 3.CM-3 T03964: SOER 96-02 Recommendation 1; "3D Nodal Simulator Model" (Steps 4.4.4)4.CM-4 T03980: SOER 96-02 Recommendation 2; "Use of Additional Design Margins" (Steps 4.4.9.2) 5.CM-5 T03981: SOER 96-02 Recommendation 3; "Perform Monitoring of Core Design" (Step 4.4.9.1)
: 1.     ATTACH the Multi-Cycle Analysis report and completed Attachment 1 DVG to the FCP, or
NF-AB-110 Page 7 of 13 6.CM-6 T03970: SOER 96-02 Recommendation 1; "Design Should Account for Operating Data" (Step 4.4.9.2)
: 2.       ENTER the Multi-Cycle Analysis report and completed Attachment 1 DVG in ERMS (SRRS## 3B.107).
CM-7 T03976: SOER 96-02 Recommendation 2; "Control of Design Inputs and Calculation Methods" (Step 4.4.9) 6.2.Cross References 6.2.1.CC-AA-310, Transmittal of Design Information 6.2.2.HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Review 6.2.3.NF-AA-100, Reload Control Procedure 6.2.4.NF-AA-100-1000, Core Reload and Cycle Management Configuration Changes 6.2.5.NF-AA-101, Nuclear Fuel Assembly and Core Component Design and Fabrication Process Changes 6.2.6.NF-AA-105-1000, Energy Utilization Plan Development 6.2.7.NF-AB-105, Managing Cycle Design Inputs and Requirements 6.2.8.NF-AB-110-2000, Bundle Design - GENIE 6.2.9.NF-AB-110-2070, Fuel Storage Reactivity 6.2.10.NF-AB-110-2210, Core Loading Pattern Development 6.2.11.NF-AB-1 10-3005, CASMO-4 Lattice Physics Calculations 6.2.12.NF-AB-110-3020, Hot Operating K-effective (BWR) 6.2.13.NF-AB-110-3025, Cold Critical K-effective (BWR) 6.2.14.NF-AB-110-4010, Core Maneuvering Characteristics Determination 7.ATTACHMENTS 7.1.Attachment 1, Design Verification Guide -
5.4. CONSIDER documenting in the FCP (or in other less formal ways) any additio nal information that describes the development of the Design Cycle bundle and core designs, if that information will provide valuable reference material in the future for individuals who were not directly involved with the design.
MULTI-CYCLE ANALYSIS Acceptance Review NF-AB-110Revision 10 Page 8 of 13 ATTACHMENT 1 Design Verification Guide -
: 6.       REFERENCES 6.1. Station Commitments 6.1.1  Limerick and Peach Bottom
MULTI-CYCLE ANALYSIS Acceptance Review Page 1 of 6 1.DOCUMENT NAMEREV NO.II.LIST OF PROCEDURES AND T&RM USED The following is a list of Procedures / T&RM which were used in the performance and verification of the analysis:
: 1.       CM-1 T03969: SOER 96-02 Recommendation 1; "Evaluate Mixed Cores During Design" (Steps 1.2, 4.3)
: 2.       CM-2 T03971: SOER 96-02 Recommendation 1; "Fuel Storage Reactivity" (Steps 4.4.8, 5.2.3)
: 3.       CM-3 T03964: SOER 96-02 Recommendation 1; "3D Nodal Simulator Model" (Steps 4.4.4)
: 4.     CM-4 T03980: SOER 96-02 Recommendation 2; "Use of Additional Design Margins" (Steps 4.4.9.2)
: 5.     CM-5 T03981: SOER 96-02 Recommendation 3; "Perform Monitoring of Core Design" (Step 4.4.9.1)
 
NF-AB-110 Revision                                                            10 Page 7 of 13
: 6. CM-6 T03970: SOER 96-02 Recommendation 1; "Design Should Account for Operating Data" (Step 4.4.9.2)
CM-7 T03976: SOER 96-02 Recommendation 2; "Control of Design Inputs and Calculation Methods" (Step 4.4.9) 6.2. Cross References 6.2.1. CC-AA-310, Transmittal of Design Information 6.2.2. HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Review 6.2.3. NF-AA-100, Reload Control Procedure 6.2.4. NF-AA-100-1000, Core Reload and Cycle Management Configuration Changes 6.2.5. NF-AA-101, Nuclear Fuel Assembly and Core Component Design and Fabrication Process Changes 6.2.6. NF-AA-105-1000, Energy Utilization Plan Development 6.2.7. NF-AB-105, Managing Cycle Design Inputs and Requirements 6.2.8. NF-AB-110-2000, Bundle Design - GENIE 6.2.9. NF-AB-110-2070, Fuel Storage Reactivity 6.2.10. NF-AB-110-2210, Core Loading Pattern Development 6.2.11. NF-AB-1 10-3005, CASMO-4 Lattice Physics Calculations 6.2.12. NF-AB-110-3020, Hot Operating K-effective (BWR) 6.2.13. NF-AB-110-3025, Cold Critical K-effective (BWR) 6.2.14. NF-AB-110-4010, Core Maneuvering Characteristics Determination
: 7.     ATTACHMENTS 7.1. Attachment 1, Design Verification Guide - MULTI-CYCLE ANALYSIS Accepta nce Review
 
NF-AB-110 Revision 10 Page 8 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 1 of 6
: 1. DOCUMENT NAME                                                                REV NO.
II. LIST OF PROCEDURES AND T&RM USED The following is a list of Procedures / T&RM which were used in the performance and verificati on of the analysis:
PROCEDURE/
PROCEDURE/
T&RMREVISION TITLE NF-AA-100Reload Control Procedure NF-AB-110Bundle and Core Design (BWR)
T&RM                REVISION        TITLE NF-AA-100                                Reload Control Procedure NF-AB-110                                Bundle and Core Design (BWR)
Note: Procedures and T&RM not used should be lined out.
Note: Procedures and T&RM not used should be lined out.
Ill.PROCEDURE/T&RM DEVIATIONS The following deviations and associated justifications from the Procedures / T&RM listed in Section I I were used for this analysis (mark N/A if no deviations were used):
Ill. PROCEDURE / T&RM DEVIATIONS The following deviations and associated justifications from the Procedures / T&RM listed in Section I I were used for this analysis (mark N/A if no deviations were used):
NF-AB-110Revision 10 Page 9 of 13 ATTACHMENT I Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 2 of 6 IV.CHECKLIST Independent Reviewer#Item for Verification Indicate Y, N, or N/A and Initial SECTION A - General Checks (N/A if not provided in report)
 
Al Does the report refer to the appropriate unit and fuel cycle(s)?
NF-AB-110 Revision 10 Page 9 of 13 ATTACHMENT I Design Verification Guide - MULTI -CYCLE ANALYSIS Acceptance Review Page 2 of 6 IV.     CHECKLIST Independent
A2 Are the identifying numbers and/or names for new fuel designs correct?A3 Is the correct fuel product line used for the reload and downstream fuel cycles?
      #   Item for Verification                                                       Reviewer Indicate Y, N, or N/A and Initial SECTION A - General Checks (N/A if not provided in report)
A4 Does the EOC N-1 projection cycle energy correspond to the value specified in the CDIR?
Al   Does the report refer to the appropriate unit and fuel cycle(s)?
A5 Does the reload fuel cycle achieve the desired EOC energy specified in the CDIR?
Are the identifying numbers and/or names for new fuel designs A2 correct?
A6 Do the batch size and splits for the reload fuel cycle match the batch size and splits used in Exelon's core model?
A3   Is the correct fuel product line used for the reload and downstream fuel cycles?
A7 Are the batch sizes for downstream fuel cycles reasonable (e.g.
Does the EOC N-1 projection cycle energy correspond to the value A4 specified in the CDIR?
consider planned power uprates)?
Does the reload fuel cycle achieve the desired EOC energy specified A5 in the CDIR?
Do the batch size and splits for the reload fuel cycle match the batch A6 size and splits used in Exelon's core model?
Are the batch sizes for downstream fuel cycles reasonable (e.g.
A7 consider planned power uprates)?
SECTION B - Design Basis (N/A if not provided in report)
SECTION B - Design Basis (N/A if not provided in report)
If provided in the report, do the criteria in the design basis of the report correspond to the CDIR criteria for:
If provided in the report, do the criteria in the design basis of the report correspond to the CDIR criteria for:
131 Thermal limit margins?
131 Thermal limit margins?
B2 Hot excess reactivity?
B2   Hot excess reactivity?
B3 Cold and SLCS shutdown margin?
B3   Cold and SLCS shutdown margin?
134 Fuel exposure limits and/or NEXRAT?
134 Fuel exposure limits and/or NEXRAT?
B5 Cell Friction Metric thresholds?
B5   Cell Friction Metric thresholds?
NF-AB-1 10 Revision 10 Page 10 of 13 ATTACHMENT 1 Design Verification Guide -
 
MULTI-CYCLE ANALYSIS Acceptance Review Page 3 of 6 IV.CHECKLIST Independent Reviewer#Item for Verification Indicate Y, N, or N/A and Initial SECTION B - Design Basis (NIA if not provided in report)
NF-AB-1 10 Revision 10 Page 10 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 3 of 6 IV.     CHECKLIST Independent
B6 Rated or maximum reactor thermal power?
      #   Item for Verification                                                         Reviewer Indicate Y, N, or N/A and Initial SECTION B - Design Basis (NIA if not provided in report)
B7 Rated or maximum core flow and capability?
B6   Rated or maximum reactor thermal power?
B8 Feedwater temperatures?
B7   Rated or maximum core flow and capability?
Bg Cycle extension techniques?
B8   Feedwater temperatures?
B10 Hot and cold critical eigenvalue basis, including local cold k-critical adjustment?
Bg Cycle extension techniques?
Hot and cold critical eigenvalue basis, including local cold k-critical B10 adjustment?
1311 Estimated OLMCPRs?
1311 Estimated OLMCPRs?
B12 Thermal Mechanical limits?
B12 Thermal Mechanical limits?
Line 638: Line 1,049:
B15 Core loading pattern strategy?
B15 Core loading pattern strategy?
B16 Target control rod pattern strategy?
B16 Target control rod pattern strategy?
NF-AB-1 10 Revision 10 Page 11 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 4 of 6 IV.CHECKLIST Independent Reviewer#Item for Verification Indicate Y, N, or N/A and Initial SECTION C -Calculations and Results (N/A if not provided in report)
 
C1 Does the reload fuel cycle show conformance with all specified design goals / limits?
NF-AB-1 10 Revision 10 Page 11 of 13 ATTACHMENT 1 Design Verification Guide - MULTI -CYCLE ANALYSIS Acceptance Review Page 4 of 6 IV. CHECKLIST Independent
C2 Do the downstream fuel cycles show reasonable conformance with all specified design goals / limits?
      #   Item for Verification                                                       Reviewer Indicate Y, N, or N/A and Initial SECTION C -Calculations and Results (N/A if not provided in report)
Do the results correspond reasonably to Exelon's core model of the C3 reload fuel cycle?
Does the reload fuel cycle show conformance with all specified C1 design goals / limits?
C4 Does the core loading match Exelon's core model of the reload fuel cycle?C5 Are the core loadings for downstream fuel cycles reasonable?
Do the downstream fuel cycles show reasonable conformance with C2 all specified design goals / limits?
C6 Do the target rod patterns match Exelon's core model for the reload fuel cycle?
Do the results correspond reasonably to Exelon's core model of the C3 reload fuel cycle?
C7 Have all calculations been performed satisfactorily? (e.g. case convergence to k-effective, modeling of FOR point)
Does the core loading match Exelon's core model of the reload fuel C4 cycle?
Based on acceptability of reload fuel cycle results and general C8 reasonability of downstream fuel cycle results, are the fresh fuel bundle designs acceptable for use for their entire projected lifetime?
C5 Are the core loadings for downstream fuel cycles reasonable?
SECTION D - Other D1 If any issues were identified with the Multi-Cycle Analysis, was an IR Document IR #
Do the target rod patterns match Exelon's core model for the reload C6 fuel cycle?
written?D2 Is the Multi-Cycle Analysis used by Fuel Supply for fuel amortization provided for the appropriate cycles (refer to NF-AA-100)?
Have all calculations been performed satisfactorily? (e.g. case C7 convergence to k-effective, modeling of FOR point)
NF-AB-110 Revision 10 Page 12 of 13 ATTACHMENT 1 Design Verification Guide -- MULTI-CYCLE ANALYSIS Acceptance Review Page 5 of 6 V.DESIGN VERIFICATION COMMENT SHEET PAGEOF VERIFIER COMMENTS I FUEL VENDOR RESOLUTION IVERIFIER RESPONSE (Date and sign each comment I resolution)
Based on acceptability of reload fuel cycle results and general C8 reasonability of downstream fuel cycle results, are the fresh fuel bundle designs acceptable for use for their entire projected lifetime?
NF-AB-110 Revision 10 Page 13 of 13 ATTACHMENT 1 Design Verification Guide -
SECTION D - Other If any issues were identified with the Multi-Cycle Analysis, was an IR D1                                                                            Document IR #
MULTI-CYCLE ANALYSIS Acceptance Review Page 6 of 6 VI.SIGNATURES THE ACTIVITIES REVIEWED BY THIS DVG ARE COMPLETE, ACCURATE, AND ADEQUATE FOR APPLICATION TO: (Plant/Unit and Cycle)
written?
PREPARER: FUEL VENDOR NAMEDATE OF REPORT REVIEW: CORE DESIGNERSIGNATURE DATE APPROVAL: MANAGER SIGNATURE DATE NOTE: This DVG should be attached to the Multi-Cycle Analysis report when it is placed into Records.
D2 Is the Multi-Cycle Analysis used by Fuel Supply for fuel amortization provided for the appropriate cycles (refer to NF-AA-100)?
 
NF-AB-110 Revision 10 Page 12 of 13 ATTACHMENT 1 Design Verification Guide -- MULTI -CYCLE ANALYSIS Acceptance Review Page 5 of 6 V. DESIGN VERIFICATION COMMENT SHEET PAGE        OF VERIFIER COMMENTS       I FUEL VENDOR RESOLUTION I          VERIFIER RESPONSE (Date and sign each comment I resolution)
 
NF-AB-110 Revision 10 Page 13 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 6 of 6 VI.     SIGNATURES THE ACTIVITIES REVIEWED BY THIS DVG ARE COMPLETE, ACCUR ATE, AND ADEQUATE FOR APPLICATION TO:
(Plant / Unit and Cycle)
PREPARER:
FUEL VENDOR NAME                                            DATE OF REPORT REVIEW:
CORE DESIGNER                                  SIGNATURE                                      DATE APPROVAL:
MANAGER                                       SIGNATURE                                       DATE NOTE: This DVG should be attached to the Multi-Cycle Analysis report when it is placed into Records.
The completed DVG is sufficient documentation of the Owner's Acceptance Review.
The completed DVG is sufficient documentation of the Owner's Acceptance Review.
ATTACHMENT 7 EGC Transmittal of Design Information document NF1000236 Rev. 1 Brian Henning Prepared bySignature John Wheeler Reviewed by SignatureDate NUCLEAR FUELS TRANSMITTAL OF DESIGN INFORMATION q SAFETY RELATED x NON-SAFETY RELATED q REGULATORY RELATED Originating Organization Nuclear Fuels 0 Other (specify)
 
N/A NF ID#NF100(Revision*1 SRRS #3A130 Page 1 of 29 Station: Quad CitiesUnit: 1Cycle: 22Generic:N/A  
ATTACHMENT 7 EGC Transmittal of Design Information document NF1000236 Rev. 1
 
NUCLEAR FUELS TRANSMITTAL OF DESIGN INFORMATION q SAFETY RELATED                       Originating Organization        NF ID#        NF100(
x NON-SAFETY RELATED                   Nuclear Fuels                  Revision*    1 q REGULATORY RELATED               0 Other (specify)   N/A             SRRS #       3A130 Page 1 of 29 Station:   Quad Cities            Unit:   1        Cycle: 22          Generic:     N/A


==Subject:==
==Subject:==
Quad Cities Unit 1 Cycle 22 Cycle D esign and Input Requirements (CDIR)
Quad Cities Unit 1 Cycle 22 Cycle Design and Input Requirements (CDIR)
To: Lyfliam Sekkat (BWR Design, Westinghouse)EC/ECR#: 377652 Jill Fisher Approved bySignature Status of Information:Z Verified q Unverified Action Tracking # for Method and Schedule of Verification for Unverified DESIGN INFORMATION.
To:   Lyfliam Sekkat (BWR Design, Westinghouse)                       EC/ECR#: 377652 Brian Henning Prepared by                  Signature John Wheeler Reviewed by                  Signature                                  Date Jill Fisher Approved by                  Signature                                  Date Status of Information:         Z Verified q Unverified Action Tracking # for Method and Schedule of Verification for Unverified DESIGN                                       N/A INFORMATION.
Desc ription of Information:
Description of Information:   Revision 0: Provides design limits and targets for the Q1 C22 bundle design, core design, and cycle operation.
Revision 0: Provides design limits and targets for the Q1 C22 bundle design, core design, and cycle operation.
Revision 1 - Update information for cycle energy requirements, feedwater temperature operating domain, Boraflex degradation, and Zirio Lead Use Channels to close Unverified Assumptions of Revision 0.
Revision 1 - Update information for cycle energy requirements, feedwater temperature operating domain, Boraflex degradation, and Zirio Lead Use Channels to close Unverified Assumptions of Revision 0.
Items A.25.b, A.27,b, A.32, and A.33 were adjusted to match the Revised Exelon CDIR procedure format.
Items A.25.b, A.27,b, A.32, and A.33 were adjusted to match the Revised Exelon CDIR procedure format.
Purpose of Information:Provides the bundle and core constraints for the cycle design.Source of Information:As documented in the attachment.
Purpose of Information:       Provides the bundle and core constraints for the cycle design.
Supplemental Distribution:
Source of Information:       As documented in the attachment.
E - Mail:Hard Gooy:
Supplemental Distribution:   E - Mail:                           Hard Gooy:
Craig Schneider (Quad Cities)Cantera Records Management Scott Vrtiska (Westinghouse)
Craig Schneider (Quad Cities)       Cantera Records Management Scott Vrtiska (Westinghouse)
Date N/A Quad Cities 1 Cycle 22 CDIR TODI NF1000236 Rev. 1 Pa g e 2 of 29 SecItem Core De sign Data Related Value Referen ce s/Comments Items Criteria Al.Cycle N-1 shutdown date 519/11 Exelon Nuclear Group's Planned Outage Schedule Rev. 28, 11111 /10 A.2 Cycle N startup date 6/8/11 Exelon Nuclear Group's Planned Outage Schedule, Rev. 28, 11/11/10 A.3 Cycle N shutdown date
 
-------- -------4/1/13'Exelon Nuclear Group's Planned Outage Schedule, Rev. 28, 11/11/10 A.4 EOC N-1 nominal exposure 2049 GWd / 16,382 MWD/MT TODI NF1000203 Rev. 0 EUP Engineering input. Same band A.5 EOC N-1 minimum exposure 16,000 MWD/MT as 01021 and Q2C21 (-400 Cycle N EUP required FOR A.6.a energy and design power 1,842 GWD / 14,765 MWd/MTU;;. 2957 MWt TODI NF1000203 Rev. 0 (EUP)'
Quad Cities 1 Cycle 22 CDIR           TODI NF1000236 Rev. 1 Page 2 of 29 Section A -*
level A'6'b FOR exposure acceptance N/A N/A band, if applicable A.7 Cycle N EUP total cycle 1,921 GWD 115,391 MWd/MTU TODI NF1000203 Rev. 0 (EUP)'
Item    Core Design Data Related                                     Value                         Referen ces/Comments Items Criteria Al. Cycle N-1 shutdown date     519/11                                                 Exelon Nuclear Group's Planned Outage Schedule Rev. 28, 11111 /10 A.2   Cycle N startup date         6/8/11                                                 Exelon Nuclear Group's Planned
ever I Exelon Reload Schedule Rev. 28 issued 11/11/10 revised the Q1 R22 outage date from 4/1/13 used in the EUP to 3/11/13. Because the bundle design and core design were already completed at the time of this revision, no changes to the EUP or CDIR energy requirements were made due to the Rev. 28 Outage Schedule. All of the Cycle N energies specified here are based on a 4/1/13 shutdown date.
                                          -- ------ - ------                                  Outage Schedule, Rev. 28, 11/11/10 A.3   Cycle N shutdown date         4/1/13'                                                 Exelon Nuclear Group's Planned Outage Schedule, Rev. 28, 11/11/10 A.4   EOC N-1 nominal exposure     2049 GWd / 16,382 MWD/MT                               TODI NF1000203 Rev. 0 EUP Engineering input. Same band A.5   EOC N-1 minimum exposure     16,000 MWD/MT                                           as 01021 and Q2C21 (-400 Cycle N EUP required FOR A.6.a   energy and design power       1,842 GWD / 14,765 MWd/MTU;;. 2957 MWt                 TODI NF1000203 Rev. 0 (EUP)'
II4Z°'t A.8 I Design MFLPD limit 0,86 Quad Cities 1 Cycle 22 CDIRTODI NF1 000236 Rev. 1 Page 3 of 29 Engineering input - no reference needed.Design Limit must be met in both MICROBURN-132 and POLCA7.
level FOR exposure acceptance A'6'b   band, if applicable          N/A                                                     N/A A.7   Cycle N EUP total cycle ever                        1,921 GWD 115,391 MWd/MTU                               TODI NF1000203 Rev. 0 (EUP)'
Design must meet these limits WITH applicable power/flow dependentlimits (off-rated penalties) and WITHany applicable penalties from control blade history,:r- Jz,Ir0 Quad Cities 1 Cycle 22 COIRTODI NFl 000236 Rev. 1 Page 4 of 29 Engineering input -- no reference needed.Design Limit must be met in both MI CROBU RN
I Exelon Reload Schedule Rev. 28 issued 11/11/10 revised the Q1 R22 outage date from 4/1/13 used in the EUP to 3/11/13. Because the bundle design and core design were already completed at the time of this revision, no changes to the EUP or CDIR energy requirements were made due to the Rev. 28 Outage Schedule. All of the Cycle N energies specified here are based on a 4/1/13 shutdown date.
-82 and POLCA7.Design must meet these limits WITH applicable power/flow dependent limits (off
II4Z°'t
-rated penalties) and WITH any applicable penalties from control blade history.
 
A.9 Design MFLCPR limit 0.90 Design Limit is 0.90. Prior to incorporating CBH R
Quad Cities 1 Cycle 22 CDIR      TODI NF1 000236 Rev. 1 Page 3 of 29 Engineering input - no reference needed.
-Factors, Design Limits should be 0.89 BOC
Design Limit must be met in both MICROBURN-132 and POLCA7.
-MOC and 0.88 MOC-EOC to provide margin to CBH R-Factors.If SCORE loading is used to eliminate or mitigate CBH R-Factors in limiting locations, 0.90 can be used for the whole cycle.
A.8 I Design MFLPD limit 0,86 Design must meet these limits WITH applicable power/flow dependent limits (off-rated penalties) and WITH any applicable penalties from control blade history,
Additionally, the limit of 0.90 applies once the CBH R-Factors are Olamented.1 2, 2 1!o Quad Cities 1 Cycle 22 CDIRT 1 NF1 0 236 Rev. 1 Page 5 of 29A.10Design MAPRAT limit 0.86 Engineering input - no reference needed.Design Limit must be met in both MICROBURN-B2 and POLCA7.
:r- Jz,Ir0
1.38% margin value using best estimate target curve per NF-AB-110-Maintain1,38% Ak/k based on short window using 3080, Rev. 10 and 1% margin to A.1 1 Cold shutdown margin limit best estimate cold target curve Tech Spec limit of 0.38%.(apply this in both POLCA and MICROBURN-82) 1.38% margin is to include determination of SDM at most limiting temperature.
 
W methodology incurs no penalty for distributed vs. local critical k-effectives.
Quad Cities 1 Cycle 22 COIR      TODI NFl 000236 Rev. 1 Page 4 of 29 Engineering input -- no reference needed.
A 0.04% Ak/k penalty A.12 SD M adjustments 0.04% Ak/k from inverted boron tubes is from ACTS 0920-01. Functionally, this 0.04% Ak/k penalty must be included in the SDM as part of the cycle R value.<51- k 2'L 4%0 Quad Cities 1 Cycle 22 CDIR TOM NF1 0236 Rev. 1 Page 6 of 29 NF-AB-110-2060 Rev. 7 nominal values.Peak set to minimize control rod 80% dk BOG ? 0 density and maintain operational
Design Limit must be met in both MI CROBU RN-82 and POLCA7.
.Peak <1 70% . k flexibility; flatness set to minimize rod
Design must meet these limits WITH applicable power/flow dependent limits (off -rated penalties) and WITH any applicable penalties from control blade history.
_.motion.A.13 Hot excess reactivity limit (Peak - Minimum) < 0.50% Ak Design goals are as shown to Based on Nominal EOC N-1 maintain rod pattern flexibility and to reduce control rod inventory for Cycle Exposure shadow corrosion channel distortion mitigation.
A.9 Design MFLCPR limit 0.90 Design Limit is 0.90. Prior to incorporating CBH R- Factors, Design Limits should be 0.89 BOC - MOC and 0.88 MOC - EOC to provide margin to CBH R -Factors. If SCORE loading is used to eliminate or mitigate CBH R-Factors in limiting locations, 0.90 can be used for the whole cycle.
Hot excess is to be driven as low as possible under the limit of 1.70%, subject to complying with other design criteria.
Additionally, the limit of 0.90 applies once the CBH R-Factors are O lamented.
MICROBURN-B2POLCA MB2 initial estimate target the same GWD/MTKeffGWD/MTKell as Q2C21 Eigenvalue Design Cycle N hot operating k-1.00550 0 1 1.00 227 55 Analysis QDC-0000-N-1756 Rev. 0, A.14 effective target 0 4.00 7 0.0 1.0015 POLCA target eigenvalue from NF-13.01.001512.10.99100 BEX-10-74 Rev. 1.
2, 2 1 !o 1
17.01.006018.00.99100 Quad Cities 1 Cycle 22 CIDIRT of NF10 236 Rev. 1 P ag e 7 of 29 A. 15 Cycle N cold k-effective target MICROBURN-B2POLCA GWD/MTKeffGWD/MTKeff 01.002001.0045 2.00.99954.01.0020 18.00.999518.01.0020 MB2 initial estimate target the same as Q2C21 Eigenvalue Design Analysis QDC-0000-N-1 756 Rev. 0.
 
POLCA target elgenvalue from NF-BEX-10-74 Rev. 1.
Quad Cities 1 Cycle 22 CDIR                  T 1 NF1 0 236 Rev. 1 Page 5 of 29 Engineering input - no reference needed.
Cold k-effective target SDM goal was increased from 1.0%
A.10  Design MAPRAT limit        0.86 Design Limit must be met in both MICROBURN-B2 and POLCA7.
A.16 adjustments in addition to None to 1.38% since best estimate cold those in A.15 target was used.
1.38% margin value using best estimate target curve per NF-AB-110-Maintain 1,38% Ak/k based on short window using 3080, Rev. 10 and 1% margin to A.1 1 Cold shutdown margin limit best estimate cold target curve                 Tech Spec limit of 0.38%.
(apply this in both POLCA and MICROBURN-82) 1.38% margin is to include determination of SDM at most limiting temperature.
W methodology incurs no penalty for distributed vs. local critical k-effectives. A 0.04% Ak/k penalty A.12 SD M adjustments           0.04% Ak/k                                     from inverted boron tubes is from ACTS 0920-01. Functionally, this 0.04% Ak/k penalty must be included in the SDM as part of the cycle R value.
                                                                                                        <51- k 2'L 4%0
 
Quad Cities 1 Cycle 22 CDIR             TOM NF1 0236 Rev. 1 Page 6 of 29 NF-AB-110-2060 Rev. 7 nominal values.
Peak set to minimize control rod BOG ? 0 . 80% dk                             density and maintain operational Peak <_ 1 . 70% . k                           flexibility; flatness set to minimize rod motion.
A.13 Hot excess reactivity limit (Peak - Minimum) < 0.50% Ak Design goals are as shown to Based on Nominal EOC N-1                     maintain rod pattern flexibility and to Cycle Exposure                                reduce control rod inventory for shadow corrosion channel distortion mitigation. Hot excess is to be driven as low as possible under the limit of 1.70%, subject to complying with other design criteria.
MICROBURN-B2                  POLCA GWD/MT        Keff                            MB2 initial estimate target the same GWD/MT          Kell 0    1 .0055            04    1 1.00227 as Q2C21 Eigenvalue Design Cycle N hot operating k-A.14                              0.0   1.0015                      .00 7 55   Analysis QDC-0000-N-1756 Rev. 0, effective target                                                         POLCA target eigenvalue from NF-13.0  1.0015            12.1  0.99100 17.0  1.0060            18.0                BEX-10-74 Rev. 1.
0.99100
 
Quad Cities 1 Cycle 22 CIDIR                  T of NF10 236 Rev. 1 Page 7 of 29 MICROBURN-B2                    POLCA            MB2 initial estimate target the same GWD/MT        Keff        GWD/MT      Keff      as Q2C21 Eigenvalue Design Cycle N cold k-effective    0        1.0020          0      1.0045      Analysis QDC-0000-N-1 756 Rev. 0.
A. 15 target                      2.0      0.9995          4.0    1.0020      POLCA target elgenvalue from NF-18.0      0.9995          18.0    1.0020      BEX-10-74 Rev. 1.
Cold k-effective target                                                     SDM goal was increased from 1.0%
A.16 adjustments in addition to None                                             to 1.38% since best estimate cold those in A.15                                                               target was used.
NF-AB-110-3050, Rev. 10.
NF-AB-110-3050, Rev. 10.
1.38% margin value using best estimate target curve with 1%
1.38% margin value using best estimate target curve with 1%
uncertainty and 0.38 Tech Spec limit.
uncertainty and 0.38 Tech Spec limit.
Credit is taken for 30 alo enriched B-SLC shutdown margin limit, Maintain z 1.38% Ak SLCS Cold SDM based on short 10 (918 ppm natural boron A.17% Ak and boronEOC N-1 &#xa9; 918 ppm natural boron equivalentequivalent) per EC# 349585.
Credit is taken for 30 alo enriched B-SLC shutdown margin limit, Maintain z 1.38% Ak SLCS Cold SDM based on short 10 (918 ppm natural boron A.17 % Ak and boron            EOC N-1 &#xa9; 918 ppm natural boron equivalent      equivalent) per EC# 349585.
concentration (ppm)
concentration (ppm)
SLCS SDM will be determined using POLCA7.SLC SDM is to include determination of SLC SDM at most limiting temperature.
SLCS SDM will be determined using POLCA7.
Q uad Cities 1 Cycle 22 CDIR TODI NF1 236 Rev. 1 Page 8 of 29 WCAP 15836-P-A for Optimal exposure limits WCAP-15942-P-A Supplement 1 for Optima2 peak rod exposure limit of 62 GWd/MTU Peak Rod -
SLC SDM is to include determination of SLC SDM at most limiting temperature.
62 GWD/MTU Rod Nodal Exposure Limit based on Maximum Rod Nodal Exposure in A.18 Fuel exposure limits Peak Rod Nodal -
 
72 GWdD/MTU LHGR limit curve.
Quad Cities 1 Cycle 22 CDIR       TODI NF1           236 Rev. 1 Page 8 of 29 WCAP 15836-P-A for Optimal exposure limits WCAP-15942-P- A Supplement 1 for Optima2 peak rod exposure limit of 62 GWd/MTU Peak Rod - 62 GWD/MTU                 Rod Nodal Exposure Limit based on Maximum Rod Nodal Exposure in A.18 Fuel exposure limits Peak Rod Nodal - 72 GWdD/MTU           LHGR limit curve.
Peak Bundle - 58 7 GWD/M TU AmerGen Calculation No. C-1101-
Peak Bundle - 58 . 7 GWD/MTU          AmerGen Calculation No. C-1101-202-E620-4,43, Revision 0, "PWR &
.202-E620-4,43, Revision 0, "PWR &
BWR Isotopic Inventories for Spent Fuel Pool Gamma Heating Study."
BWR Isotopic Inventories for Spent Fuel Pool Gamma Heating Study." The calculation states a pin exposure limit of 58,700 MWdIMTU but EC 378207 documents that the limit is a bundle limit, not a pin limit as stated in the calculation.
The calculation states a pin exposure limit of 58,700 MWdIMTU but EC 378207 documents that the limit is a bundle limit, not a pin limit as stated in the calculation.
Engineering input - no reference needed A.19 Design allowance for Design to maintain 3% margin to all e exposure limits exposure limits 3%margin to all exposure limits based on nominal EOC for all cycles in MICROBURN-132 and POLCA7.WCAP 15836-P-A for O ptimal A.20 Residence time limits No limits for Optima2 fuel exposure limits; no residence time se-cified.110 Quad Cities 1 Cycle 22 CDIRTOO! NF1 236 Rev. 1 Page 9 of 29 A.21 Limits on design due to need to maintain full-core offload None Engineering input -No explicit reference needed to document there ca pability are no restrictions.
Engineering input - no reference needed Design allowance for Design to maintain A.19 exposure limits      3% margin to all exposure limits     3% margin to all exposure limits based on nominal EOC for all cycles in MICROBURN-132 and POLCA7.
NF-AB-110-2210 Rev. 12 A.22 Limits on design due to need to maintain symmetry in core power Load and operate to as near to 1/8 core symmetry as prat#icai.Maintain octant symmetry in exposure and bundle type as much as possible; i maintain symmetric partners during bundle shuffles as much as possible.
WCAP 15836-P-A for O ptimal A.20 Residence time limits No limits for Optima2 fuel           exposure limits; no residence time s e-cified.
Design Sequence Exposures Sequence Start Sequ ence End Engineering input - no reference needed.MWD/MTU Rod Group Date Date Sequence Exchange dates and 0 - 2700 A2 - 10 6/8/11 10/1/11 exposures are approximate.
110
Sequences listed can consist of Tar et se uence exchan 27 00- 5475 A2 - 910/2/111/28/12 different combinations (i.e. A2 can be A.23 g q ge plan 5475-8250 A2 - 10 1/29/12 5/26/12 groups 9 and/or 10).
 
8250 - 11000 A2 - 9 5/27/12 9/22/12 If necessary to support thermal margins or other design criteria the , 11000 - 12475 A210 9/23/12 11/24/12 Rod Groups can be changed as long t l d i i as no c o t d f n ro ro rema ns nser e or 12475 - 14800 A2 - 9 11/25/12 3/4/13 two consecutive sequence intervals.
Quad Cities 1 Cycle 22 CDIR                    TOO! NF1             236 Rev. 1 Page 9 of 29 Limits on design due to need                                                       Engineering input -No explicit A.21 to maintain full-core offload None                                                 reference needed to document there ca pability                                                                       are no restrictions.
14800 - 15425 ARC 3/5/13 4/1/13 Q uad Cities 1 Cycle 22 CDIR TODI NF1 236 Rev. 1 Page 10 of 29 NF-AB-440-1002 Rev. 2 (pending WCMS cutover)
NF-AB-110-2210 Rev. 12 Limits on design due to need A.22 to maintain symmetry in core Load and operate to as near to 1/8 core symmetry as prat#icai .                                         Maintain octant symmetry in exposure power                                                                              and bundle type as much as possible;             i maintain symmetric partners during bundle shuffles as much as possible.
Design                                         Engineering input - no reference Sequence                Sequence  Sequ ence needed.
Exposures                  Start      End MWD/MTU   Rod Group       Date       Date Sequence Exchange dates and 0 - 2700     A2 - 10     6/8/11     10/1/11   exposures are approximate.
Sequences listed can consist of 27 00 - 5475   A2 - 9      10/2/11    1/28/12   different combinations (i.e. A2 can be Tar g et se q uence exchan ge A.23                                                                                   groups 9 and/or 10).
plan                           5475   - 8250 A2 - 10     1 / 29/12   5/26/12 If necessary to support thermal 8250 - 11000     A2 - 9     5/27/12     9/22/12   margins or other design criteria , the Rod Groups can be changed as long 11000 - 12475   A2    10  9/23/12   11/24/12   as no c o n t ro l rod rema i ns i nsert e d f or 12475 - 14800   A2 - 9     11/25/12     3/4/13 two consecutive sequence intervals.
14800 - 15425       ARC       3/5/13     4/1/13
 
Q uad Cities 1 Cycle 22 CDIR                         TODI NF1           236 Rev. 1 Page 10 of 29 NF-AB-440-1002 Rev. 2 (pending WCMS cutover)
NF-AB-440-1003 Rev. 1.
NF-AB-440-1003 Rev. 1.
Specific control rod pattern Avoid sequence exchanges in June-August; avoid This avoids reliance on complex rod guidance and limitations significant rod pattern changes during sequence, if maneuvers to maintain adequate A.24a possible; avoid insertion of control rods from 48 in the thermal margin, etc; requires Periods to maintain constant a sequence middle of relatively flat hot excess reactivity; to rod pattern maintain MAPRAT within the design limits, rods may need to be insertediwithdrawn in the middle of a sequence; last sequence will allow for rods to be withdrawn to ARO Specific control rod pattern guidance and limitations A.24b+/-0.5 mk (+/-0.0005 k)
Specific control rod pattern                                                           This avoids reliance on complex rod guidance and limitations      Avoid sequence exchanges in June-August; avoid significant rod pattern changes during sequence, if     maneuvers to maintain adequate A.24a                               possible; avoid insertion of control rods from 48 in the thermal margin, etc; requires Periods to maintain constant                                                           relatively flat hot excess reactivity; to rod pattern                   middle of a sequence maintain MAPRAT within the design limits, rods may need to be insertediwithdrawn in the middle of a sequence; last sequence will allow for rods to be withdrawn to ARO Specific control rod pattern guidance and limitations A.24b                               +/-0.5 mk (+/-0.0005 k)                                       NF-AB-110-3040 Rev. 11 Maximum deviation from target k-effective N F-A B-110-3040 Rev. 11 Specific control rod pattern                                                           NF-AS-440-1002 Rev. 2 (pending guidance and limitations     Avoid use of B sequence rods; any rod that was           WCMS cutover)
NF-AB-110-3040 Rev. 11 Maximum deviation from target k-effective N F-A B-110-3040 Rev. 11 Specific control rod pattern NF-AS-440-1002 Rev. 2 (pending guidance and limitations Avoid use of B sequence rods; any rod that was WCMS cutover)
A.24c                               inserted during the previous rod sequence must be         NF-AB-440-1003 Rev. 1 Control rod locations which   fully withdrawn during the next sequence                 Fuel reliability concerns if any portion cannot be used                                                                         of a bundle is continuously controlled for much longer than one control rod sequence
A.24c inserted during the previous rod sequence must be NF-AB-440-1003 Rev. 1 Control rod locations which fully withdrawn during the next sequence Fuel reliability concerns if any portion cannot be used of a bundle is continuously controlled for much longer than one control rod sequence Quad Cities I Cycle 22 C iR TODi NF10 0 0 236 Rev. 1 Page 11 of 29 Engineering input -- no reference needed.Preferred notch positions for control rods are 00, 06 Attempt to eliminate use of shallow Specific control rod pattern 20, and 48; control rods may be placed at rod positions in design except as guidance and limitations notch positions 22-28, if necessary; intermediate required to provide MAPLHGR A.24d control rods not allowed at notch positions 02, 04, margin; goal is to maintain Rules for use of shallow rods and 30-46, but some shallow rod positions may be operational flexibility; avoid fuel required for portions of the cycle; minimize banking of conditioning limit problems rod groups at one position associated with intermediate rod movement Engineering input - no reference Specific control rod pattern needed A.24e guidance and limitations Assume no QOS rods It is assumed that any rods that are OOS rods COS at the end of the current cycle will be repaired and returned to service for the design cycle Specific control rod pattern Ensure adequate thermal margins exist for rod Engineering input - no reference A,24f guidance and limitations patterns at +/- 3 mk around the hot target eigenvalue needed Flexibility curve Maintain adequate flexibility -
 
Quad Cities 1 Cycle 22 CDIRTODI N F1 00 0236 Rev. 1 Page 12 of 29 Engineering input - no reference Specific control rod pattern needed A.24g guidance and limitations Preferred order of withdrawal to FOR (last sequence) is group 9B, 9C, then 9A last.
Quad Cities I Cycle 22 C iR                         TODi NF10 0 0 236 Rev. 1 Page 11 of 29 Engineering input -- no reference needed.
Operational preference. This order is Misc not required for compliance with core design requirements.
Preferred notch positions for control rods are 00, 06 20, and 48; control rods may be placed at               Attempt to eliminate use of shallow Specific control rod pattern                                                          rod positions in design except as guidance and limitations     intermediate notch positions 22-28, if necessary; control rods not allowed at notch positions 02, 04,     required to provide MAPLHGR A.24d                                                                                      margin; goal is to maintain Rules for use of shallow rods and 30-46, but some shallow rod positions may be required for portions of the cycle; minimize banking of operational flexibility; avoid fuel rod groups at one position                             conditioning limit problems associated with intermediate rod movement Engineering input - no reference Specific control rod pattern                                                         needed guidance and limitations A.24e                               Assume no QOS rods                                     It is assumed that any rods that are OOS rods                                                                             COS at the end of the current cycle will be repaired and returned to service for the design cycle Specific control rod pattern                                                         Engineering input - no reference guidance and limitations     Ensure adequate thermal margins exist for rod patterns at +/- 3 mk around the hot target eigenvalue   needed A,24f Flexibility                   curve Maintain adequate flexibility -
Engineering input - no reference needed A 25 a Core flow window to use in 98 - 100%Design with nominal 98% core flow
 
..design (96.04 MIbfhr) for all state-points, increasing to 100% (98 Mlb/hr) at EOR.No ICF assumed.
Quad Cities 1 Cycle 22 CDIR                          TODI N F1 000236 Rev. 1 Page 12 of 29 Engineering input - no reference Specific control rod pattern                                                           needed guidance and limitations     Preferred order of withdrawal to FOR (last sequence)
A.24g is group 9B, 9C, then 9A last.                           Operational preference. This order is Misc                                                                                   not required for compliance with core design requirements.
Engineering input - no reference needed Core flow window to use in   98 - 100%                                                 Design with nominal 98% core flow A . 25 . a                                                                                        (96.04 MIbfhr) for all state-points, design increasing to 100% (98 Mlb/hr) at EOR.
No ICF assumed.
Low Limit - 100% (100% CTP @ 98 Mlb/hr - No ICF)
Low Limit - 100% (100% CTP @ 98 Mlb/hr - No ICF)
Unit Rod-line Restrictions High Limit - 101.7% (100% CTP @ 95.5 Mlb/hr)
High Limit - 101.7% (100% CTP @ 95.5 Mlb/hr)
Engineering input - no reference A (Low and High) for operation CTF7i x 100 needed for High & Low FCL Limit 25 b..FCL=to beused in the reload sign.d design.22 .191 + .$9714 x i^t?,^+ - ^.{^J1 19S x Wr%
Unit Rod-line Restrictions                                                              Engineering input - no reference (Low and High) for operation                                                           needed for High & Low FCL Limit A . 25 . b                               FCL =                     CTF7i    x 100 to be used in the reload 22 .191 + .$9714 x i^t ,^ + - ^.{^J1 19S x Wr% ? TODI QDC-02-028.01 and QCOP sign.
TODI QDC-02-028.01 and QCOP Where CTP% = percent of rated thermal power 0202-13 Rev. 15 for FCP equation Wt&deg;/o = percent of rated core flow Quad Cities 1 Operating License DLO I SLO Values shown are expected to A.26 Cycle N-1 SLMCPR values 1.11 / 1.13 (Optima2) continue to be used in Q1C22 based on Westinghouse calculated values for previous Optima2 cycles. These are the values currently in Quad Cities Tech Specs.
design.
cI+t 7.i lio Quad Cities 1 Cycle 22 CDR TODI NF1 0 236 Rev. 1 Page 13 of 29 Engineering input - no reference needed Estimates shown are for Base Case NSS and no EGOS. Based on Current Estimates based on previous cycles.'
Where CTP% = percent of rated thermal power             0202-13 Rev. 15 for FCP equation Wt&deg;/o = percent of rated core flow Quad Cities 1 Operating License DLO I SLO                                               Values shown are expected to A.26     Cycle N-1 SLMCPR values       1.11 / 1.13 (Optima2)                                   continue to be used in Q1C22 based on Westinghouse calculated values for previous Optima2 cycles. These are the values currently in Quad Cities Tech Specs.
Q2C21 actual limits of 1.43 BOC -
c    I+ 7.
14,000 MWd/MTU / 1.45 from 14,000 A.27.a Estimated OLMCPRs to use Optima:?.:
t i lio
and Q1C21 actual limits of 1.43 from in design 1.44 for <11,000 MWD/MTU BOC to 13,275 MWdIMTU and 1.49 1.48 for X11,000 MWD/MTU from 13,275 MWd/MTU.
 
Quad Cities 1 Cycle 22 CDR               TODI NF1 0 236 Rev. 1 Page 13 of 29 Engineering input - no reference needed Estimates shown are for Base Case NSS and no EGOS. Based on Current Estimates based on previous cycles.'   Q2C21 actual limits of 1.43 BOC -
14,000 MWd/MTU / 1.45 from 14,000 A.27.a Estimated OLMCPRs to use   Optima:?.:                                     and Q1C21 actual limits of 1.43 from in design                   1.44 for <11,000 MWD/MTU                       BOC to 13,275 MWdIMTU and 1.49 1.48 for X11,000 MWD/MTU                       from 13,275 MWd/MTU.
MCPR OL must be >1.41 for LOCA analysis per Westinghouse document NF-BEX-06-44-P (April 2006) and Quad Cities Design Analysis OPTIMA2-TR021 QC-LOCA Rev. 5.
MCPR OL must be >1.41 for LOCA analysis per Westinghouse document NF-BEX-06-44-P (April 2006) and Quad Cities Design Analysis OPTIMA2-TR021 QC-LOCA Rev. 5.
N/A - OPRM amplitude setpoint values will be A.27.b Range of acceptable OPRM adjusted on the OPRM system and will not set the cycle OLMCPR limit Engineering input - no reference amplitude setpoint values needed Assume no setpoint change for design
N/A - OPRM amplitude setpoint values will be Range of acceptable OPRM   adjusted on the OPRM system and will not set the A.27.b                              cycle OLMCPR limit                               Engineering input - no reference amplitude setpoint values                                                   needed Assume no setpoint change for design
'The values include margin to the best-estimate expected OLMCPR. The SLMCPR values may change due to the issue identified in IR 1129289, which would impact the Q1 C22 OLMCPR. The impact of the SLMCPR issue for Q1 C22 is unknown, and these values remain a reasonable estimate of the Q1C22 OLMCPR..nom I, (l t)l4Ii Quad Cities 1 Cycle 22 IR TODI N F1000236 Rev. 1 Page 14 of 29 Lattice Burnup Bottom Plenum Middle Plenum GWd/MTU Bottom and Middle and To NF-BEX-09-188 Rev. 0 (Q2C21 RLR) 0 8.50 8.50 8.60 Based on limiting Q2C21 UE21 bundle design with --0.1 reduction in 129.009.00 9.20 MAPLHGR limits
'The values include margin to the best-estimate expected OLMCPR. The SLMCPR values may change due to the issue identified in IR 1129289, which would impact the Q1 C22 OLMCPR. The impact of the SLMCPR issue for Q1 C22 is unknown, and these values remain a reasonable estimate of the Q1C22 OLMCPR.
.A 28 APLHGR Limits to use in 24 9 50950.d i..9.80 Better MAPLHGR estim te b es gn a s may e 30 9 50950 9 80 used based on preliminary
                                                                                                          .nom I, (l t)l4Ii
...MAPLHGR li i m t calculations 50 9.50 9.50 9.90 performed by Westinghouse during the design as they become available.
 
60 9,50 9.50 9.80 72 9.50 9.50 9.80 GWD/MT LHGR Li mit LHGR Limits to use in 0 13.11 NF-BEX-09-188 Rev. 0 (Q2C21 RLR)
Quad Cities 1 Cycle 22         IR             TODI N F1000236 Rev. 1 Page 14 of 29 Lattice Burnup           Bottom Plenum Middle Plenum GWd/MTU     Bottom   and Middle     and To       NF-BEX-09-188 Rev. 0 (Q2C21 RLR) 0       8.50       8.50         8.60       Based on limiting Q2C21 UE21 bundle design with --0.1 reduction in 12      9.00        9.00         9.20       MAPLHGR limits .
A.29 design 14.0 13.11 72.0 6.48 Standard OPTJMA2 LHGR limits A 30 Transient LHGR Limits to N/A.use in design Transient LHGR limits not used Off-rated thermal limits to Assume off-rated thermal limits will be A.31 use in desi See NF-BEX-09-1 B8 (Q2C21 RLR) similar to those calculated for Q2C21.
A . 28 APLHGR Limits to use in         24       9 . 50      9 . 50      9.80 d es i gn                                                                    Better MAPLHGR estim ate s may be used based on preliminary 30      9 . 50      9 . 50      9. 80    MAPLHGR li m i t calculations 50       9.50       9.50                   performed by Westinghouse during 9.90 the design as they become available.
n g These multipliers can be found in the NF-
60       9,50       9.50         9.80 72       9.50       9.50         9.80 GWD/MT     LHGR Li mit 0          13.11                             NF-BEX-09-188 Rev. 0 (Q2C21 RLR)
-.BEX-09-188 Rev. 0.
LHGR Limits to use in A.29                                 14.0         13.11 design 72.0         6.48                             Standard OPTJMA2 LHGR limits A . 30 Transient LHGR Limits to use in design              N/A                                               Transient LHGR limits not used Assume off-rated thermal limits will be A.31   Off-rated thermal limits to                                                  similar to those calculated for Q2C21.
Quad Cities 1 Cycle 22 CDIR TODI NF1 236 Rev. 1 Page 15 of 29 OLMCPR1.41.GE SIL 320 Supplement 3 For fluence related bow:
use in desi g n            See NF-BEX-09-1 B8 (Q2C21 RLR)
NF-AB-105 Revision 13 GE SIL 320 Supplement 3 summation criterion <2 Bundles with X25 000 effective inch-(C lattice criterion).
These multipliers can be found in the NF-
, days and less than 40 GWD/MT Fuel channel distortion For shadow corrosion (Inch-days), at FOG Q1 C22 channel exposure may be monitored A.32 limitations and mitigation
-.                                                                                   BEX-09-188 Rev. 0.
, no cell shall contain:
 
rather than being re-channeled strategy Any bundle with 225,000 effective inch-days and MCPR OL must be X1.41 for Channel 240 GWDIMT channel exposure Distortion analysis per Westinghouse document NF-BEX-10-70 Rev. 0 OR (May 21, 2010), which transmitted the 3 or more bundles with x20,000 effective inch-report BTF 0!9-0275, rev 2, days and X30 GWD/MT channel exposure."Westinghouse methodology for enhanced channel bow indications." I^rl 1`t2" Quad Cities 1 Cycle 22 CDIR TODI NF1000236 Rev. 1 Page 16 of 29 Minimum allowed calculated reactor period and maximum allowed single notch worth for in-sequence notch worth determinations A.33.a Individual notch worths between 04-36 and total worth between 36-48 must be less than the Ak that would result in a 50 second period NF-AB-130-2 220 Rev. 8 Notch worths must consider sequence steps beyond Groups 1-4 (Groups 7,8,9, and/or 10) as necessary to satisfy requirements up to +3.5% keff around critical.
Quad Cities 1 Cycle 22 CDIR                   TODI NF1           236 Rev. 1 Page 15 of 29 OLMCPR        1.41.
A.33.b Range around the expected cold critical eigenvalue when the notch worth and step worth limits are to be enforced-1.2&deg;1 to +3.5% keff around the expected cold critical eigenvalue evaluated at 120 OF and 320 T.
GE SIL 320 Supplement 3 For fluence related bow:
NF-AB-130-2620 Rev. 8 Bounding range based on uncertainties for non BOG criticals Step worth (sum of all notch worths between movement limits) must be less than the minimum of:
NF-AB-105 Revision 13 GE SIL 320 Supplement 3 summation criterion <2 (C lattice criterion).                          Bundles with X25 , 000 effective inch-days and less than 40 GWD/MT Fuel channel distortion   For shadow corrosion (Inch-days), at FOG Q1 C22 ,    channel exposure may be monitored A.32 limitations and mitigation no cell shall contain:                               rather than being re-channeled strategy MCPR OL must be X1.41 for Channel Any bundle with 225,000 effective inch-days and 240 GWDIMT channel exposure                     Distortion analysis per Westinghouse document NF-BEX-10-70 Rev. 0 OR                                             (May 21, 2010), which transmitted the 3 or more bundles with x20,000 effective inch- report BTF 0!9-0275, rev 2, days and X30 GWD/MT channel exposure.           "Westinghouse methodology for enhanced channel bow indications."
0.005 Ak NF-AB-130-2620 Rev. 8 Maximum rod step worth or (7/64)* [ ko- Wl-0.0112 ] Ak)
I^rl 1`t2"
Notch worths must consider A.33.c allowed in the defined range around cold criticality for the where sequence steps beyond Groups 1-4 startup sequence ko= the cold critical target keff at each exposure (Groups 7,8,9, and/or 10) as point of interest necessary to satisfy requirements up kARI = the cold all rods inserted keff at each to +3.5% keff around critical.
 
exposure point of interest A.A.35 Code packages which are to be utilized in the determination of cold in-sequence notch worth values Core loading restrictions to be used to minimize risk of abnormally high notch worths during in-sequence critical POLCA7 for official notch worth determination as reported in the CMR.
Quad Cities 1 Cycle 22 CDIR                       TODI NF1000236 Rev. 1 Page 16 of 29 NF-AB-130-2 220 Rev. 8 Minimum allowed calculated reactor period and maximum     Individual notch worths between 04-36 and total        Notch worths must consider A.33.a allowed single notch worth      worth between 36-48 must be less than the Ak that     sequence steps beyond Groups 1-4 for in-sequence notch worth      would result in a 50 second period                     (Groups 7,8,9, and/or 10) as determinations                                                                          necessary to satisfy requirements up to +3.5% keff around critical.
Criteria must also be satisfied in MICROBURN-B2.
Range around the expected cold critical eigenvalue when                                                           NF-AB-130-2620 Rev. 8 A.33.b the notch worth and step         -1.2&deg;1 to +3.5% keff around the expected cold critical worth limits are to be          eigenvalue evaluated at 120 OF and 320 T.             Bounding range based on enforced                                                                                uncertainties for non BOG criticals Step worth (sum of all notch worths between movement limits) must be less than the minimum of:
Minimize clumping of once-burned bundles in high importance areas, such as between the edge of the fresh fuel zone and the periphery Engineering input No reference needed ATI 271647
0.005 Ak                                               NF-AB-130-2620 Rev. 8 or Maximum rod step worth         (7/64)   * [ ko- Wl-0.0112 ] Ak)                       Notch worths must consider allowed in the defined range   where A.33.c                                                                                        sequence steps beyond Groups 1-4 around cold criticality for the startup sequence               ko= the cold critical target keff at each exposure     (Groups 7,8,9, and/or 10) as point of interest                                 necessary to satisfy requirements up kARI = the cold all rods inserted keff at each         to +3.5% keff around critical.
-02,"Determine the apparent cause of the high notch worths for D3C19 and address the other recommendations in the lR." 1/14/2005.
exposure point of interest Code packages which are to be utilized in the             POLCA7 for official notch worth determination as A. determination of cold in-       reported in the CMR. Criteria must also be satisfied    Engineering input No reference sequence notch worth            in MICROBURN-B2.                                        needed values Core loading restrictions to                                                           ATI 271647 -02, "Determine the be used to minimize risk of     Minimize clumping of once-burned bundles in high        apparent cause of the high notch A.35  abnormally high notch          importance areas, such as between the edge of the       worths for D3C19 and address the worths during in-sequence      fresh fuel zone and the periphery                       other recommendations in the lR."
Quad Cities 1 Cycle 22 CIR TOQI NF1 000236 Rev. 1 Page 17 of 29 As documented in OPTIMA2-TR038QC-EOP, Rev.1,"SVEA-96 OPTIMA2 Fuel Input to the Emergency Operating Procedures." Westinghouse letter NF-BEX-06-282 Parameters XB-cld-nat, XB-hot-nat, and MSWBP (dated 12/14/06) contains the must be verified for 01 C22.
critical                                                                                1/14/2005.
Optimal EPG parameters.
 
A.36 Hot SLC SDM requirement is ^ 0.38% Ak/k and QGA 101 Rev. 13.
Quad Cities 1 Cycle 22 CIR                         TOQI NF1 000236 Rev. 1 Page 17 of 29 As documented in OPTIMA2-TR038QC-EOP, Rev.1, "SVEA-96 OPTIMA2 Fuel Input to the Emergency Operating Procedures."
EOP/EPG/SAG Parameters 700 ppm natural boron consistent with OPTIMA2-TR038QC-EOP, Rev.1.
Westinghouse letter NF-BEX-06-282 Parameters XB-cld-nat, XB-hot-nat, and MSWBP             (dated 12/14/06) contains the must be verified for 01 C22.                             Optimal EPG parameters.
MSBWP is to include determination Reactor is shutdown by 0.38% Ak/k xenon free at the of shutdown conditions at most most reactive temperature with the strongest rod at limiting temperature.
Hot SLC SDM requirement is ^ 0.38% Ak/k and               QGA 101 Rev. 13.
position 48.
A.36 EOP/EPG/SAG Parameters   700 ppm natural boron consistent with OPTIMA2-TR038QC-EOP, Rev.1.
QGA 101 requires all other rods inserted to at least position 04. The site desired alternative to 04 is to evaluate all rods at 02 and change QGA 101 to require that all rods in to at least position 02 for MSBWP.NF-AB-110-3040 Rev. 11.
MSBWP is to include determination Reactor is shutdown by 0.38% Ak/k xenon free at the       of shutdown conditions at most most reactive temperature with the strongest rod at       limiting temperature.
Proposed core designs and rod MCO < 0.1 % per GENE MCO analysis model patterns will be checked via GENE MCO analysis model. This will A.37 MCO limit Maximum bundle RPF < 1.65 and no more than 3 require performance of calculations to bundles with RPF > 1.60 within 3X3 array.
position 48. QGA 101 requires all other rods inserted to at least position 04. The site desired alternative to 04 is to evaluate all rods at 02 and change QGA 101 to require that all rods in to at least position 02 for MSBWP.
determine moisture carryover compliance based on results from MICROBURN-B2. RPF criteria are from NF T&RM NF-AB-110-3040
NF-AB-110-3040 Rev. 11.
__m._Rev. 11.
Proposed core designs and rod MCO < 0.1 % per GENE MCO analysis model                 patterns will be checked via GENE MCO analysis model. This will A.37 MCO limit                                                                         require performance of calculations to Maximum bundle RPF < 1.65 and no more than 3 bundles with RPF > 1.60 within 3X3 array.               determine moisture carryover compliance based on results from MICROBURN-B2. RPF criteria are from NF T&RM NF-AB-110-3040
Quad Cities 1 Cycle 22 CDI R TODI NF1 236 Rev. 1 Page 18 of 29 Final Task Report for TSD DQW04-024 (Attachment 1 to NF-BEX-05-155 1.EOC CAVEX < 37,000 MWD/MT Rev. 0)2.Initial core weight < 126.0 MTU A.38 Decay heat analysis 3.Initial core average enrichment between 3.90 and These criteria are the limits of 5.00 w/o U-235 (treat all fuel as fresh).
__m._                                                          Rev. 11.
applicability of the Westinghouse decay heat analysis. These criteria were also assumed in the Westinghouse evaluation of alternate source term.
 
GNF- GE-NE-A22-401033-02 Rev. 0, Axial and radial power profiles consistent (similar)
Quad Cities 1 Cycle 22 CDI R                       TODI NF1           236 Rev. 1 Page 18 of 29 Final Task Report for TSD DQW04-024 (Attachment 1 to NF-BEX-05-155
August 2000.
: 1. EOC CAVEX < 37,000 MWD/MT                           Rev. 0)
A.39 Fluence assumptions with past analyses -- 1 natural U node at bottom, 2 at Westinghouse - NE BEX 45-114 top; similar radial per profiles loading least reactive Rev. 0 8f23f45.fuel on edge
: 2. Initial core weight < 126.0 MTU A.38 Decay heat analysis         3. Initial core average enrichment between 3.90 and     These criteria are the limits of 5.00 w/o U-235 (treat all fuel as fresh).           applicability of the Westinghouse decay heat analysis. These criteria were also assumed in the Westinghouse evaluation of alternate source term.
, EC 349583 Rev. 1 Section 4.1.16.
GNF- GE-NE-A22-401033-02 Rev. 0, Axial and radial power profiles consistent (similar)   August 2000.
WCAP 16081-P-A Addendums I and CPR correlation limitations 2 to CPR correlation have been A.40 on core design None.approved by NRC and incorporated into POLCA, WCMS, and MICROBURN-B2.-d0 Quad Cities 1 Cycle 22 CDIRTODI NF1 2 6 R ev. 1 Page 19 of 29 These criteria are bounding values, and they were the basis for spent fuel 1.Bundle enrichment range from 3.38 to 4.50 w/o pool heating calculations as U235 documented in AmerGen Calculation 2.Bundle uranium loading (nominal)
A.39 Fluence assumptions         with past analyses -- 1 natural U node at bottom, 2 at top; similar radial per profiles loading least reactive Westinghouse - NE BEX 45-114 fuel on edge                                            Rev. 0 , 8f23f45.
<197.0 kgU No, C-1101-202-E620-443, Revision 41 A Spent fuel pool gamma 3.Bundle exposure < 58,700 MWD/MTU 0, "PWR & BWR Isotopic Inventories
EC 349583 Rev. 1 Section 4.1.16.
.heating constraints 4.Average bundle power (reactor rated thermal for Spent Fuel Pool Gamma Heating power / # bundles)
WCAP 16081-P-A Addendums I and CPR correlation limitations                                                         2 to CPR correlation have been A.40                             None.                                                   approved by NRC and incorporated on core design into POLCA, WCMS, and MICROBURN-B2.
<5.586 MWt Study." The calculation states a pin 5.Radial Peaking Factor
                                                                                                                      -d      0
<2.00 exposure limit of 58,700 MWd/MTU 6.Axial Peaking Factor < 1.80 but EC 378207 documents that the limit is a bundle limit, not a pin limit as stated in the calculation.
 
Fuel which resided on the periphery (on the first or A.42 Fuel promotion limitations second row) during the current or previous cycles NF-AB-110-2210 Rev. 12 should not be p romoted inwardast the fourth row.
Quad Cities 1 Cycle 22 CDIR                      TODI NF1             2 6 Rev. 1 Page 19 of 29 These criteria are bounding values, and they were the basis for spent fuel
Feedwater Temperature Operating Domain based on nominal FW temperature of: TODI QDC-10-026 Rev. 0 (T)
: 1. Bundle enrichment range from 3.38 to 4.50 w/o   pool heating calculations as U235                                             documented in AmerGen Calculation
A 43 Reactor dome pressure and
: 2. Bundle uranium loading (nominal) < 197.0 kgU       No, C-1101-202-E620-443, Revision Spent fuel pool gamma     3. Bundle exposure < 58,700 MWD/MTU                 0, "PWR & BWR Isotopic Inventories A . 41 heating constraints       4. Average bundle power (reactor rated thermal       for Spent Fuel Pool Gamma Heating power / # bundles) < 5.586 MWt                   Study." The calculation states a pin
-98`(Q/Qs)2 + 242.78'(Q/QR) + 210.82
: 5. Radial Peaking Factor < 2.00                     exposure limit of 58,700 MWd/MTU
.feedwater temperature Where: Q/QR is CTP/
: 6. Axial Peaking Factor < 1.80                       but EC 378207 documents that the limit is a bundle limit, not a pin limit as stated in the calculation.
rated CTP NF-BEX-05-24, Revision 2.(Pressure)
Fuel which resided on the periphery (on the first or A.42   Fuel promotion limitations second row) during the current or previous cycles   NF-AB- 110-2210 Rev. 12 should not be promoted inward ast the fourth row.
Q1 C21 Reactor Dome Pressure OratinDomain Quad Cities 1 Cycle 22 COIR TODII NF1236 Rev. 1 Page 20 of 29 Engineering input - No reference needed 1. Face adjacent shuffles within a cell increases shuffle time dramatically 1.Eliminate face adjacent within cell shuffles if
Feedwater Temperature Operating Domain based on nominal FW temperature of:
: 2. Cross quadrant shuffles will affect possible ability to move to another quadrant A 44 Miscellaneous Shuffling should an SRM fail during shuffling
TODI QDC-10-026 Rev. 0 (T)
.Restrictions 2.Avoid cross quadrant shuffles (i.e. compromise quadrant 3.Control cell re-use independence).
A . 43 Reactor dome pressure and -98`(Q/Qs)2 + 242.78'(Q/QR) + 210.82 feedwater temperature                                                           NF-BEX-05-24, Revision 2.
Where: Q/QR is CTP/ rated CTP (Pressure)
Q1 C21 Reactor Dome Pressure O      ratin Domain
 
Quad Cities 1 Cycle 22 COIR                     TODII NF1          236 Rev. 1 Page 20 of 29 Engineering input - No reference needed
: 1. Face adjacent shuffles within a cell increases shuffle time dramatically
: 1. Eliminate face adjacent within cell shuffles if 2. Cross quadrant shuffles will affect possible ability to move to another quadrant
. Miscellaneous Shuffling                                                      should an SRM fail during shuffling A 44 Restrictions            2. Avoid cross quadrant shuffles (i.e. compromise quadrant independence).
: 3. Control cell re-use
: 3. Insure that any bundle that was controlled during the last sequence in Q1C21 will have adequate periods of uncontrolled operation before being controlled in QiC22.
: 3. Insure that any bundle that was controlled during the last sequence in Q1C21 will have adequate periods of uncontrolled operation before being controlled in QiC22.
A.45 Axial Peaking for small Maximum peak in top half of core (node 20) < 1.6 Westinghouse memo NF-BEX break LOCA Anal sis 246 ,1110 Quad Cities 1 Cycle 22 CD1AT OD I NF1000236 Rev. 1 Page 21 of 29 Item Fuel Bundle Design Data Related It em s Criteria Value Referen ces/Comments a8 Gd rods per lattice CN-EXELONBWR-05-24, Revision 0 Gd w/o based on Boraflex degradation and Gd w/o as OPTIMA2-TRO25DQ-SFP, "Optimal Spent Fuel defined in CN-EXELONBWR-Pool & New Fuel Vault Evaluation --
Axial Peaking for small                                                   Westinghouse memo NF-BEX A.45                        Maximum peak in top half of core (node 20) < 1.6 break LOCA Anal sis                                                       246
05-24 Rev. 0 Dresden/Quad Cities" Rev. 0 Average enrichment of the Boraflex Degradation assumptions to be used B.1 Spent fuel pool criticality criteria peripheral rods in each lattice based on preliminary interpretation of NETCO is less than the lattice-average Boraflex degradation results, enrichment a)7 inch uniform gap b)50% loss in thickness No Gadolinia rods on bundle c)5% loss in width periphery Face adjacent Gd rods shall be counted as a 30 cm top and 15 cm bottom single rod.
                                                                                                          ,1110
 
Quad Cities 1 Cycle 22 CD1A                      TODI NF1000236 Rev. 1 Page 21 of 29 Item Fuel Bundle Design Data Related Items                Value                           References/Comments Criteria a8 Gd rods per lattice CN-EXELONBWR-05-24, Revision 0 Gd w/o based on Boraflex degradation and Gd w/o as       OPTIMA2-TRO25DQ-SFP, "Optimal Spent Fuel defined in CN-EXELONBWR-         Pool & New Fuel Vault Evaluation --
05-24 Rev. 0                     Dresden/Quad Cities" Rev. 0 Average enrichment of the       Boraflex Degradation assumptions to be used B.1 Spent fuel pool criticality criteria   peripheral rods in each lattice based on preliminary interpretation of NETCO is less than the lattice-average Boraflex degradation results, enrichment                               a) 7 inch uniform gap b) 50% loss in thickness No Gadolinia rods on bundle             c) 5% loss in width periphery Face adjacent Gd rods shall be counted as a 30 cm top and 15 cm bottom       single rod.
Natural U blankets.
Natural U blankets.
OPTIMA2-TRO25QQ-SFP, "Optima2 Spent Fuel Pool & New Fuel Vault Evaluation -
OPTIMA2-TRO25QQ-SFP, "Optima2 Spent Fuel Pool & New Fuel Vault Evaluation -
Max 235U Enrichment:4.95%Dresden/Quad Cities" Rev. 0 B.2 New fuel vault criticality criteria Min. # Gd rodsa Min Gd 203 Enrichment 5.5%
Max 235U Enrichment: 4.95%       Dresden/Quad Cities" Rev. 0 B.2 New fuel vault criticality criteria     Min. # Gd rods            a Min Gd 203 Enrichment 5.5%       Face adjacent Gd rods shall be counted as a single rod.
Face adjacent Gd rods shall be counted as a single rod.
1a
1a Quad Cities 1 Cycle 22 CDIRTOQt NF1 236 Rev. 1 Page 22 of 29 Section B -
 
Item Fuel Bundle Design Data Related It em s Value RefereeComments Criteria Max 236U Enrichment:5 0%NRC Certificate of Compliance No. 9292 for the
Quad Cities 1 Cycle 22 CDIR                        TOQt NF1         236 Rev. 1 Page 22 of 29 Section B -
.Min Gd203 Enrichment 4.0%
Item   Fuel Bundle Design Data Related Items                Value                             Referee    Comments Criteria Max 236U Enrichment: 5 . 0%     NRC Certificate of Compliance No. 9292 for the Min Gd203 Enrichment 4.0%       Model No. PATRIOT Package, Revision 6, No Gd rods on edge or corner     Docket Number 71-9292.
Model No. PATRIOT Package, Revision 6, No Gd rods on edge or corner Docket Number 71-9292.
Fuel shipping containers criticality     # Gd rods (96 fuel B.3                                                Gd rod diagonally rods)       Letter, Brian Beebe "to Daniel Redden, NF-@EX-criteria                                                             12 symmetric
B.3 Fuel shipping containers criticality
                                                # Gd rods (92 fuel rods) 10     05-166 Revision 0, Additional Information on Gd rod diagonally symmetric   Patriot Container Certificate of Compliance " ,
# Gd Gd rod rods (96 fuel diagonally rods) 12 symmetric Letter, Brian Beebe to Daniel Redden, NF-@EX-criteria# Gd rods (92 fuel rods) 10
                                                # Gd rods (84 fuel rods) 8       October 21, 2005 and NF-BEX-06-67, "Additional Min 2 Lid rods per quadrant Information on Patriot Container Certificate of Compliance " , March 9 2006.
" Additional Information on 05-166 Revision 0," Gd rod diagonally symmetric Patriot Container Certificate of Compliance
10 CFR 70.24 criticality monitoring B&deg;4  exem tionS                              N/A                               No criticality monitoring exemptions.
,"# Gd rods (84 fuel rods)8 October 21, 2005 and NF-BEX-06-67, Additional Min 2 Lid rods per quadrant Information on Patriot Container Certificate of
B.5   Fuel type to be manufactured for this reload                                  Optimal                           Engineering input - No reference needed BTK04-164, "Mechanical Data Input to Nuclear and Thermal-   BTKO4-164, "Mechanical Data Input to Nuclear B.6 Source for fuel product line dimensions   Hydraulic Design, Quad Cities   and Thermal-Hydraulic Design, Quad Cities I & 2 1 & 2 and Dresden 2 & 3         and Dresden 2 & 3 SVEA-96 Optimal" , Rev. 1, SVEA-96 Optimal", Rev. 1,       3/24/05 3/24/05 B.7 Channel bow assumptions                   None                             No channel bow assumptions for bundle desi n.
" Compliance
CPR correlation limitations on bundle     C} ptima2 bet sub- bundle factors 6.8                                            must be between 0 . 89 R and    SEA WCAP-1            P. Optimal CPR correlation design                                                                     is valid within this range.
, March 9 2006.
mu No explicit lattice or power/exposure peaking B.9  Limitations on lattice power peaking      None                            limits for bundle design. This includes local peaking factors.
B&deg;4 10 CFR 70.24 criticality monitoring N/A No criticality monitoring exemptions.
Core loading restrictions due to "other" Load bundles with high 8.10                                           control/inch days in non-       Engineering input - no reference needed factors nodded locations on periphery.
exem tionS B.5 Fuel type to be manufactured for this Optimal Engineering input - No reference needed reload BTK04-164, "Mechanical Data Input to Nuclear and Thermal-BTKO4-164, "Mechanical Data Input to Nuclear B.6 Source for fuel product line dimensions Hydraulic Design, Quad Cities and Thermal-Hydraulic Design, Quad Cities I & 2
I i I =(,0
" 1& 2 and Dresden 2 & 3
 
, Rev. 1, and Dresden 2 & 3 SVEA-96 Optimal SVEA-96 Optimal", Rev. 1, 3/24/05 3/24/05 B.7 Channel bow assumptions None No channel bow assumptions for bundle desi n.
Quid Cities 1 Cycle 22 CDIR                       T'OD NF1 0236 Rev.
6.8 CPR correlation limitations on bundle 89 R and factors C}must be ptima2 bet sub-between 0 bundle u SEA WCAP-1P.Optimal CPR correlation design.m is valid within this range.
No explicit lattice or power/exposure peaking B.9Limitations on lattice power peakingNone limits for bundle design. This includes local peaking factors.
Core loading restrictions due to "other" Load bundles with high 8.10 factors control/inch days in non-Engineering input - no reference needed nodded locations on periphery.
I i I=(,0 Quid Cities 1 Cycle 22 CDIR T'OD NF1 0236 Rev.
Page 23 of 29 Section B -
Page 23 of 29 Section B -
It em Fuel Bundle Des ign Data R e lated items Value RefereComments Criteria 1.No gad rods face adjacent to central portion of water cross or face adjacent to other gad rods (gad rods 1.Basis is modeling uncertainty. No explicit Limitations on the are allowed to be next to reference.
Item    Fuel Bundle Design Data Related items               Value                             Refere    Comments Criteria
B.11 piacement/enrichment of gadolinia connecting wings of water bearing rods cross); no gad rods in part 2.Appendix B of Westinghouse DR/QC length rods; no gad rods in contract.Manufacturing instrumentation outer row of bundle sensitivity 2.Concentrations in a single rod must differ b z 2 w/o Standard concentrations Maximum 8 w/o B.12 Limitations on the amount of gadolinia allowed in the fuel pellets 3 concentrations per reload Appendix B of Westinghouse DR/QC contract 2 concentrations per rod type (including zero concentration Engineering input - no reference needed.
: 1. No gad rods face adjacent to central portion of water cross or face adjacent to other gad rods (gad rods     1. Basis is modeling uncertainty. No explicit Limitations on the                           are allowed to be next to       reference.
13,13 Target bundle enrichments for this reload design 3.84-4.20%
B.11 piacement/enrichment of gadolinia           connecting wings of water bearing rods                                 cross); no gad rods in part 2. Appendix B of Westinghouse DR/QC length rods; no gad rods in     contract. Manufacturing instrumentation outer row of bundle             sensitivity
Bundle enrichments in the lower part of this range may be used to help maintain the peak hot excess reactivity, if necessary.
: 2. Concentrations in a single rod must differ b z 2 w/o Standard concentrations Maximum 8 w/o Limitations on the amount of gadolinia B.12                                                                          Appendix B of Westinghouse DR/QC contract allowed in the fuel pellets             3 concentrations per reload 2 concentrations per rod type (including zero concentration Engineering input - no reference needed.
8.14 Target minimum length for bottom fuel None Westinghouse fuel design does not have this zone limitation on the bundle desi n.
Target bundle enrichments for this 13,13                                          3.84-4.20%                     Bundle enrichments in the lower part of this reload design range may be used to help maintain the peak hot excess reactivity, if necessary.
Quad Cities 1 Cycle 22 CDI R TOD F1 236 Rev. 1 Page 24 of 29 Sec tion B -Item F u el Bundle Design Data Related Items Value Referen ces/Comments Criteria EC 349583, `Implement Westinghouse Optimal Nuclear Fuel" 30 cm at top of bundle (2 nodes)Consistent with previous Optimal core designs.
8.14 Target minimum length for bottom fuel                                   Westinghouse fuel design does not have this zone                                     None limitation on the bundle desi n.
Length of natural U zones at top and 8.15 bottom of bundle 15 cm at bottom of bundle (1 A more detailed reactor fluence evaluation is node)required before the natural U top zone can be decreased to less than 30 cm per information resented in EC 349883.
 
Engineering input - no reference needed W CBH evaluation tool has been developed for use during bundle design. Design strategy is to keep the final maximum MFLCPR with CBH penalty added to below, or as close as possible, Maximum allowable control blade Select pin enrichments to to the design goal of 0.90.
Quad Cities 1 Cycle 22 CDI R                   TOD F1 236 Rev. 1 Page 24 of 29 Section B -
8.16 history delta MFLCPR penalty minimize CBH impact on CPR W will also evaluate proposed rod patterns for GBH impacts.
Item   Fuel Bundle Design Data Related Items               Value                         References/Comments Criteria EC 349583, `Implement Westinghouse Optimal Nuclear Fuel" 30 cm at top of bundle (2 nodes)                       Consistent with previous Optimal core designs.
Length of natural U zones at top and 8.15 bottom of bundle                         15 cm at bottom of bundle (1 A more detailed reactor fluence evaluation is node)                       required before the natural U top zone can be decreased to less than 30 cm per information resented in EC 349883.
Engineering input - no reference needed W CBH evaluation tool has been developed for use during bundle design. Design strategy is to keep the final maximum MFLCPR with CBH penalty added to below, or as close as possible, Maximum allowable control blade         Select pin enrichments to   to the design goal of 0.90.
8.16   history delta MFLCPR penalty             minimize CBH impact on CPR W will also evaluate proposed rod patterns for GBH impacts.
CBH will impact the bundle design and the extent to which fresh bundles are placed in cells that are tanned to be controlled during operation.
CBH will impact the bundle design and the extent to which fresh bundles are placed in cells that are tanned to be controlled during operation.
6 17 Maximum rotated bundle delta CPR None Westinghouse methods do not have this limit
Maximum rotated bundle delta CPR         None                         Westinghouse methods do not have this limit 6 . 17 allowed Fuel manufacturing constraints which                                 No special restraints.
.allowed 8 18 Fuel manufacturing constraints which None No special restraints.
8 . 18                                          None may be challen ed                                                                             ____
.may be challen ed
No explicit lattice or power{exposure peaking Exposure peaking limits on the lattice                               limits for bundle design. This includes local B , 19                                          None designs                                                                   akin factors.
____No explicit lattice or power{exposure peaking B 19 Exposure peaking limits on the lattice None limits for bundle design. This includes local
2
, designs akinfactors.2 Quad Cities 1 Cycle 22 C DIR TODI NF1 000 236 Rev. 1 Page 25 of 29 Section B It Fue l Bundle Des ign Data Related It em s Value Re e fersCorntits Criteria Current limitation of Westinghouse automated fuel rod loading machines is 15 rod types. If >15 Manufacturing preference is rod types, then special hand loading is required.
 
20 B Maximum number of unique rod types maximum of 15 fuel rod types Note that unique rod types are characterized by
Quad Cities 1 Cycle 22 C DIR                       TODI NF1 000236 Rev. 1 Page 25 of 29 Section B It     Fuel Bundle Design Data Related Items                Value                             Reefers      Corn    tits Criteria Current limitation of Westinghouse automated fuel rod loading machines is 15 rod types. If >15 Manufacturing preference is     rod types, then special hand loading is required.
.in a given Optima2 bundle the enrichment, gadolinia concentration and axial type distribution, rod length (113 PLR, 2/3 PLR, or full
maximum of 15 fuel rod types   Note that unique rod types are characterized by B . 20 Maximum number of unique rod types      in a given Optima2 bundle       the enrichment, gadolinia concentration and axial type                           distribution, rod length (113 PLR, 2/3 PLR, or full length rods), and whether the rod is a tie rod, spacer-capture rod, or a regular fuel rod.
Per Westinghouse process, gadolinia power 8.21  Target gadolinium suppression penalty    Not Allowed                    -suppression is not allowed for Optimal fuel.
SER-WCAP-16081-P Optirna2 CPR correlation is valid within this range of R-factor.
Optimal sub-bundle R-factor 8.22  Target R-Factors                        must be Z! 0.89 and 5 1.11      This will be checked and documented during bundle design work to ensure that the bundle designs comply with the R-factor requirements, rather than relying on the check to be done within the CPR correlation subroutine Westinghouse design methods do not require or use a target local peaking factor. Peaking 8.23  Target local peaking factors            N/A factors must meet the design criteria identified in other sections of the CDIR.
NF-AB-440 Rev. 13 Verify lattices are bounded by  NF-AB-0-1002 Rev. 2 (pending WCMS Optimal REMACCX PO curve cutover) in NF-AB-440-1002 8.24  Lattice Design for Fuel Reliability (LRG vs.nodalnexposure) if If lattice does not satisfy limits, notify site that (LH R v s. o is used for core generic Optimal PO curve does not apply for this monitoring
_-            ^-                                                  c qLe i.e. will be altered slightly iil12ir
 
Quad Cities 1 Cycle 22 CDIR                TOM NF1000236 Rev. 1 Page 26 of 29 Section C -                            Value                      Retere    Comments Ite m M iscella neous Data Related Items Criteria NF-AB-440-1002 Rev. 2 (pending WCMS In accordance with Exelon C.1  Fuel conditioning guidelines                                          cutover) procedure                NF-AB-400-1003 Rev. 1 Engineering decision - no explicit reference.
Computer code(s) to be used for lattice PHOENIX-4 / POLCA-7 and  Goal is to provide adequate design margins in C.2  and core design, and any special CASMC)-4 / MICRC}BURN-B2  both code packages; differences will be version requirements managed b en ineerin judgment.
Standard Quad Cities      TOOl NF0400171 Rev. 0.
C.3  Power-flow map Power/Flow None planned.
Technical Specification changes for the                              SLMCPR is the only identified potential Technical C.4                                              SLMCPR may be required cycle                                                                Specification change.
based on SLMCPR analysis results C.5  Planned core component changes              None                      N/A Studies or projects in progress which C.6  could impact the reload design or          None                      None licensing i I (m /to
 
Quad Cities 1 Cycle 22 CDIR                    TODI N F 10. 0236 Rev. 1 Page 27 of 29 Section C -
item                                                          Value                      References/Comments Miscellaneous Data Related Items Criteria Bundle exposure <62 NF-AB-110-2210 Rev. 12.
GWD/MT Reg. Guide 1.183 "Alternative Radiological Peak rod average exposure <
Source Terms for Evaluating Design Basis 54 GI/MT Accidents at Nuclear Power Reactors';
Or, if Peak rod average exposure C.7  AST analysis limits                                                  QC design analyses QDC-0000-N-1267, Rev.
54 GWDlMT, then peak rod 1A, " Re-analysis of Fuel Handling Accident average LHGR must not (FHA) Using Alternative Source Terms " , and exceed 6 . 3 kW/ft .
QDC-0000-N-1268, Rev. 2A, "Re-analysis of Control Rod Drop Accident (CRDA) Using Maximum RPF must be <
Alternate Source Terms."
1.70.
C.8  Items which are critical to quality      See CDIR Attachment 1      Engineering Input Fuel Type(s) to be used in downstream                                Assume Optimal will be loaded in all future C.9                                            Optima2 multi-cycle analyses                                                  cycles for multi-cycle analyses C.9.1 Discharge Average Burnup Target          48 GWd/MTU                  Westinghouse DRJQC contract.
Design Cycle + 4 future C.9.2 Multi-Cycle Analysis Length                                          Engineering Input.
cycles,
 
Quad Cities 1 Cycle 22 C D IR                TODI NF1000236 Rev. 1 Page 28 of 29 Section D -
Item    Planned Change Related Items                      Value                      References/Comments Criteria ECs 374977, 365821, 373667, 366920.
The effects of the turbine modifications are D.1 Planned station modifications that may    Turbine Overhaul included in Exelon TODl QOC 10-27 Rev. 0, impact reload analysis                    Modifications            " OPL-W Parameters for Quad Cities Unit 1 Cycle 22 Transient Analysis", 9/15/14.
TODI QDC -1 0-026 Rev. 0, "Quad Cities Units I Planned setpoint changes that may        Feedwater Temperature D.2                                                                    & 2 Nominal Feed Water Temperature Curve " ,
impact reload analysis                    Operating Domain revision 8/24110.
Planned station minor modifications that D3  m impact reload analysis                  None .                    N/A D'4 Planned component changes that may impact reload analysis                    None                      N/A A
102,1110
 
Quad Cities 1 Cycle 22 CDIR                        TOO1 NF1 0        239 Rev. 1 Page 29 of 29 CDIR Atta c hm ent 1 (ite m C.8)
Q1C22 Exelon Fuel Design CTQ's t . Attempt to reduce redundancies during the reload design effort such that both Exelon and W share in the preparation and review of output documents.
: 2. Design in compliance with CDIR specifications and meet Major Milestones in the Design & Licensing Schedule.
: 3. W performs multi-cycle analysis consistent with Exelon EUP following completion of Reference Loading Pattern. Exelon will utilize this information for economic/amortization purposes.
: 4. Open access to W core and bundle design procedures by Exelon. Mutual agreement on how to resolve conflicts, documenting exceptions when taken.
: 5. Improve design quality through collaboration.
: 6. Attempt to minimize the number of shadow corrosion induced channel bow susceptible bundles within a bladed cell. The objective is to minimize Exelon`s risk of channel-control blade interference utilizing Exelon and industry operating experience,
: 7. Adhere to SIL 320 (latest supplement) loading pattern restrictions to minimize fluence gradient induced bow impacts on channel to rod interference. All fuel types are to be treated as GNF fuel from a SIL 320 perspective.
: 8. Resulting design is acceptable when satisfactory results are obtained with both POLCA-7 and MICROBURN-82.
: 9. Expanded notch worth criteria and step worth criteria are met and margin to the limits is communicated.
lt(72 /tO
 
ATTACHMENT 8 EGG Calculation QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan"
 
CC-AA-309-1 001 Revision 6 Page 1 of 9 ATTACHMENT 1 Design Analysis Major Revision Cover Sheet Page I of 5 Design Analysis (Major Revision)                                                  Last Page No. 9 B14 Analysis No.: '            QDC-0000-N-1 804                              Revision:      0
 
==Title:==
3                    Quad Cities Unit 1 Cycle 22 Core Loading Plan EC/ECR No.: 4              377652                                        Revision:      0 Station(s):                              Quad Cities                                            Component(s): "
Unit No.:                                1                                N/A Discipline: a                            NUDC Descrip. Code/Keyword: 'a                N01 Safety/OA Class:                          SR System Code:                              N/A Structure:        'a                      N/A CONTROLLED DOCUM ENT REFERENCES Document No.:                                              From/To        Document No.:                                    From/To OPTIMA2-TR038QC-EOP                                        From          QDC-0000-N-1806                                  To QDC-0000-N-1805                                            From QDC-0000-N-1653                                            From Is this Design Analysis Safeguards Information? 1e                              Yes[] Noo            If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions?"                      Yes[] No Z            If yes, ATI/AR#:        VIA This Design Analysis SUPERCEDES: 1e                        N/A                                                      in its entirety.
Description of Revision (list changed pages when all pages of original analysis were not changed): 19 Original Issue Seth Spooner                            -^J                                              / 1 e l t.^ r !
Preparers 20                    Brandon de Graaf                                        j                                y f fS hf Print Name                                  SI  Name            ^''          Date Method of Review: _'              Detailed Review                Alternate Calculatl      (attached) q          Testing q Reviewer: 22                    Eric Bruch                                                                              412 91 11 Print Name                                  Sign Name                          Date Review Notes: "                  Independent review 0                  Peer review Sep                  11 5.
ITPR : J oh n R e i mer        ^    ^ -^      ^      `^/      fzv r (Fm External Analyses Only)
External Approver: u            N/A                                      N/A                                            N/A Print Name                                  Sign Name                          Date Exelon Reviewer: 25              N/A                                      N/A                                              N/A Print Name                                  SI  Name                          Date Independent 3`d Party Review Reqd? m                          Yes      No Exelon Approver. 17              Jill Fisher                                                                              5          f
                                                                                ,1oLIfri.__-_
Print Name                                  Sin Name                            Date
  -f- Rtv;W o'                              cd          blade-    m aps b" A l    e,
 
Exek,n.                Nuclear Fuels - BWR Design          QDC-0000-N-1804 Rev. 0 Nuclear        Quad Cities Design Analysis                    Page 2 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN DESIGN ANALYSIS TABLE OF CONTENTS SECTION:                                          SUB-PAGE PAGE NO.
NO.
CC-AA- 309-1001 Major Revision Cover Page                  1                N/A Design Analysis Table of Contents                          2                N/A 1.0 Purpose                                                3                N/A 2.0 Inputs                                                  3                N/A 3.0 Assumptions                                            4                N/A 4.0 References                                              4                N/A 5.0 Identification of Computer Programs                    5                N/A 6.0 Method of Analysis                                      5                N/A 7.0 Numeric Analysis                                        6                N/A 8.0 Results                                                7                N/A 9.0 Conclusion                                              7                N/A Attachment A - Core Loading Plan                        Al - All            N/A Attachment B - Independent Review Information          B1 - B14            N/A
 
xel !,n.            Nuclear Fuels - BWR Design                                        QDC-0000-N-1804 Rev. 0 Nuclear          Quad Cities Design Analysis                                                                Page 3 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN List of Tables Table 1 - Computer Codes Used In Analysis .............................................................................5 Table 2 - CAKWFT 1.4 Output .....................................................................................................6 1.0    Purpose The purpose of this design analysis is to document the Core Loading Plan Revision 0 for Q1 C22 per T&RM NF-AB-130-2200, Revision 1 (Reference 4.3).
The current revision of this Design Analysis (Revision 0) is part of the work covered under the Fuel Change Package FCP 377652, Quad Cities Unit 1 Cycle 22 Reload Design (Reference 4.5).
2.0    Inputs 2.1    Core Loading by Assembly ID and Fuel Type The fuel assembly ID loading map is from the Westinghouse Reference Loading Pattern (Reference 4.6, pg Al -15, Al -16), modified with two changes: 1) replace bundle QAA21 1 in location 05-40 with bundle QAA1 39; and 2) replace bundle QAA213 in location 05 -22 with bundle QAA1 41. Both QAA1 39 and QAA1 41 were planned to be discharged in the Reference 4.6 loading pattern. These changes were made to address a suspected fuel failure detected in 01 C21 (see Reference 4.9 for loading changes).
The fuel type loading map is from the Q1C22 Design Analysis for the MICROBURN-132 basedeck (Reference 4.9). This design analysis incorporates the replaced bundles, therefore no modifications were made to the map.
2.2    Core Weight The as-built core weight is determined using the Q1C22 Design Analysis for the MICROBURN-B2 basedeck (Reference 4.9). Fresh fuel as-built bundle mass values were used as input to the deck, and the MICROBURN-B2 output file shows overall core weight as 124.83 MTU.
2.3    Control Blade Information This information is from the 01 C21 CLP (Reference 4.10) together with the changes described in Quad Cities blade replacement TODI (Reference 4.11), which shows details regarding planned blade replacements and shuffles during Q1 R21. Blade IQs are compared to Reference 4.4 Attachment 11 to determine which of the Marathon control blades are part of etch indication population.
 
Exe kn.                Nuclear Fuels - BWR Design              QDC-0000-N-1804 Rev. 0 Nuclear          Quad Cities Design Analysis                            Page 4 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 2.4  EPG/SAG Information Parts of this information are directly from the Reference 4.6 report (# of Optima2 bundles on page `A1-14') and Reference 4.7 Westinghouse report (Optimal active fuel lengths, total mass of U02, total mass of clad and channels on pages 9-12). There is also calculated information (Optima2 bundles' mass, total mass of Uranium), which is presented in Section 7.0 Numeric Analysis.
3.0  Assumptions None.
4.0  References 4.1. Exelon Procedure, CC-AA-309, Revision 10, "Control of Design Analyses".
4.2. Exelon T&RM, CC-AA-309-1001, Revision 6, "Guidelines For Preparation And Processing of Design Analyses".
4.3. Exelon T&RM, NF-AB-130-2200, Revision 1, "Core Loading Plan Generation".
4.4. Exelon T&RM, NF-AB-135-1410, Revision 8, "BWR Control Blade Lifetime Management".
4.5. Exelon Fuel Change Package 377652-000, "Quad Cities Unit 1 Cycle 22 Core Reload Design".
4.6. Westinghouse Report NF-BEX-10-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22", dated 12/3/10. (Attachment 14 to FCP 377652-000) 4.7. Exelon Design Analysis OPTiMA2-TRO38QC-EOP, Revision 1, "SVEA-96 OPTIMA2 Fuel Input to the Emergency Operating Procedures," dated 12/18/06. (QC record available via EDMS) 4.8. Westinghouse Report NF-BEX-10-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22", dated 11/18/10. (Attachment 18 to FCP 377652-000) 4.9. Exelon Design Analysis QDC-0000-N-1805, Revision 1, "Quad Cities 1 Cycle 22 MICROBURN- B2 Basedeck," dated April 2011.
4.10. Exelon Design Analysis QDC-0000-N-1 653, Revision 1, "Quad Cities 1 Cycle 21 Core Loading Plan," dated 5/11/09.
4.11. TODI QDC-11-025, Revision 1, "Q1 R21 Control Rod Blade and LPRM Replacements,"
dated 4/21/11.
 
Exekrn.              Nuclear Fuels - BWR Design                QDC-0000-N-1 804 Rev. 0 Nuclear        Quad Cities Design Analysis                          Page 5 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 5.0    Identification of Computer Programs Table 1 - Computer Codes Used In Analysis Cksum Name                                        Ref                DTSQA Info IiP/UX 11.1 CAKWFT (1.4)          N/A for PC        See DTSQA Database      EX0006886 Level CC 6.0    Method of Analysis This Design Analysis was prepared and reviewed in accordance with the governing procedure CC-AA-309 (Reference 4.1) and the associated T&RM CC-AA-309- 1001 (Reference 4.2). The Core Loading Plan Generation T&RM was utilized to create the CLP. This procedure can be found in Reference 4.3.
 
ExeIn.                      Nuclear Fuels - BWR Design                  QDC-0000-N-1804 Rev. 0 Nuclear              Quad Cities Design Analysis                                  Page 6 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 7.0 Numeric Analysis Core Average kW/ft This value is the rated core power (2957 MWt) divided by the summation of all the linear ft of fuel pellet stack in the core. The calculation includes part length correction as shown in the table below. The table also shows the linear ft calculation for each of the different bundle types loaded in the core by cycle.
The Gad rod correction is not required for Optimal fuel. This correction is for fuel types in which the Gad rods have a shorter length than full length rods, which must be accounted for in the linear ft of fuel calculation (GE 14 fuel has this characteristic). This cycle will contain all Optimal fuel, which contains Gad rods that -hayd do not have a
                                                                              ^sts^alu unique length.
sqj SAM Information for input into CAKWFT 1.4 was obtained from Reference 4.6.
Table 2 - CAKWFT 1.4 Output Unit/Cycle                                Quad Cities nit J. Cycle 22 Fuel Type                                    ODt mat      Oates              O-Number of Bundles                                236          268              220
          # Total rods per bundle                          96            96                96
          # Full length rods                                84          84                84
          # Gad rods                                      # N/A        #N/A              #N/A
          # Short Part length rods                          4            4                4
          # Long Part length rods                            8            8                8 Full length rode, In.                          145.28      145.28            145.28 Gad rod length, In.                            145.28      145.28            145.28 Short Part length rods, in.                    50.39        50.39            50.39 Long Part length rods, in.                    99.61          99.61            99.61 Total fuel length, In., per bundle          13201.96      13201.96          13201.96 Total fuel length, In. (by fuel type)      3115662.56 3538125.28            2904431.2 Total fuel length, ft. (by fuel type)    259638.5467 294843 .7733 242035.9333 Total fuel length, ft. (for the core)    796518.2533 Rated thermal power (MWth)                      2957 Core Average KW/ft Core Average kW/ft                          3.712 kW/ft


length rods), and whether the rod is a tie rod, spacer-capture rod, or a regular fuel rod.
Exe1n.               Nuclear Fuels - BWR Design             QDC-0000-N-1804 Rev. 0 Nuclear         Quad Cities Design Analysis                           Page 7 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN EOP Information Bundles mass = (Mass of U02 + Mass of Clad/channel)/# of assemblies For Optimal, using Reference 4.7, Table 1 values:
8.21 Target gadolinium suppression penalty Not Allowed Per Westinghouse process, gadolinia power-suppression is not allowed for Optimal fuel.
Optimal bundles mass = (316877 Ibm + 134824 Ibm)/724 = 623.9 Ibm Total mass of uranium is calculated per Reference 4.3. Total mass of U02 value is based on full core of Optimal fuel per Reference 4.7, Table 1:
SER-WCAP-16081-P Optirna2 CPR correlation is valid within this Optimal sub-bundle R-factor range of R-factor.
Total mass of uranium = total mass of U02 / ratio of the molecular weight of U02 to U 316877 Ibm / (270/238) 279321 Ibm Core weight is calculated per Reference 4.3. The core weight value is based on full core of Optimal fuel per Reference 4.7, Table 1:
8.22 Target R-Factors must be Z! 0.89 and 5 1.11 This will be checked and documented during bundle design work to ensure that the bundle designs comply with the R-factor requirements, rather than relying on the check to be done within the CPR correlation subroutine Westinghouse design methods do not require or 8.23 Target local peaking factors N/A use a target local peaking factor. Peaking factors must meet the design criteria identified in other sections of the CDIR.
Verify lattices are bounded by NF-AB-440 Rev. 13 NF-AB-0-1002 Rev. 2 (pending WCMS Optimal REMACCX PO curve cutover)8.24 Lattice Design for Fuel Reliability in NF-AB-440-1002 (LRG vs. nexposure) if nodal o (LHv R is used for core s.If lattice does not satisfy limits, notify site that monitoring generic Optimal PO curve does not apply for this
_-^-c qLei.e.will be altered slightly iil12ir Quad Cities 1 Cycle 22 CDIRTOM NF1000236 Rev. 1 Page 26 of 29 Ite m Section C -
Value RetereComments M isce lla neo us Data Related Items Criteria NF-AB-440-1002 Rev. 2 (pending WCMS C.1 Fuel conditioning guidelines In accordance with Exelon cutover)procedure NF-AB-400-1003 Rev. 1 Engineering decision - no explicit reference.
Computer code(s) to be used for lattice PHOENIX-4 / POLCA-7 and C.2 and core design, and any special CASMC)-4 / MICRC}BURN-B2 Goal is to provide adequate design margins in version requirements both code packages; differences will be managed b en ineerinjudgment.C.3 Power-flow map Standard Quad Cities TOOl NF0400171 Rev. 0.
Power/Flow None planned.
C.4 Technical Specification changes for the SLMCPR may be required SLMCPR is the only identified potential Technical cycle based on SLMCPR analysis Specification change.
results C.5 Planned core component changes None N/A Studies or projects in progress which C.6 could impact the reload design or None None licensing i I (m /to Quad Cities 1 Cycle 22 CDIR TODI N F 10.
0236 Rev. 1 Page 27 of 29 item Section C-Miscella neo us Data Related Items Criteria Value References/Comments Bundle exposure <62 NF-AB-110-2210 Rev. 12.
GWD/MT Peak rod average exposure <
Reg. Guide 1.183 "Alternative Radiological 54 GI/MT Source Terms for Evaluating Design Basis Or, if Accidents at Nuclear Power Reactors';
C.7 AST analysis limits Peak rod average exposure 54 GWDlMT, then peak rod QC design analyses QDC-0000-N-1267, Rev.
" average LHGR must not 1A, Re-analysis of Fuel Handling Accident exceed 6 3 kW/ft" (FHA) Using Alternative Source Terms , and..QDC-0000-N-1268, Rev. 2A, "Re-analysis of Maximum RPF must be <
Control Rod Drop Accident (CRDA) Using 1.70.Alternate Source Terms."C.8Items which are critical to quality See CDIR Attachment 1 Engineering Input C.9 Fuel Type(s) to be used in downstream Optima2 Assume Optimal will be loaded in all future multi-cycle analyses cycles for multi-cycle analyses C.9.1 Discharge Average Burnup Target 48 GWd/MTU Westinghouse DRJQC contract.
C.9.2 Multi-Cycle Analysis Length Design Cycle + 4 future Engineering Input.
: cycles, Quad Cities 1 Cycle 22 C D IRTODI NF1000236 Rev. 1 Page 28 of 29 Section D -
Item Planned Change Related Items Value References/Comments Criteria ECs 374977, 365821, 373667, 366920.
Planned station modifications that may Turbine Overhaul The effects of the turbine modifications are D.1 impact reload analysis Modifications included in Exelon TODl QOC 10-27 Rev. 0," OPL-W Parameters for Quad Cities Unit 1 Cycle 22 Transient Analysis", 9/15/14.
Planned setpoint changes that may Feedwater Temperature TODI-1 0-026 Rev.0, "Quad Cities Units I QDC D.2 impact reload analysis Operating Domain revision
"& 2 Nominal Feed Water Temperature Curve , 8/24110.D 3 Planned station minor modifications that None N/A mimpact reload analysis
.D'4 Planned component changes that may None N/A A impact reload analysis 102,1110 Quad Cities 1 Cycle 22 CDIRTO O1 NF1 0 239 Rev. 1 Page 29 of 29 CDIR Atta c hm ent 1 (ite m C.8)Q1C22 Exelon Fuel De sign CTQ's t .Attempt to reduce redundancies during the reload design effort such that both Exelon and W share in the preparation and review of output documents.
2.Design in compliance with CDIR specifications and meet Major Milestones in the Design & Licensing Schedule.
3.W performs multi-cycle analysis consistent with Exelon EUP following completion of Reference Loading Pattern. Exelon will utilize this information for economic/amortization purposes.
4.Open access to W core and bundle design procedures by Exelon. Mutual agreement on how to resolve conflicts, documenting exceptions when taken.
5.Improve design quality through collaboration.
6.Attempt to minimize the number of shadow corrosion induced channel bow susceptible bundles within a bladed cell. The objective is to minimize Exelon`s risk of channel-control blade interference utilizing Exelon and industry operating experience, 7.Adhere to SIL 320 (latest supplement) loading pattern restrictions to minimize fluence gradient induced bow impacts on channel to rod interference. All fuel types are to be treated as GNF fuel from a SIL 320 perspective.
8.Resulting design is acceptable when satisfactory results are obtained with both POLCA-7 and MICROBURN-82.
9.Expanded notch worth criteria and step worth criteria are met and margin to the limits is communicated.
lt(72/tO ATTACHMENT 8 EGG Calculation QDC-0000-N-1804 Rev. 0,"Quad Cities Unit 1 Cycle 22 Core Loading Plan" CC-AA-309-1 001 Revision 6 Page 1 of 9 ATTACHMENT 1 Design Analysis Major Revision Cover Sheet Page I of 5 Design Analysis (Major Revision)Last Page No. 9 B14 Analysis No.: 'QDC-0000-N-1 804Revision:0 Title: 3Quad Cities Unit 1 Cycle 22 Core Loading Plan EC/ECR No.:
4377652Revision:0 Station(s):Quad Cities Component(s): " Unit No.:1 N/A Discipline: aNUDC Descrip. Code
/Keyword: 'aN01 Safety/OA Class:SR System Code:N/A Structure:
'aN/A CONTROLLED DOCUM ENT REFERENCES Document No.:
From/To Document No.:
From/To OPTIMA2-TR038QC-EOP From QDC-0000-N-1806 To QDC-0000-N-1805 From QDC-0000-N-1653 From Is this Design Analysis Safeguards Information?
1eYes[]NooIf yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions?"Yes[]No ZIf yes, ATI/AR#:VIA This Design Analysis SUPERCEDES:
1eN/Ain its entirety.
Description of Revision (list changed pages when all pages of original analysis were not changed): 19 Original Issue Seth Spooner-^J/ 1 e l t.^ r !
Preparer s 20Brandon de Graafjy f fS hf Print NameSIName^''Date Method of Review:
_'Detailed Review Alternate Calculatl(attached) qTesting q Reviewer: 22Eric Bruch4 12 9 1 11 Print NameSign NameDate Review Notes: "Independent review 0Peer review Sep1 1 5.fzv^ITPR i^JhR`^r: mer^o n e-^/(Fm External Analyses Only)
External Approver:
uN/AN/AN/A Print NameSign NameDate Exelon Reviewer:
25N/AN/AN/A Print NameSINameDate Independent 3`d Party Review Reqd? mYesNo Exelon Approve r.17Jill Fisher,1oLIfri.__-_5f Print NameSin NameDate-f- Rtv;W o' b" A l e, cd blade-m aps Exe k ,n.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage 2 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN DESIGN ANALYSIS TABLE OF CONTENTS SECTION: PAGE NO.SUB-PAGE NO.CC-AA-309-1001 Major Revision Cover Page 1 N/A Design Analysis Table of Contents 2 N/A 1.0 Purpose 3 N/A 2.0 Inputs 3 N/A 3.0 Assumptions 4 N/A 4.0 References 4 N/A 5.0 Identification of Computer Programs 5 N/A 6.0 Method of Analysis 5 N/A 7.0 Numeric Analysis 6 N/A 8.0 Results 7 N/A 9.0 Conclusion 7 N/A Attachment A - Core Loading Plan Al-All N/A Attachment B - Independent Review Information B1 - B14 N/A xel !,n.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage 3 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN List of Tables Table 1 - Computer Codes Used In Analysis .............................................................................5 Table 2 - CAKWFT 1.4 Output .....................................................................................................6 1.0 Purpose The purpose of this design analysis is to document the Core Loading Plan Revision 0 for Q1 C22 per T&RM NF-AB-130-2200, Revision 1 (Reference 4.3).
The current revision of this Design Analysis (Revision 0) is part of the work covered under the Fuel Change Package FCP 377652, Quad Cities Unit 1 Cycle 22 Reload Design (Reference 4.5).2.0 Inputs 2.1 Core Loading by Assembly ID and Fuel Type The fuel assembly ID loading map is from the Westinghouse Reference Loading Pattern (Reference 4.6, pg Al -15, Al -16), modified with two changes: 1) replace bundle QAA21 1 in location 05-40 with bundle QAA1 39;and 2) replace bundle QAA213 in location 05
-22 with bundle QAA1 41. Both QAA1 39 and QAA1 41 were planned to be discharged in the Reference 4.6 loading pattern. These changes were made to address a suspected fuel failure detected in 01 C21 (see Reference 4.9 for loading changes).
The fuel type loading map is from the Q1C22 Design Analysis for the MICROBURN-132 basedeck (Reference 4.9).
This design analysis incorporates the replaced bundles, therefore no modifications were made to the map.
2.2 Core Weight The as-built core weight is determined using the Q1C22 Design Analysis for the MICROBURN-B2 basedeck (Reference 4.9).
Fresh fuel as-built bundle mass values were used as input to the deck, and the MICROBURN-B2 output file shows overall core weight as 124.83 MTU.
2.3 Control Blade Information This information is from the 01 C21 CLP (Reference 4.10) together with the changes described in Quad Cities blade replacement TODI (Reference 4.11), which shows details regarding planned blade replacements and shuffles during Q1 R21. Blade IQs arecompared to Reference 4.4 Attachment 11 to determine which of the Marathon control blades are part of etch indication population.
E x e kn.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage 4 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 2.4 EPG/SAG Information Parts of this information are directly from the Reference 4.6 report (# of Optima2 bundles on page `A1-14')
and Reference 4.7 Westinghouse report (Optimal active fuel lengths,total mass of U02, total mass of clad and channels on pages 9-12). There is also calculated information (Optima2 bundles' mass, total mass of Uranium), which is presented in Section 7.0 Numeric Analysis.
3.0 Assumptions None.4.0 References 4.1.Exelon Procedure, CC-AA-309, Revision 10, "Control of Design Analyses".
4.2.Exelon T&RM, CC-AA-309-1001, Revision 6, "Guidelines For Preparation And Processing of Design Analyses".
4.3.Exelon T&RM, NF-AB-130-2200, Revision 1, "Core Loading Plan Generation".4.4.Exelon T&RM, NF-AB-135-1410, Revision 8, "BWR Control Blade Lifetime Management".
4.5.Exelon Fuel Change Package 377652-000, "Quad Cities Unit 1 Cycle 22 Core Reload Design".4.6.Westinghouse Report NF-BEX-10-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22", dated 12/3/10. (Attachment 14 to FCP 377652-000)4.7.Exelon Design Analysis OPTiMA2-TRO38QC-EOP, Revision 1, "SVEA-96 OPTIMA2 Fuel Input to the Emergency Operating Procedures," dated 12/18/06. (QC record available via EDMS)4.8.Westinghouse Report NF-BEX-10-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22", dated 11/18/10. (Attachment 18 to FCP 377652-000) 4.9.Exelon Design Analysis QDC-0000-N-1805, Revision 1, "Quad Cities 1 Cycle 22 MICROBURN-B2 Basedeck," dated April 2011.
4.10.Exelon Design Analysis QDC-0000-N-1 653, Revision 1, "Quad Cities 1 Cycle 21 Core Loading Plan," dated 5/11/09.
4.11.TODI QDC-11-025, Revision 1, "Q1 R21 Control Rod Blade and LPRM Replacements," dated 4/21/11.
E xekrn.Nuclear Fuels - BWR DesignQDC-0000-N-1 804 Rev. 0 NuclearQuad Cities Design AnalysisPage 5 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 5.0 Identification of Computer Programs Table 1 - Computer Codes Used In Analysis Name Cksum IiP/UX 11.1 Ref DTSQA Info CAKWFT (1.4)
N/A for PC See DTSQA Database EX0006886 Level CC 6.0 Method of Analysis This Design Analysis was prepared and reviewed in accordance with the governing procedure CC-AA-309 (Reference 4.1) and the associated T&RM CC-AA-309-1001 (Reference 4.2).
The Core Loading Plan Generation T&RM was utilized to create the CLP. This procedure can be found in Reference 4.3.
E xeIn.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage 6 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 7.0 Numeric Analysis Core Average kW/ft This value is the rated core power (2957 MWt) divided by the summation of all the linear ft of fuel pellet stack in the core. The calculation includes part length correction as shown in the table below. The table also shows the linear ft calculation for each of the different bundle types loaded in the core by cycle.The Gad rod correction is not required for Optimal fuel. This correction is for fuel types in which the Gad rods have a shorter length than full length rods, which must be accounted for in the linear ft of fuel calculation (GE 14 fuel has this characteristic). This cycle will contain all Optimal fuel, which contains Gad rods that-hayd do not have aunique length.^sts^alu sqj SAM Information for input into CAKWFT 1.4 was obtained from Reference 4.6.
Table 2 - CAKWFT 1.4 Output Unit/C ycleQuad Cities nit J.
Cycle 22 Fuel TypeODt matOatesO -Number of Bundles236268220# Total rods per bundle969696# Full length rods848484#Gad rods#N/A#N/A#N/A#Short Part length rods444# Long Part length rods888 Full length rode, In.145.28145.28145.28 Gad rod length, In.145.28145.28145.28 Short Part length rods, in.50.3950.3950.39 Long Part length rods, in.99.6199.6199.61 Total fuel length, In., per bundle13201.9613201.9613201.96 Total fuel length, In. (by fuel type)3115662.56 3538125.282904431.2 Total fuel length, ft. (by fuel type) 259638.5467 294843
.7733 242035.9333 Total fuel length, ft. (for the core)796518.2533 Rated thermal power (MWth)
Core Average KW/ft Core Average kW/ft 2957 3.712 kW/ft Exe1n.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage 7 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN EOP Information Bundles mass = (Mass of U02 + Mass of Clad/channel)/# of assemblies For Optimal, using Reference 4.7, Table 1 values:
Optimal bundles mass = (316877 Ibm + 134824 Ibm)/724 = 623.9 Ibm Total mass of uranium is calculated per Reference 4.3.Total mass of U02 value is based on full core of Optimal fuel per Reference 4.7, Table 1:
Total mass of uranium
=total mass of U02 / ratio of the molecular weight of U02 to U 316877 Ibm / (270/238) 279321 Ibm Core weight is calculated per Reference 4.3.The core weight value is based on full core of Optimal fuel per Reference 4.7, Table 1:
Core weight = total mass of uranium in Ibm converted to MTU
Core weight = total mass of uranium in Ibm converted to MTU
= 279321 Ibm x (1 ST/2000 Ibm) x (1 MT/1.10231 ST)
                          = 279321 Ibm x (1 ST/2000 Ibm) x (1 MT/1.10231 ST)
= 126.70 MTU 8.0 Results Attachment A contains the Core Loading Plan for 01 C22.
                          = 126.70 MTU 8.0 Results Attachment A contains the Core Loading Plan for 01 C22.
9.0 Conclusion Using the inputs from Section 2 and the Numeric Calculations from Section 7, the Core Loading Plan document was created as Attachment A of this Design Analysis. The data presented in Attachment A meets the requirements of T&RM NF-AB-1 30-2200, Revision 1.
9.0 Conclusion Using the inputs from Section 2 and the Numeric Calculations from Section 7, the Core Loading Plan document was created as Attachment A of this Design Analysis. The data presented in Attachment A meets the requirements of T&RM NF-AB-1 30-2200, Revision 1.
E xeItn_Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage Al of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Attachment A Core Loading Plan E xeI^n-Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage A2 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN REVISION 0 This Design Analysis section contains the Core Loading Plan as specified by NF-AB-130-2200, Revision 1, "Core Loading Plan Generation".Revision Status: 0 Revision Notes: N/A CLP Distribution:Dave Schumacher (QC Reactor Engineering Manager)Mike Priaulx (QC Unit 1 Lead QNE)Jill Fisher (NF Manager BWR Design)Chuck Alguire (QC Manager Mechanical Design)Benone Lohan (Westinghouse)Charles Kuebel (QC Operations Staff)
E x e k ,n..Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage A3 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Table of Contents Item of Interest:Page Fuel Bundle InventoryA4 As-Built Core WeightA5 Core Average kW/ftA5 Fuel Assembly ID Loading MapA6 Fuel Type Loading MapA8 Control Blade InventoryA9 Control Blade Identification ArrayAl0 Emergency Operating Procedure (EOP) InformationAl 1 E xe1tn.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage A4 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Fuel Bundle Inventory Bundle No. of Fuel Bundle Description Cycle Initial Batch lD Range Type Bundles Loaded 25-29 56 Opt2-3.99-15GZ8.00-3G6.00 (QA20) 20 QAAOO1 -QAA056 Cycle 20, Non-control - Bundle Type # 25 Cycle 20,713 Locations - Bundle Type # 26 Cycle 20, 7C locations -
Bundle Type # 27 Cycle 20, 9C locations -
Bundle Type # 28 Cycle 20, 10A locations - Bundle Type # 29 30-36 114 Opt2-4.00-13GZ8.00-3G6.00 (QB20) 20 0AA057-0,AA192 Cycle 20, Non-control -Bundle Type # 30 Cycle 20, 7C locations -
Bundle Type # 31 Cycle 20, 8A locations -
Bundle Type # 32 Cycle 20, 8B locations -
Bundle Type # 33 Cycle 20, 9B locations - Bundle Type # 34 Cycle 20, 913 locations -
Bundle Type # 35 Cycle 20, 108 locations - Bundle T W e # 36 40-44 66 Opt2-4.05-12GZ7.00-2G6.00 (QC20) 20 QAA193-QAA260 Cycle 20, Non-control - Bundle Type # 40 Cycle 20, 7A locations -
Bundle Type # 41 Cycle 20, 8A locations - Bundle Type # 42 Cycle 20, 8B locations -
Bundle Type # 43 Cycle 20, 10C locations - Bundle Type # 44 4-10, 148 Opt2.3.98.18GZ8.00 (QD21) 21 QABOO1-0AB148 12-14 Cycle 21, Non-control - Bundle Type # 4 Cycle 21, 98 locations -
Bundle Type # 5 Cycle 21, 9C locations - Bundle Type # 6 Cycle 21, 9D locations - Bundle Type # 7 Cycle 21,10A locations -
Bundle Type # 8 Cycle 21,108 locations -
Bundle Type # 9 Cycle 21, 7A locations - Bundle Type # 10 Cycle 21, 78 locations -
Bundle Type # 12 Cycle 21, 7C locations - Bundle Type # 13 Cycle 21, 8A locations - Bundle Type # 14 xe1c'n-Nuclear Fuels- BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage A5 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 45-47 64 Opt2-3.99-16GZ8.00 (QE21) 21 QAB149-QAB212 Cycle 21, Non-control - Bundle Type # 45 Cycle 21, 7C locations - Bundle Type # 46 C y cle 21, 8B locations - Bundle T yp e # 47 21-24 56 Opt2-4.01-14GZ6.00 (0F21) 21 QAB213-GAB268 Cycle 21, Non
-control - Bundle Type # 21 Cycle 21, 91) locations - Bundle Type # 22 Cycle 21, 10C locations - Bundle Type # 23 C y cle 21 813 locations - Bundle T yp e # 24 60 104 0p12-4.07-19GZ7.50/5.50 (0022) 22 QAC001-QAC104 Cycle 22, Non
-control - Bundle Type # 60 80 56 Opt2-4.07-17GZ7.
50/5.50 (QH22) 22 QAC105-QAC160 Cycle 22, Non-control - Bundle Type # 80 90-92 60 Opt2-4.12-12G5.50-2GZ5.50 (0122) 22 QAC161-QAC220 Cycle 22, Non-control - Bundle Type # 90 Cycle 22, 9D locations - Bundle Type # 91 Cycle 22, 10C locations - Bundle Type # 92 As-Built Care Weight Quad Cities Unit I Cycle 22 124.83 MT1J' Core Average kW/tt Rated Thermal Power
=2957 MWth Quad Cities Unit 1 Cycle 22 3.712' kW/ft For Use with INPO Fuel Reliability Index Calculation Nuclear Fuels - BWR Design Attachment 1 to Q1C22 CLP Quad Cities Unit 1 Cycle 22 Care Loading Plan QDC-0000-N-1604 Revision 0 Page A6 of A11 Fuel Assembl y ID Loadin g Ma p 010305070911131517192123252729 OAA033 QAA248 CM177 QAAOSS QAA161 QM049 QAAOBI OAA201 OAA225 QAA025 QAA017 QAA169 OAA041 QAA075 QM153 OAA209 0M217 QA8001 Q46009 QM009 QAA003 0M233 QAA057 QAA129 QM113 QAA123 QA8253 QAC181 0A0133 QAC169 QM257 OAAOB7 OAA241 QAA193QA8125QABO37 QACI77 QAC185 QAB281 QACIOS 0AB061 QAAOOI OAA085 QA1097 0M185 QABIO1 QAC193 QAC201 QAC113 QABO29 OAC121 Omni QAC129 QM171 QAA235 QM243 QAA187 QMO99 QAB181 OAC209 QAC001 OAB189 OAB205 OAC009 QAC137 QABIO9 0AA043 0AA059 QM195 Q3.B103 QA81830AC217QAC017 QAB023 QABO77 QAB173 QAB053 QAC025 OAB237 QAA073 OM131 QAB12T QAC195 QAC211 QAC019 QABO17 QACA33 QAG145 OAC041 QAB213 QAC153 QA8045 QAA051 QAA155QAA115QABai9 QAC203 CA0003 QAB021 QACO35 QA8141 OAC049 QAB163 OAC057 QABO71 QACOBS QAA035 QAA083 QM139 0AA121 QAC119 QAC115 QA8191 OABO79 OAC147 QAC051OA8119QAa245QA8085 OAC073 QA8197 OAA251 OAA203 QAA219 QA8255 QAC157 QA8031 QAB297 OA8175 0AC043 QAB167OAB247OAB117 QACO81 0A8229 QA8093 QM179 QAA227 QAB003 QAC163 QAB283 QAC123 OAC011 QABO55 0AB215 OACO59 QAB087 QAC083 QAB1S7 OACO89 QA8149 QM091 OAA027 QA8011QA8135QAC107 0A8223 OAC139 QAC027 QAC155 QABOB9 QAC075 OA5231 QAC091 QABI59 QAC097 QM163 QM019 QMOII QAC171QA8083QAC131 QAB111 064,0239 QA6047 QAC067 QAB199 QA809S 0A8151 QAC099 QA8143 am165 QMB21 0M013 QAC173 QA8085 QAC133 OAB113 QAB241 OABO49 QAC069 GAB201 QASO97 QAB153 QAC101 QA8145 OM093 QM029 OABO13 QAB13T QACIO9 QA8225 QAC141 QAC02B QAC157 QA5075 QAC977 QA8233 QAC093 QABI81 QAC1O3 OAA151 QAA229 QAB005 QAC165 QAB265OAC125QAC013 QABO57 GAe217 QACO61 OABO89 CAM 85 QAB163QACOOSQAB155 OAA253 0M205 QAA221 QA8257 QAC189 QABO33 12AB209 OA8177 QACO45 QA8169 OA8249 QA8123 OAC087 OAB235 QA8099 OAA037 CAA095 QAA141 OM127 QAC181 QAC117 0A8193 QAB081 QAC149 QACO53QA81210A8251 QABO91 QAC079 QA8203 OAA053 QAA157 OAA117 W0&1 OAC205 QACOOS QAB027OAC037QAS147 OACOS5 OA8171 QAC063 QABO73 OAG071OAA079OAA133 OA8129 QAC197 QAC213 QACO21 OABO19 OAC039 QACI51 OAC047 OAS219 QAC1S9 OAS051 OAAO45 QAAO81 QAA197 0*8105 0A5185OAC219QAC023 QABO25 0A8083 0A8179 0A8059 QAC031 0A9243 OM173QAA237OAA245 QM189 OM101 0A8187 OAC215 QA0007 0A5195 QA8211 OAC015 QAC143 0A8115 00007 0AA071 QM103 QM191 CAB107 QAC199 QAC207 OAC119 0A5035 0A0127 QA8227 OAC135 QAA259 OAA099 OAA247 OM199 CA8131 0A8043 0AC183 QAC191 0A8267 QAC11I 0A8087QAA005QAA238QAA063QAA135 QM119 OAA125 QA8259 QAC167 0A8139 QAC175 4M175 OAA047 00077 QAA159 0A9215GAA2230A8007 0A5015 0AA015 QAA055 QAA087 OAA207 QAA231 OAA031 OAA023 0A9039 0AA255 QAA183 OAA095 0AA187 Nuclear Fuels - BWR Design Attachment 1 to 01 C22 CLIP 313335373941434547495153555759 Legends Center ASYID Exel 6 n_Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0Nuclear Quad Cities Design AnalysisPage A8 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Fuel Type Loading Map 1 1 3 5 7 9 11 131161 17 19 21 23 25 27 29 31 33 35 37 60 59 1528 42 30 35 34 34 35 30 42 5829 30 40 43 25 25 25 25 43 40 56 55 14130126130 30 40 44 13 9 27 27 9 13 44 64S 43 31 30 30 36 24 9010 92 92 10 90 24 52 51 1340 8 4 90 90 22 505 90 90 80 9 80 19ii30 47 1290 60 45 45 60 462691 60 4 4 45 12 4460 7 60 80 60 21 4 229 30 30 40 39 10 28 3090 80 FIKTTJ 60 602 14 3842 40 44 24 90 9 45 45 60 47 W4 6 38 35 9 30 43 13 90 22 84 60 12 21 60 3435 2510802180 60 60 14 31 8 34 2592 8F)28 27 7 35 25 9 1 10 1 80 121 80 46 6 0 14 60 21 12 60 80 22 90 13 43 30 26 23 602 47 60 45 45 9 90 24 44 40 42 2 60 45 45 60 14 21 4 60 80 4 45 80 90 3fi 40 30 28 22 14 60 60 14 60 47 60 4 60 4 60 80 6216080 60 7 6 80 60 90 4290 90 4 40 8 30 30 29 30 20 18 , 3 4r2of t j 21 80 601 80 14 560 14 21 4 60 80 4 45 80 90 3fi 40 30 28 4 4 80 80 24 24 601 1E 614604760 4 60 4 60 90 4 30 30 294 80 21 80 60 80 12 45 4 4 21 0 80 9 45 39 0 80 7 60 90 90 8 30 304 30 40 36 80 41143 1 45 47 49 51 53 55 57 59 31 43 1 25 4 8 90 90 60 5 30 3033 25 45 0 30 40 43 91 45 5 40 31 40 40 33 1 41 1 401 26130143 13 90 22 80 24 23 6 28 18 19 5 14 15 4 12 r17 0- 11 3 7 2 4 4 3 1 2 34 25 27 92 4 80 4 24 6 60 45 4 46 41 33 40 40 8 4 25 43 31 0 26 A 2 Iq 6 D E 18 30 30 0 30 29 140 44 2 30 30 26 43 40 30 30 45 90, 60 45 45 25 33 30 31 40 5 4560 4 4 45 12 30 30 36 4 90 6060 4 60 47 60 8 90 90 60 7 60 80 60.21 90 90 945 60 47 21 801 45 4 80 60 4 21 14 60 12 21 60 14 90 14 90 40 401 30 42 F 28 22 60 90 36 601 24 90 10 92 92 10 90 24 36 90 23 60 9 3H 26 13 9 27 27 9 13 43 25 25 25 25 43 80 22 30 35 34 34 35 46 60 60 60 21 80 80 21 0 80 4 4 80 60 80 60 46 60 601 46 60 0 23 4 24 60 2 45 4 444 80 60 45 45 46 60 46 60 4 4 4 1 60 46 60 60 0 60 60 34 46 4 45 60 6 24 4 60 4 922725 34 30-4 3 - 3 6 -
66 23 60 14 80 60 80 21 80 10 9 25 35 3 4 4 23 60 4 21 47 60 45 45 9 90 24 44 40 42 38 4fi 4 45 60 6 24 4 80 4 92 27 25 34 3 60 30 604228 iCLM 38 N 50 40 0 90 40 30 14 80 60 80 21 80 10 9 25 35 26 4 8 30 30 29 90 31 43 25 261 30 30 60 900 90 4 5 5 91 40 45 30 40 30 40 5 33 33 41 31 43 25 P 54 26 30 16 12 10 R 8 8 4 2 58 E xe l6 n-Nuclear Nuclear Fuels - SWR DesignQDC-0000-N-1804 Rev. 0 Quad Cities Design AnalysisPage A9 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Control Blade Inventory INITIAL BLADE IDENTIFIER RANGE PPLX BLADE TYPE NO.NO. OF BLADES CONTROL BLADE DESCRIPTION CX019R, CX062R, CY016R, D0075 1 4 GE D-100 (GE STD) 0077 to 0165, C0140 1 9 GE D-120 D104 to D691 2 15 GE D-230 M532 to M534, M545 to M551 3 10 GE Marathon with Etch Indication M1061 to M1432, M1537 to M1658, M1869 to M1878 3 42 GE Marathon AA050 to AA078 1 4 Westinghouse (ABB) CR-82 02-431 to 02-776, 02-1200 to 02-1351 1 93 Westinghouse (ABB) CR82M-1 E xe k n.Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0Nuclear Quad Cities Desi g n AnalysisPage Al 0 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Control Blade Identification Array TOOL No.ODC-11-025 Rev. I I Page 3 of Q1C22 Blade Map 2610141822263034384246505458 59 59 Maft1 14551 ABS02 02-60902-63 D2-606#8882 02-617 55 1RB882 1102.661 55 Mann 141330 ABB62 02-705 rii 18882 02-776 2 07-696 GE230 D579 A8582 AA0 76 GESTD D0075 A8802 02-462 U18Ba2 8882 02139 ABB82 882 AB882 ABB82 A8892 51 ABea2 Marie A8882 A8882 Martn 51 47 M533 02-704 02-455 02-695 02-607 02-463 02-768 02-1204 141328 02-1205 02-662 47 43 Mann M532 Mann GE230#8802 X48862 GE230 (#8882 GESTD Cx062R 43 Mam Mann 888 18882 2 IMartri M1657 060902-1242 02-457 ABM 02-431 39 35 31 27 23 39 35 31 27 23 48052 0E230 48882 Mam 882 Marts 06120 Marts 06120 Mann I 82 ann 8882 0882 ast 02-630 0304 02-693 141873 02-1203 M1427 0092 147062 0387 141072 02-1244 41876 2-772 2-452 348 GE230 48887.#0882 36230 Marro Marro Marro A8862 Mann 06120 Mann A8882 Mann 48882 GE37D 0499 02-440 D2-701 069014107014549 141073 02-1202 M1061 C0140 M1323 02450 M1068 2 4 49 YO16R A^BWB-8-2 Marts 82 Marls 06120 Maim A19B82 Mann#8882 Mann 06120 Mann A8882 Mann 82 02-632 141329 02-703 M1538 0165 14595 02 459 M546 02-775 M547 079 41658 2-702 1428 2-1339 2#8852 2 882 14ann A8882 862 62 Maim 2 Marta 682#8882 682 82 02-453 02-461 02-699 02-059 147326 02-1207 02-77302-1201M1655 112.1206 147430 02-45102-70002-664 3 2 06230 AW82 Marls 08982 Marts GE12D Mann GE120 Mann 2 MaM Afl882 GE230 2 02-633 D303 02-760 141870 02.1200 141653 0D79 141064 0003 141320 02-660 141878 02-7239 0202 02-616#5882 82-610 141822 AB382 02.769AB882 02.615GESTD CX019R GE230 D58146505458 19 ABB82 02-1351 15 GE230 Dt04 09882 02-1340[ABN2 02-446 Mann 141652 Marla#8682 02-657 Mann M534 GE120 0017 Mann M1071 141869 11 46882 GE230 0494 7 1.8 882 02.606 ASS82 78#8882 02-774GE230 0306 j0BB82 02-1240 02-443 2 10 iAB Mann M1066 IA6582 02-1241 02-1345 A8 Maim 141069 02-767 02-766 Mann 141429 802 02.663 06230 1 D691 46882 02-1219 26: 30 Maim M1327 AB802 02-634 Maim 14550GE230 0583 A8892 AAB50 Mann M1654 A8882 02-447 A8882 02-698 A8882 02-697 A8882 02-635 34GE230 D580 Mann M1063 882 02-770[AM 82 02.448[ASU B2 02.636 38 Mann M1656 AB882 02-12438882 02-456 r 82 02-1344 42 19 15 7 V/t 'Ile 2J6 s/^/ I w iE xel Vf_Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0NuclearQuad Cities Design AnalysisPage A11 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN EMERGENCY OPERATING PROCEDURE (EOP) INFORMATION Bundle T yp e Mass Ibm Westinghouse Optima2 Bundles 623.9 Parameter Value Number of Westinghouse Optimal Bundles 724 Optimal Active Fuel Length (in.)
145.28 Core Weight (MT) 126.70 Total Mass of Uranium (Ibm) 279,321 Total Mass of U02 (Ibm) 316,877 Total Mass of Clad and Channels (Ibm) 134,824 E xekn-Nuclear Fuels - BWR DesignQDC-0000-N-1804 Rev. 0 NuclearQuad Cities Design AnalysisPage B1 of B14 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Attachment B Independent Review Information C'aC.-0000-tJ -1$o P,ei 0 AHuc.ti^,+8 2. o 4 i e) k '
ATTACHMENT 2 Design Verification Guide - Core Loading Plan Page 1 of 5 1.DOCUMENT NUMBER QOC
-&deg;&deg;&deg;" N'ig&deg;y REV NO.0 (if applicable) ihLIST OF PROCEDURES AND T&RMs USED The following Is a list of Procedures and/or T&RMs that were used In the performance and verification of the analysis:NF-AB-130-2200 Revision 1 Page 5 of 16 NY-a PROCEDURE!
T&RM_REVISION NF-AB-130-2200_
I CC-AA-3a910 CC-AA-3W-icito t__--AIN-135-1106a *BTU Core Loading Pattern Generation Gan^rdl 01 tk s c r A nv ly SeS (yuiictincs^rtj'cc;on ar.A rbLxss.*+^O Z i's. c.,.5n A+.41 ys c^
3u)R co..i,-di Qi%aie., t.:L4t1;^w. f4L u i ..t-Note: P rocedures and T&RM not used should be lined out.
III.PROCEDURE I T&RM DEVIATIONS The following deviations and associated justifications from the Procedures
/T&RM listed in Section II were used for this analysis (mark N/A if no deviations were used):
N/A t)ew :t .t.Sior+S 40 i'ti+M p"..oeLs ,rVM Pst .n*o x^tecir o*+tfe prff' r ko.L L,.LA th grrca..'1 iht .^+w . i:s.on :4 bJF
-A6- "W-21c* dOCS n.4 4-* "- 41.4 rti.if% &(- +k
+{--Cr- tte tallow:
h..ry&#xa2; L 'mss
(-%.p.It^Lr lty^l:.;^^-^uJsGLlritd a.4 c. tits q s +a+.s1+JSs'S
'4444l k...Z 1L-h LaiaY.^k'0.17^t M1^2} ^dl eQf'a6G.,.,l..fm^:nn>>
t;NL.w'j,)'1Me [.,^.u^.gL*i Usa.^^ F.rt1 f bl+.
d^^n.^ra {:^Cv^ti4tCLr .2.Lu+s^ .*r l-t c$v: r^^d:.tt^Q e L_P ._.Sr QJ^ 1-15 .
L 44 rJ-.4* .. 4f n..ai .., :.. i ^.y C 4. ^
j*+s ne..^ra.r:s.o.n -M Nt=-A$ % iiH{0 d4[] w.
tU.rtlr 4 srt}U{!y++C i'*:S ,c u+^-tr+#- ijCt.Laier'C rrClcZQ VMo.31310 0 513/i f Qc - cxx O -
is - I E by Rw a a .tk^.t.. S ATTACHMENT I Design Verification Guide - Core Loading Plan Page 2 of 5 IV.CHECKLIST Core Designer Independent Item for Verification Reviewer Indicate V. N, or N/A and Initial SECTION A - Verification Checklist is the full core loading pattern Identical to the latest map sent to site n 0-Al Reactor Engineering for generation of the Core Component Transfer?th i ti Sh t A f&#xb6;'i or za on ee s u y A2 Are the fresh fuel types Identified by actual handle Identifications?
lbc.'t'Y Do the fuel types and quantities match those described In the N e^.+c ti:,s}.?Reference Loading Pattern document or subsequent revision?
G.,..*g*+D A4 Do the number of bundles add up to a full core size?B 6A D f Wii an in-core shuffle be performed during the refueling outage?
If no 1t g, D A5 (i.e., full core off-load), beginning-of
-cycle shutdown margin data may 1`-'.need to be incorporated Into the CLP.
'ef^A6 Do all pages contain the correct revision number of the report?
JD'l Al Is the control blade Inventory table correct?Y Is the control blade Identification array Identical to the latest map W lr^8n^t^N//'t-c3 A8 sent to site Reactor Engineering for generation of the Core Component Transfer Authorization Sheets?A9 is the EPG/SAG Information correct?
BMD xx, Y Ili All}Is the core average kW/ft calculation correct?
.)d M O Y U-3-Rl l^rovr^i k .T49 t"d ie Sw"^S MPt^4{^r^. '(p . W '
44 1wre'.NF-AB-130-2200 Revision 1 Page 6 of 16 R,Lv d (p(- -Dodo-V-a- ttc1NF AB %30-2200 Revision I Page 7 of 16 ATTACHMENT I Design Verification Guide -
Core Loading Plan Page 3 of 5 IV.CHECKLIST (Cont'd.)Item for Verification Core Designer Indicate Y, N, or N/
A and Initial SECTION B - Documentation Are all of the following included in the Core Loading Plan: "Full core" core loading pattern with actual handle identifications for flesh fuel; I Quantity, description, NFT number, handle ID range, and cycle loaded B2 for each unique bundle type in the core, as modeled by the reactor simulator code; 83 I Core weight In appropriate units?
SECTION C - Additional Questions B1 QDC-0000-N-1804 Revision 0 Attachment B B5 of B14 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 4 of 5 V.DESIGN VERIFICATION COMMENT SHEET Review ofQDC-0000-N-1804 Document In Review:ODC-0000-N-1804 Quad Cities Unit 1 Cycle 22 Core Loading Plan Responsible EngineerBrandon De Graaf/ Seth Spooner Independent Review Engineer. Eric Bruch
#Independent Review Comment Responsible Engineer Response Independent Review Closure Page 2, 9.0 Conclusions should read Page 2 changed to 'Conclusion'.
Concur.'Conclusion', or the title on page 7 SJS 04/22/11 EB 4/28/11 1 should be adjusted.
This only needs to be fixed if either page must be redone anyway.
E B 4120/11 Section 2.1, no mention is made for a Text added to mention suspected fuel Concur, reference of where the 2 fuel shuffles failure in 01 C21 and Reference 4.9.
EB 4/28/11 2 came from. Indicate the reference SJS 04/22/11 and that the rationale behind the moves is explained there.
ES 4/20/11 Section 2.2, the calculated core This is an input as it must be listed in No need to yell.
weight should not be included in this the results and is coming from an ES 4/28/11 3 section since it is for the 1nyuta only.external analysis.
No change made EB 4/20/11 to address this comment.
SJS 04122/11 Reference 4.7 is actually an Exelon While this is both an Exelon andda Concur.Design Analysis. Change the wording Westinghouse document (with the ES 4/28/11 4 to indicate that.
same name), it will be updated to ES 4/20111 refer to the Exelon reference for consistency.
SJS 04/22/11 Reference 4.8, add a 0 to make it Comment incorporated.
Concur.5 377652-000 to remain consistent.
SJS 04/22111 ES 4/28/11 EB 4/20/11 Reference 4.11 updated for Revision Concur 6 1 of Quad blade TODI.
.EH 4118/11 SJS 04/22/11 Page Al0 updated for Revision 1 of Concur 7 Quad blade TOOL
.E8 42111 SJS 04/22/11 8 9 10N F-AB+130.2200 Revision 2 Page 8 of 16 Page 1 of 1 NF-AB-130-2200 Revision 1 Page 9 of 16 ATTACHMENT I Design Verification Guide -Core Loading Plan Page 5 of 5 VII.SIGNATURES THE ACTIVITIES REVIEWED BY THIS DVG ARE COMPLETE, ACCURATE, AND ADEQUATE FOR APPLICATION TO: (Plant ! Unit andlor Cycle)
REVIEW: Erg c ay-UC At VERIFIER" APPROVAL: DATE SIGNATURE MANAGER"SIGNATUREDATE'THE PREPARER AND REVIEWER MUST BE DIFFERENT INDIVIDUALS Proiect Name Responsible Engineer-Rev, Page Quad Cities 1 Cycle 22 Brian Henning 2 - 4127/11 1 of 5 Purpose and Scope Ensure that all recommendations made by the Core Manager, Core Designer, Independent Reviewer, etc. for the identified core reload design are evaluated and resolved prior to the appropriate milestone (e.g. start up date).
1 Measurement Standard All recommendations resolved & documented.
Generic Analysis Checks Item Nu_mber Act i o Responsible Engineer Independent Reviewer 1 Is the responsible engineer qualified to perform the analysis?
8MC Y 615 2 Is the independent reviewer qualified to perform the analysis? S 63 T Project Name Responsible Engineer Rev.Page Quad Cities 1 Cycle 22 Brian Henning 2 - 4127/11 2 of 5 Generic Reload Items Resolution Completed Milestone Actio n Reference Responsible Independent Reviewer byiate Co d e Engineer Not to be All Work Design Analysis/Product page numbering and IR 521107 6c^,G closed.formattingareincompliancewiththe$i3 Applicable t ll requirements of CC-AA-309 and CC-AA-309-o a new items.1001 or other CC process. See IR 521107 for rI further information (internal OPEX).
Rod Ensure that the axial power peaking for hot Q1 C18 Transient WA SO -V Patterns, cases does not exceed 1.8 for Quad Cities and Selection/A Exp Acct Dresden due to a constraint on the LOCA analysis.CMR, Ensure that the core average void fraction for GE-NE-A2200103-N1A BSA Exp Acct Quad Cities units is less than 48.3%. If this is 56-01-QRev.0 y/'\it not satisfied, Tech Support should be notified.(TaskReport 0611)PPLX POWERPLEX Input Deck IR 662389 t3/A 9. b A typo or incorrect print of the LHGRFAC table heading in the "SVL" file generated by the POWERPLEX-III USEP05 version was identified.
When the "SHARE LIMITS" card is used in the POWERPLEX-III input, the heading"LHGRFAC FOR FUEL TYPE" in the "SVL" file should be "LHGRFAC FOR LIMIT TYPE" instead. This has no impact on calculations.
MYH Proeect Name Quad Cities 1 Cycle 22 Responsible Engineer Brian Henning Rev.2-4/27/11 Pane 3of5 Resolution Completed Milestone Action Reference Responsible Independent Reviewer by/D ate Code Engineer Keep until MB2/An error in AREVA document EMF-2480.
IR 679312 Nl A g,r k, the error PPLX"Maneuvering and Conditioning Criteria for is corrected ATRIUM-10 Fuel (REMACCX
-10)to Use with POWERPLEX-III," Revision 3 was identified.
in the The unit for the parameter "beta" in Equations EMF-(3.5) and (3.12) is incorrect. It should be in units 2480 of kW/ft, instead of MWd
/kgU. (MYH) document Deck, In EMF-2147,"MICROBURN
-B2 User's IR 637562 t l^fl^B.*D PPLX-Manual," Revisions 10 and 11 (Page 5-270), the IIl/M-B2 default value for EDGE
_DECQNDITIONING keyword is I. This is correct in the Default field, but is incorrect in the Description field. The fc3`6 t'31 current description states:
i f this keyword is not input, the deconditioning exposure increment will default to the nodal average." The correct description should read, if this keyword is not input, the deconditioning exposure increment will default to the rod nodal ex p osure."MYH)Powerplex AREVA Condition Report CR 2010
-6767 IR 01123765 Nf A 9-#-%Z>Decks identified a code error in MICROBURN-B2 and POWERPLEX-i11.The error occurs when the OVERRIDE subgroup is used to force specific


assemblies to use a different REMA ACCX type.The use of OVERRIDE caused the incorrect
ExeItn_          Nuclear Fuels - BWR Design    QDC-0000-N-1804 Rev. 0 Nuclear        Quad Cities Design Analysis            Page Al of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Attachment A Core Loading Plan
 
ExeI^n-                  Nuclear Fuels - BWR Design            QDC-0000-N-1804 Rev. 0 Nuclear            Quad Cities Design Analysis                    Page A2 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN REVISION 0 This Design Analysis section contains the Core Loading Plan as specified by NF-AB-130-2200, Revision 1, "Core Loading Plan Generation".
Revision Status: 0 Revision Notes: N/A CLP Distribution:
* Dave Schumacher (QC Reactor Engineering Manager)
* Mike Priaulx (QC Unit 1 Lead QNE)
* Jill Fisher (NF Manager BWR Design)
* Chuck Alguire (QC Manager Mechanical Design)
* Benone Lohan (Westinghouse)
* Charles Kuebel (QC Operations Staff)
 
Exek,n..              Nuclear Fuels - BWR Design    QDC-0000-N-1804 Rev. 0 Nuclear          Quad Cities Design Analysis            Page A3 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Table of Contents Item of Interest:                                          Page Fuel Bundle Inventory                                      A4 As-Built Core Weight                                      A5 Core Average kW/ft                                        A5 Fuel Assembly ID Loading Map                              A6 Fuel Type Loading Map                                      A8 Control Blade Inventory                                    A9 Control Blade Identification Array                        Al0 Emergency Operating Procedure (EOP) Information            Al 1
 
Exe1tn.                      Nuclear Fuels - BWR Design          QDC-0000-N-1804 Rev. 0 Nuclear                Quad Cities Design Analysis                    Page A4 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Fuel Bundle Inventory Bundle    No. of Fuel Bundle Description                          Cycle Initial Batch lD Range Type    Bundles                                                  Loaded 25-29      56    Opt2-3.99-15GZ8.00-3G6.00 (QA20)                    20    QAAOO1 -QAA056 Cycle 20, Non-control - Bundle Type # 25 Cycle 20,713 Locations - Bundle Type # 26 Cycle 20, 7C locations - Bundle Type # 27 Cycle 20, 9C locations - Bundle Type # 28 Cycle 20, 10A locations - Bundle Type # 29 30-36      114  Opt2-4.00-13GZ8.00-3G6.00 (QB20)                  20    0AA057-0,AA192 Cycle 20, Non-control - Bundle Type # 30 Cycle 20, 7C locations - Bundle Type # 31 Cycle 20, 8A locations - Bundle Type # 32 Cycle 20, 8B locations - Bundle Type # 33 Cycle 20, 9B locations - Bundle Type # 34 Cycle 20, 913 locations - Bundle Type # 35 Cycle 20, 108 locations - Bundle TWe # 36 40-44      66    Opt2-4.05-12GZ7.00-2G6.00 (QC20)                  20    QAA193-QAA260 Cycle 20, Non-control - Bundle Type # 40 Cycle 20, 7A locations - Bundle Type # 41 Cycle 20, 8A locations - Bundle Type # 42 Cycle 20, 8B locations - Bundle Type # 43 Cycle 20, 10C locations - Bundle Type # 44 4-10,    148  Opt2.3.98.18GZ8.00 (QD21)                          21    QABOO1-0AB148 12-14 Cycle 21, Non-control - Bundle Type # 4 Cycle 21, 98 locations - Bundle Type # 5 Cycle 21, 9C locations - Bundle Type # 6 Cycle 21, 9D locations - Bundle Type # 7 Cycle 21,10A locations - Bundle Type # 8 Cycle 21,108 locations - Bundle Type # 9 Cycle 21, 7A locations - Bundle Type # 10 Cycle 21, 78 locations - Bundle Type # 12 Cycle 21, 7C locations - Bundle Type # 13 Cycle 21, 8A locations - Bundle Type # 14
 
xe1c'n-                        Nuclear Fuels- BWR Design                      QDC-0000-N-1804 Rev. 0 Nuclear                  Quad Cities Design Analysis                            Page A5 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 45-47      64      Opt2-3.99-16GZ8.00 (QE21)                                      21  QAB149-QAB212 Cycle 21, Non-control - Bundle Type # 45 Cycle 21, 7C locations - Bundle Type # 46 Cycle 21, 8B locations - Bundle T ype # 47 21-24      56    Opt2-4.01-14GZ6.00 (0F21)                                      21  QAB213-GAB268 Cycle 21, Non-control - Bundle Type # 21 Cycle 21, 91) locations - Bundle Type # 22 Cycle 21, 10C locations - Bundle Type # 23 Cycle 21 813 locations - Bundle Type # 24 60      104    0p12-4.07-19GZ7.50/5.50 (0022)                                  22  QAC001 -QAC104 Cycle 22, Non-control - Bundle Type # 60 80        56    Opt2-4.07-17GZ7.50/5.50 (QH22)                                  22  QAC105-QAC160 Cycle 22, Non-control - Bundle Type # 80 90-92      60    Opt2-4.12-12G5.50-2GZ5.50 (0122)                                22  QAC161 -QAC220 Cycle 22, Non-control - Bundle Type # 90 Cycle 22, 9D locations - Bundle Type # 91 Cycle 22, 10C locations - Bundle Type # 92 As-Built Care Weight Quad Cities Unit I Cycle 22                124.83 MT1J' Core Average kW/tt Rated Thermal Power = 2957 MWth Quad Cities Unit 1 Cycle 22                3.712' kW/ft For Use with INPO Fuel Reliability Index Calculation


calculation of the preconditioned state for those specific assemblies.
Attachment 1 to Q1C22 CLP Nuclear Fuels - BWR Design Quad Cities Unit 1 Cycle 22 Care Loading Plan                                              Revision                Page QDC-0000-N-1604                                                                                0                  A6 of A11 Fuel Assembly ID Loading Ma p 01      03      05      07    09      11    13    15      17    19    21      23    25    27    29 OAA033  QAA248 CM177 QAAOSS QAA161 QM049 QAAOBI OAA201 OAA225 QAA025 QAA017 QAA169 OAA041 QAA075 QM153 OAA209 0M217 QA8001 Q46009 QM009 QAA003 0M233 QAA057 QAA129 QM113 QAA123 QA8253 QAC181 0A0133 QAC169 QM257 OAAOB7  OAA241 QAA193  QA8125 QABO37 QACI77 QAC185 QAB281 QACIOS 0AB061 QAAOOI OAA085 QA1097 0M185 QABIO1 QAC193 QAC201 QAC113 QABO29 OAC121 Omni QAC129 QM171 QAA235 QM243 QAA187 QMO99 QAB181 OAC209 QAC001 OAB189 OAB205 OAC009 QAC137 QABIO9 0AA043 0AA059 QM195 Q3.B103 QA8183 0AC217 QAC017 QAB023 QABO77 QAB173 QAB053 QAC025 OAB237 QAA073 OM131 QAB12T QAC195 QAC211 QAC019 QABO17 QACA33 QAG145 OAC041 QAB213 QAC153 QA8045 QAA051 QAA155 QAA115 QABai9 QAC203 CA0003 QAB021 QACO35 QA8141 OAC049 QAB163 OAC057 QABO71 QACOBS QAA035 QAA083 QM139 0AA121 QAC119 QAC115 QA8191 OABO79 OAC147 QAC051 OA8119 QAa245 QA8085 OAC073 QA8197 OAA251 OAA203 QAA219 QA8255 QAC157 QA8031 QAB297 OA8175 0AC043 QAB167 OAB247 OAB117 QACO81 0A8229 QA8093 QM179 QAA227 QAB003 QAC163 QAB283 QAC123 OAC011 QABO55 0AB215 OACO59 QAB087 QAC083 QAB1S7 OACO89 QA8149 QM091 OAA027 QA8011 QA8135 QAC107 0A8223 OAC139 QAC027 QAC155 QABOB9 QAC075 OA5231 QAC091 QABI59 QAC097 QM163 QM019 QMOII QAC171 QA8083 QAC131 QAB111        064,0239 QA6047 QAC067 QAB199 QA809S 0A8151 QAC099 QA8143 am165 QMB21 0M013 QAC173 QA8085 QAC133 OAB113 QAB241 OABO49 QAC069 GAB201 QASO97 QAB153 QAC101 QA8145 OM093 QM029 OABO13 QAB13T QACIO9 QA8225 QAC141 QAC02B QAC157 QA5075 QAC977 QA8233 QAC093 QABI81 QAC1O3 OAA151 QAA229 QAB005 QAC165 QAB265 OAC125 QAC013 QABO57 GAe217 QACO61 OABO89 CAM 85 QAB163 QACOOS QAB155 OAA253 0M205 QAA221 QA8257 QAC189 QABO33 12AB209 OA8177 QACO45 QA8169 OA8249 QA8123 OAC087 OAB235 QA8099 OAA037 CAA095 QAA141    OM127 QAC181 QAC117 0A8193 QAB081 QAC149 QACO53 QA8121 0A8251 QABO91 QAC079 QA8203 OAA053 QAA157 OAA117 W0&1 OAC205 QACOOS QAB027 OAC037 QAS147 OACOS5 OA8171 QAC063 QABO73 OAG071 OAA079 OAA133 OA8129 QAC197  QAC213 QACO21 OABO19 OAC039 QACI51 OAC047 OAS219 QAC1S9 OAS051 OAAO45 QAAO81  QAA197 0*8105 0A5185 OAC219 QAC023 QABO25 0A8083 0A8179 0A8059 QAC031 0A9243 OM173 QAA237 OAA245 QM189 OM101 0A8187 OAC215 QA0007 0A5195 QA8211 OAC015 QAC143 0A8115 00007 0AA071  QM103 QM191 CAB107 QAC199 QAC207 OAC119 0A5035 0A0127 QA8227 OAC135 QAA259 OAA099 OAA247 OM199 CA8131 0A8043 0AC183 QAC191 0A8267 QAC11I 0A8087 QAA005 QAA238 QAA063 QAA135 QM119 OAA125 QA8259 QAC167 0A8139 QAC175 4M175 OAA047 00077 QAA159 0A9215 GAA223 0A8007 0A5015 0AA015 QAA055 QAA087 OAA207 QAA231 OAA031 OAA023 0A9039 0AA255 QAA183 OAA095 0AA187
Do not use OVERRIDE feature of MICROBURN-82 and POWERPLEX
 
-Ill until the error has been corrected.
Attachment 1 to 01 C22 CLIP Nuclear Fuels - BWR Design 31 33 35 37 39 41  43    45    47  49    51 53    55    57      59  Legends Center ASYID
Project Name Responsible Engineer Rev.Page Quad Cities 1 Cycle 22 Brian Henning 2 - 4/27/11 4 of 5 Resolution Completed Milestone Action Reference Responsible Independent Reviewer b Mp_ate Code Engineer CBH, Ensure that the actual exposures at which each
 
/Acp Exp. Acct, Optimal bundle type is controlled in Q1C22 are Y Target within the bounds of the assumed bundle Rod controlled exposure ranges specified in the Patterns, appropriate Westinghouse CBH evaluation. If EOC Optimal bundles are controlled for significant
Exel 6 n_                                      Nuclear Fuels - BWR Design                            QDC-0000-N-1804 Rev. 0 Nuclear                                  Quad Cities Design Analysis                                      Page A8 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Fuel Type Loading Map 1 1 3 5 7 9 11 131161 17 19 21 23 25 27 29 31 33 35 37 39 41143 1 45 47 49 51 53 55 57 59 28 42 30 35 34 34 35 30 42 0 60 59 15 58                                              29 30 40 43 25 25 25 25 43 40 30 130126130 30 40 44 13 9 27 27 9 13 44 40 56 55 14 64                      43 31 30 30 36 24 90 10 92 92 10 90 24 36                                    31 43 1 25 40 8    4 90 90 22            4   4 80 0              4    8 40 40 33 1 411 52 51 13 50                                      5 90 90 80 9 80 21 80 80 21 80 9 80                        90 5 30 3033 25 19ii              30 S                    90 60 45 45 60 80                  80          45      90 45    0 30 40 43 47 12 46                26                  91 60 4    4 45 12 601 24 24 60 12 45 4              4 60 91 45 5 40 31 44                                      60 7 60 80 60 21 80                6 80 21        80        7 60 90 90 8 30 30                          4 42              29 30 30                                          14 601 1E 14 60 47 60 4 60 4 60 90 4 30 30 29                                  4 40 39 10 28 30              90 80 FIKTTJ 60 60          2  14        5      60 14 21      4 60 80 4 45 80 90 3fi 40 30 28 401 38          42 40 44 24 90 9 45 45 60 47                          23 4    4 23 60 4 21 47 60 45 45 9 90 24 44 40 42 38 38 35 9 30 43 13 90 22 84 60 12 21 60 W46 60 46 46 60                              6                      6                        -4 3 - 3 6 -
` j f Projection periods outside of these assumed exposure Iv ranges, then additional Westinghouse CBH evaluation may be required to support Q1C22 operation.
34          35 25        10 80 21 80 60 60 14                    46 60 60        60 23 60 14 80 60 80 21 80 10 9 25 35 3 34 25        92                                      60      4 60 4fi 4 45 60 6 24 4 80 4 92 27 25 34 3 31 8 34 25 27 92 4 80 4 24 6 60 45 4 46 601 4                      4 60 46 4 45 60 6 24 4 60 4 92 27 25 34 30 28 27 7 35 25      9 1 10 1 80 121 8F) 80      14 601 23 60 46 1 60 60            60        14 80 60 80 21 80 10 9 25 35 26 26        130143 13 90 22 80 60 12 21 60 14 60 46 60 46                      0 46 6 0 14 60 21 12 60 80 22 90 13 43 30 26 24            140 44 2      90 9      45 60 47 21        60 0              23 60        2  47 60 45 45 9 90 24 44 40 42 2 23 6 28 30 30 36 90 801 45 4 80 60 4 21 14 60 45 45 60 14 21 4 60 80 4 45 80 90 3fi 40 30 28 22 29 30 30 4 90 60          60 4 60 47 60 14 60 60 14 60 47 60 4 60 4 60 90                                        30 29          20 18 19 5 30 8 90 90 60 7 60 80 60.21 80 6 6 80 21 60 80 60 7 60 90 90 8 30 30                                                    18 26 31 40 5 45          60 4    4    45 12    60  4  24  60  2 45    4 4    60 91  45  5    4 40 31    26              16 14 15 4 43 40 30 30 45 90, 60 45 45 60 80 4              4 80 60 45 45              4 30 5            40 43 30 12                      25 33 30          90 90 60 9 80 21 80 80 21                      0 90 900 90 5 30          33 25                      12 0- 11 3 r17                          41 33 40 40 8 4 90 90 22 80 4                4 80      0 90 4        8 40 40 33 41                                10 7 2                        25 43 31          36 24 90 10 92 92 10 90 24 36 30 30 31 43 25                                                  8 0 26 30 30 401        13 9 27 27 9 13                40 30 30 261                                          8 44 3 1                                        0 30 40 43 25 25 25 25 43 40 30 29                                                                4 2                                                28 42 30 35 34 34 35 30 42 28                                                                  2 A      Iq              D      E        F        3        H              iC        L      M        N           P        R 2      6                      18      22      26              34      38      42                50          54        58 4r2of t j
In exposure accounting, document this compliance by including a detailed table that summarizes the controlled periods of operation for each bundle type and the assumed ranges.Consider updating this table after every se q uence exchan g e durin g the cycle, Target Ensure that Westinghouse performs ASI W/A at-l>
                                                                                                                              , 3
Rod Patterns, compliance checks as soon as possible for any significant changes to the target rod patterns for SKI'N!t EOC the remainder of Q1C22. If the rod patterns Projection change significantly during Q1C22, then additional Westinghouse ASI compliance evaluation may be required to support 01 C22 operation.
 
ATI COLR Include WCAP-16081-P-A Addendum 1 in Q1 NJ+^o 298639-COLR. Addendum supports low bundle mass S N 07 flux correlation limits required for Optimal l bundles on the periphery.
Exel6 n -                              Nuclear Fuels - SWR Design            QDC-0000-N-1804 Rev. 0 Nuclear                              Quad Cities Design Analysis                      Page A9 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Control Blade Inventory INITIAL BLADE IDENTIFIER RANGE          PPLX    NO. OF  CONTROL BLADE BLADE    BLADES  DESCRIPTION TYPE NO.
MB2 Deck Include Reviewer comments on ASSEMBLY QDC-0000-N-1805 f l/H ef- D F_TYPE cards, Lattices 4 and 25, XMASS card, Rev. 0.cy/FLOW_TEMP_F, and QAC bundle I NAME_RANGE card as described in the commens to QDC-0000-N-1805 Rev. 0.
CX019R, CX062R, CY016R, D0075            1        4    GE D-100 (GE STD) 0077 to 0165, C0140                1        9    GE D-120 D104 to D691                  2        15    GE D-230 M532 to M534, M545 to M551            3        10    GE Marathon with Etch Indication M1061 to M1432, M1537 to M1658,          3        42    GE Marathon M1869 to M1878 AA050 to AA078                  1       4    Westinghouse (ABB) CR-82 02-431 to 02-776, 02-1200 to 02-1351      1        93    Westinghouse (ABB) CR82M-1
Responsible Engineer 2-4/27/11 Project Name Quad Cities 1 Cycle 22 Brian Henning Rev.Page 5 of 5 W .0 Q n G Resolution completed Milestone Action Reference Responsible Independent Reviewer b y/D a t e C ode Engineer ATI M132 Deck The Thermal Limit set for QDC-0000-N-1805 QDC-0000-N-1805 piA.-, 985968-POW ERP Rev. 1 is based on Rev. 0 RLR. ATI has been Rev, 1-rL 4c# a.e>^af^c4 49 LEX Deck created to verify TL set or implement changes.
 
.;+.c TL data is preliminary until RLR Rev. 1 has been aa'--OJ--ti v approved.M82 Deck Update Steam Flow to FW temperature curve QDC-0000-N-1805 tt(A%" V POWERP and add comment card stating Lattice Types 4 Rev. 1.&#xa3;-c'i^ /`LEX Deck and 25 are not used In accordance with reviewer comments to QDC-0000-N-1805 Rev. 1.
Exekn.                                           Nuclear Fuels - BWR Design                                              QDC-0000-N-1804 Rev. 0 Nuclear                                        Quad Cities Desi gn Analysis                                                            Page Al 0 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Control Blade Identification Array TOOL No.           ODC-11-025 Rev. I                                                                         Page    3      of I
OVC0", ACV K:>l L CC-AA-309-1001 Revision 6 Page 40 of 73 ik 'to-u" Me.- 4 21 ATTACHMENT 4 Design Analysis Design Review Guidance Page 1 of 3 Purpose (Section 4.3.2.)Are the statement of problem and purpose of Design Analysis sufficiently detailed to reveal how the Design Analysis will address, resolve, and approach the stated problem? yes inputs (Section 4.3.3.)
Q1C22 Blade Map 2        6        10      14      18      22        26        30      34      38      42        46      50      54      58 ABS02    #8882      02-63  Maft1    1RB882                                                59 59                                                                          14551 02-609    02-617    D2-606          1102.661 A8582  GESTD  A8802    ABB62      Mann          2    8882    GE230  rii 18882                            55 55 AA0 76  D0075  02-462    02-705    141330  07-696    02139    D579    02-776 Marie    A8882  A8892  ABB82    ABB82      ABea2  A8882    882 Martn        U18Ba2 AB882                        51 51                                                                          02-768    02-1204 141328    02-1205 02-662 M533      02-704  02-455  02-695    02-607    02-463 Mann                                                                                                          GESTD            47 47                                                                                                                        Cx062R M532 Mann                    888 2 IMartri    (#8882  Mam    GE230    Mann        18882  Mann    ABM                X48862    GE230  #8802            43 43                                                                                                      02-457    0609    02-1242 02-431  M1657 0E230    48882    Mam        882 Marts      06120      Marts  06120    Mann        82      ann    8882    0882      ast  39 39 48052                                                                                                            2- 772  2-452    348 02-630  0304      02-693    141873  02-1203 M1427    0092      147062  0387    I 141072  02- 1244 41876 48887  . #0882    36230  Marro  Marro    Marro      A8862 Mann        06120  Mann      A8882    Mann    48882    GE37D    35 35 GE230                                                                                                02450    M1068    2449      YO16R 0499    02 -440  D2- 701  0690    141070  14549    141073    02-1202 M1061    C0140  M1323 2
1.Are design inputs into the Design Analysis clearly identified and their sources listed?
A^BWB Marts          82 Marls    06120  Maim      A19B82    Mann    #8882    Mann    06120    Mann      A8882  Mann          82  31 31                                                                                                      41658      2-702            2-1339 02-632  141329    02-703  M1538    0165    14595    02 459    M546    02- 775  M547      079                          1428 2  #8852            2    882  14ann  A8882        862        62 Maim            2 Marta          682  #8882      682      82 27                                                                                                                                          27 02-453  02-461    02-699  02-059  147326  02- 1207 02- 773    02-1201 M1655    112.1206 147430    02-451    02-700  02-664        3 2  06230    AW82    Marls    08982 Marts      GE12D      Mann    GE120    Mann          2  MaM      Afl882 GE230          2  23 23                                                                                                                                  02-616 02-633  D303    02- 760  141870  02.1200 141653    0D79      141064  0003    141320  02-660  141878    02-7239 0202 GE230    09882 [ABN2        Mann    ABB82 Mann          iAB      Mann    GE230    Mann    AB382      Maim    AB802 19                                                                                                                                          19 Dt04      02-1340 02-446  141652  02-1351 M1066      02-1241 M1654    D580    M1656    02.769    M1327  02-634
yes 2.Was the i nput information obtained from the correct revision of the source document?
                        #8682    Mann      Mann    GE120  Marla    IA6582 Maim        A8882    Mann    AB882 AB882        Maim    GE230            15 15                                                                                            02-1243 02.615      14550  0583 02-657  M534      M1071    0017    141869    02-1345 141069    02-447  M1063 1.8882  ASS82  GE230    46882    A8                  A8882      882  8882    GESTD A8892 11 02.606        78 0494    02-774    02-767    02-766  02-698  02-770  02-456  CX019R AAB50 j0BB82 #8882    GE230      802    Mann    A8882  [AM 82    r 82    GE230                                7 7
Y es 3.Are the design inputs relevant and directly applicable to the purpose of the Design Analysis?Yes 4.Are the inputs sufficient considering the purpose of the Design Analysis? Yes 5.Is the use of engineering judgment clearly documented and justified? Yee-, 6.Has all input data been used correctly? Yts 7.Do the sources of inputs used meet current technical requirements as committed under
02-1240 02-443  0306    02.663    141429  02-697  02.448  02-1344 D581
                                                            #5882    46882 06230        A8882  [ASUB2 82-610  02-1219 D6911      02-635  02.636 2                    10      14      18      22        26  :    30      34        38      42        46      50      54      58 V/t 'Ile 2J6 s/^/ I w
 
i Exel Vf_                Nuclear Fuels - BWR Design    QDC-0000-N-1804 Rev. 0 Nuclear            Quad Cities Design Analysis          Page A11 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN EMERGENCY OPERATING PROCEDURE (EOP) INFORMATION Bundle Type                        Mass Ibm Westinghouse Optima2 Bundles                                    623.9 Parameter                          Value Number of Westinghouse Optimal Bundles                            724 Optimal Active Fuel Length (in.)                                  145.28 Core Weight (MT)                                                  126.70 Total Mass of Uranium (Ibm)                                      279,321 Total Mass of U02 (Ibm)                                          316,877 Total Mass of Clad and Channels (Ibm)                            134,824
 
Exekn-         Nuclear Fuels - BWR Design      QDC-0000-N-1804 Rev. 0 Nuclear      Quad Cities Design Analysis             Page B1 of B14 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Attachment B Independent Review Information
 
aC.-0000-tJ -1$o                            P,ei 0 AHuc.ti              ^,+ 8 C'
: 2. o 4 i e) k '
NF-AB-130-2200 Revision 1 Page 5 of 16 ATTACHMENT 2 Design Verification Guide - Core Loading Plan Page 1 of 5                                                NY-
: 1.       DOCUMENT NUMBER QOC -&deg;&deg;&deg;" N'ig&deg;y                          REV NO.       0 (if applicable) a ih        LIST OF PROCEDURES AND T&RMs USED The following Is a list of Procedures and/or T&RMs that were used In the performance and verification of the analysis:
PROCEDURE!
T&RM _                REVISION          BTU NF-AB-130-2200_                  I          Loading Core        Pattern Generation CC-AA-3a9                      10              Gan^rdl 01 tk s cr          A nv ly SeS CC-AA - 3W-ici                    to            (yuiictincs          ^rtj'cc ;on ar.A            rbLxss. +^
OZ i's. c.,.5n A+.41 ys c^
t__--AIN-135-1106                a
* 3u)R co..i,-di Qi%aie., t.:L4t1;^w. f4            L u i ..t-Note: Procedures and T&RM not used should be lined out.
III.      PROCEDURE I T&RM DEVIATIONS The following deviations and associated justifications from the Procedures / T&RM listed in Section II were used for this analysis (mark N/A if no deviations were used):
N /A t)ew :t .t.Sior+S 40 i'ti+M        p".. oeLs ,rVM        Pst .n*o x^tecir o*+tfe prff'r ko.L L,.LA                th    grrca..'1 iht .^+w . i:s.on :4 bJF - A6- "W-21c* dOCS n.4 4-* "- 41.4 rti.if% &(- +k+{
.1                                                                                                              --  Cr- tte tallow:
h..ry&#xa2; L 'mss (-%.p        . It ^ Lr lty ^l:.;^^
^ibLitl^..
                                                                          ^uJsGLlritd a.4 c. tits q s +a+.s1+JSs'S
                          '4444 CA.? [.^ c. LItS**n k...Zl^ ,S* 1S 1L-h LaiaY.^k'0.17^t M1^2} ^dl eQf' a6G.,.,l..fm^:nn>>
t;NL.w    'j,)'1Me [.,^ .u^.gLi Usa.^^ F.rt1 f bl+. d^    ^n.^ra { :^Cv^ti4tCLr .2.L    u+s^ . r l-  t c$v:r^^d  :. tt^Q t%i.[1^            JL      e L_P L  ._.Sr c.
QJ^ 1-15 tti.+
                                                  .              1+ L t*n tis        44  rJ-.4* .. 4f  n..ai .., :.. i ^.y C 4. ^
j +s ne..^ ra.r:s.o.n -M Nt=-A$ % i              iH{0 d4[] w. tU. rtlr 4s          rt}U{!y ++C i' :S ,c u+^-tr+#- ijCt.Laie
    *r 1
r' C rrClc
                      .(  C1a  `*'^ ^sr^.ati8**
                                      +e          Grt bL. .(}liq ZQ VMo.31310                  0        513/i f
 
Qc - cxx O - is - I E by            Rw a a .tk^.t.. S NF-AB- 130-2200 Revision 1 Page 6 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 2 of 5 IV.      CHECKLIST Independent Core Designer Item for Verification                                                                                  Reviewer Indicate V. N, or N/A and Initial SECTION A - Verification Checklist is the full core loading pattern Identical to the latest map sent to site Reactor Engineering for generation of the Core Component Transfer                n 0-Al Auth or izati on Sh eets ?                                                  f          &#xb6;y'i A2  Are the fresh fuel types Identified by actual handle Identifications?              lbc.'t' Y
Do the fuel types and quantities match those described In the              N e^.+cti        :, s      }.?
Reference Loading Pattern document or subsequent revision?                      G.,.. g*+ D f
A4  Do the number of bundles add up to a full core size?                            B 6A D Wii an in-core shuffle be performed during the refueling outage? If no      1t g, D A5  (i.e., full core off-load), beginning-of-cycle shutdown margin data may      1  `-ef '.^ '
need to be incorporated Into the CLP.
A6  Do all pages contain the correct revision number of the report?              J              D      'l Al  Is the control blade Inventory table correct?                              Y Is the control blade Identification array Identical to the latest map      W lr^        8n^t^    N//'t- c3 A8  sent to site Reactor Engineering for generation of the Core Component Transfer Authorization Sheets?
xx, A9  is the EPG/SAG Information correct?                                                BMD                Y    Ili
                                                                                      .)
All} Is the core average kW/ft calculation correct?                                    dMO                Y U-3
            - Rl l      ^rovr            ^i k . T49    t"d ie Sw"^S          MPt^4    {^r^. '(p . W '    44 1wre'.
 
(p(- -Dodo-V-a- ttc1                          R,Lv d NF AB %30-2200 Revision I Page 7 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 3 of 5 IV.        CHECKLIST (Cont'd.)
Core Designer Item for Verification Indicate Y, N, or N/A and Initial SECTION B - Documentation Are all of the following included in the Core Loading Plan:
            "Full core" core loading pattern with actual handle identifications for B1 flesh fuel; I Quantity, description, NFT number, handle ID range, and cycle loaded B2 for each unique bundle type in the core, as modeled by the reactor simulator code; 83 I Core weight In appropriate units?
SECTION C - Additional Questions
 
QDC-0000-N-1804 Revision 0 Attachment B B5 of B14 N F-AB+130.2200 Revision 2 Page 8 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 4 of 5 V.      DESIGN VERIFICATION COMMENT SHEET Review ofQDC-0000-N-1804                                                                          Page 1 of 1 Document In Review:                ODC-0000-N-1804 Quad Cities Unit 1 Cycle 22 Core Loading Plan Responsible Engineer                Brandon De Graaf/ Seth Spooner Independent Review Engineer. Eric Bruch
          #  Independent Review Comment                                                        Independent Responsible Engineer Response Review Closure Page 2, 9.0 Conclusions should read        Page 2 changed to 'Conclusion'.        Concur.
              'Conclusion', or the title on page 7      SJS 04/22/11                            EB 4/28/11 1  should be adjusted. This only needs to be fixed if either page must be redone anyway.
E B 4120/11 Section 2.1, no mention is made for a      Text added to mention suspected fuel    Concur, reference of where the 2 fuel shuffles    failure in 01 C21 and Reference 4.9. EB 4/28/11 2  came from. Indicate the reference          SJS 04/22/11 and that the rationale behind the moves is explained there.
ES 4/20/11 Section 2.2, the calculated core          This is an input as it must be listed in No need to yell.
weight should not be included in this      the results and is coming from an        ES 4/28/11 3  section since it is for the 1nyuta only. external analysis. No change made EB 4/20/11                                to address this comment.
SJS 04122/11 Reference 4.7 is actually an Exelon      While this is both an Exelon andda      Concur.
Design Analysis. Change the wording Westinghouse document (with the                ES 4/28/11 4  to indicate that.                          same name), it will be updated to ES 4/20111                                refer to the Exelon reference for consistency.
SJS 04/22/11 Reference 4.8, add a 0 to make it          Comment incorporated.                  Concur.
5  377652-000 to remain consistent.            SJS 04/22111                            ES 4/28/11 EB 4/20/11 Reference 4.11 updated for Revision 6                                              1 of Quad blade TODI.                  Concur.
SJS 04/22/11                            EH 4118/11 Page Al0 updated for Revision 1 of Concur .
7                                            Quad blade TOOL SJS 04/22/11                            E8 42111 8
9 10
 
NF-AB-130-2200 Revision 1 Page 9 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 5 of 5 VII. SIGNATURES THE ACTIVITIES REVIEWED BY THIS DVG ARE COMPLETE, ACCURATE, AND ADEQUATE FOR APPLICATION TO:
(Plant ! Unit andlor Cycle)
REVIEW:
Erg c ay-UC At VERIFIER"                                  SIGNATURE                            DATE APPROVAL:
MANAGER"                                  SIGNATURE                            DATE
                        'THE PREPARER AND REVIEWER MUST BE DIFFERENT INDIVIDUALS
 
Proiect Name                            Responsible Engineer          -          Rev,                      Page Quad Cities 1 Cycle 22                            Brian Henning                  2 - 4127/11                    1 of 5 Purpose and Scope Ensure that all recommendations made by the Core Manager, Core Designer, Independent Reviewer, etc. for the identified core reload design evaluated and resolved prior to the appropriate milestone (e.g. start up date).                                                          are 1Measurement Standard All recommendations resolved & documented.
Generic Analysis Checks Item      Actio                                                                Responsible  Independent Nu
_mber                                                                          Engineer      Reviewer 1      Is the responsible engineer qualified to perform the analysis?          8MC Y  615 2      Is the independent reviewer qualified to perform the analysis?      y*v 63                                        T S
 
Project Name                          Responsible Engineer                      Rev.              Page Quad Cities 1 Cycle 22                          Brian Henning                    2 - 4127/11          2 of 5 Generic Reload Items Resolution Completed  Milestone  Action                                          Reference          Responsible      Independent Reviewer byiate Cod e                                                                          Engineer Not to be    All Work  Design Analysis/Product page numbering and          IR 521107          6c^,G closed.                formatting are in compliance with the                                  $i3 Applicable            requirements of CC-AA-309 and CC-AA-309-to all new 1001 or other CC process. See IR 521107 for                                                rI items.
further information (internal OPEX).
Rod    Ensure that the axial power peaking for hot      Q1 C18 Transient    WA SO -V Patterns, cases does not exceed 1.8 for Quad Cities and    Selection Exp Acct  Dresden due to a constraint on the LOCA                                                        /A analysis.
CMR,    Ensure that the core average void fraction for  GE-NE-A2200103-    N1A BSA Exp Acct  Quad Cities units is less than 48.3%. If this is 56-01-Q Rev. 0          y                      /'\
not satisfied, Tech Support should be notified.                                                    it (Task      Report 0611)
PPLX    POWERPLEX Input Deck                                IR 662389      t3/A 9. b A typo or incorrect print of the LHGRFAC table heading in the "SVL" file generated by the POWERPLEX-III USEP05 version was identified. When the "SHARE LIMITS" card is used in the POWERPLEX-III input, the heading "LHGRFAC FOR FUEL TYPE" in the "SVL" file should be "LHGRFAC FOR LIMIT TYPE" instead. This has no impact on calculations.
MYH
 
Proeect Name                          Responsible Engineer                      Rev.                  Pane Quad Cities 1 Cycle 22                            Brian Henning                  2-4/27/11                3of5 Resolution Completed  Milestone  Action                                              Reference      Responsible        Independent Reviewer by/Date                                                                                    Engineer Code Keep until    MB2/    An error in AREVA document EMF-2480.                    IR 679312        A g,r k, the error                                                                                  Nl PPLX    "Maneuvering and Conditioning Criteria for is                    ATRIUM-10 Fuel (REMACCX -10) to Use with corrected              POWERPLEX-III," Revision 3 was identified.
in the                The unit for the parameter "beta" in Equations EMF-                  (3.5) and (3.12) is incorrect. It should be in units 2480                  of kW/ft, instead of MWd/kgU. (MYH) document Deck,    In EMF-2147, "MICROBURN -B2 User's                      IR 637562  t l^fl^ B. D PPLX-    Manual," Revisions 10 and 11 (Page 5-270), the IIl/M-B2  default value for EDGE_ DECQNDITIONING keyword is I. This is correct in the Default field, but is incorrect in the Description field. The                                                  f  c3 t'
current description states: if this keyword is not                                              `    6 31 input, the deconditioning exposure increment will default to the nodal average." The correct description should read, if this keyword is not input, the deconditioning exposure increment will default to the rod nodal ex posure."    MYH)
Powerplex  AREVA Condition Report CR 2010 -6767                  IR 01123765 Decks    identified a code error in MICROBURN -B2 and                          Nf A 9- #-%Z>
POWERPLEX-i11. The error occurs when the OVERRIDE subgroup is used to force specific assemblies to use a different REMA    ACCX type.
The use of OVERRIDE caused the incorrect calculation of the preconditioned state for those specific assemblies.
Do not use OVERRIDE feature of MICROBURN-82 and POWERPLEX - Ill until the error has been corrected.
 
Project Name                          Responsible Engineer                      Rev.              Page Quad Cities 1 Cycle 22                          Brian Henning                    2 - 4/27/11            4 of 5 Resolution Completed  Milestone  Action                                            Reference        Responsible      Independent Reviewer bMp_ate    Code                                                                            Engineer CBH,    Ensure that the actual exposures at which each                        /A    cp Exp. Acct, Optimal bundle type is controlled in Q1C22 are                              Y Target  within the bounds of the assumed bundle Rod    controlled exposure ranges specified in the Patterns, appropriate Westinghouse CBH evaluation. If EOC    Optimal bundles are controlled for significant                                            `jf Projection periods outside of these assumed exposure                                                  Iv ranges, then additional Westinghouse CBH evaluation may be required to support Q1C22 operation.
In exposure accounting, document this compliance by including a detailed table that summarizes the controlled periods of operation for each bundle type and the assumed ranges.
Consider updating this table after every se q uence exchan g e durin g the cycle, Target  Ensure that Westinghouse performs ASI                                W/A at-l>
Rod    compliance checks as soon as possible for any SKI Patterns, significant changes to the target rod patterns for                                          'N!    t EOC    the remainder of Q1C22. If the rod patterns Projection change significantly during Q1C22, then additional Westinghouse ASI compliance evaluation may be required to support 01 C22 operation.
ATI          COLR    Include WCAP-16081-P-A Addendum 1 in Q1                                      o 298639-              COLR. Addendum supports low bundle mass NJ+^
S 07                    flux correlation limits required for Optimal                                                Nl bundles on the periphery.
MB2 Deck  Include Reviewer comments on ASSEMBLY              QDC-0000-N-1805 f l/H ef- D F_TYPE cards, Lattices 4 and 25, XMASS card,      Rev. 0.                cy                    /
FLOW_TEMP_F, and QAC bundle                                                                    I NAME_RANGE card as described in the commens to QDC-0000-N-1805 Rev. 0.
 
Project Name                          Responsible Engineer                      Rev.                          Page Quad Cities 1 Cycle 22                        Brian Henning                  2-4/27/11                          5 of 5            W .0 Q
n Resolution completed  Milestone  Action                                          Reference          Responsible                Independent Reviewer by/Date Code                                                                          Engineer ATI        M132 Deck  The Thermal Limit set for QDC-0000-N-1805        QDC-0000-N-1805  piA      .-,
985968-    POW ERP    Rev. 1 is based on Rev. 0 RLR. ATI has been      Rev, 1          -rL 4c# a.e>          ^af^c4 49          LEX Deck  created to verify TL set or implement changes.                    .;+.c c:.e .^rj+r*
TL data is preliminary until RLR Rev. 1 has been                  aa'-      -OJ--ti v approved.
M82 Deck  Update Steam Flow to FW temperature curve        QDC-0000-N-1805 tt(A %" V POWERP    and add comment card stating Lattice Types 4    Rev. 1.                &#xa3;-c' LEX Deck  and 25 are not used In accordance with reviewer                                                        i^ /`
comments to QDC-0000-N-1805 Rev. 1.
G
 
OVC                0", ACV    K:>  ik 'to -u" Me.- 4 21 lL CC-AA-309-1001 Revision 6 Page 40 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 1 of 3 Purpose (Section 4.3.2.)
Are the statement of problem and purpose of Design Analysis sufficiently detailed to reveal how the Design Analysis will address, resolve, and approach the stated problem? yes inputs (Section 4.3.3.)
: 1. Are design inputs into the Design Analysis clearly identified and their sources listed?
yes
: 2. Was the input information obtained from the correct revision of the source document?
Yes
: 3. Are the design inputs relevant and directly applicable to the purpose of the Design Analysis?                                                              Yes
: 4. Are the inputs sufficient considering the purpose of the Design Analysis? Yes
: 5. Is the use of engineering judgment clearly documented and justified? Yee-,
: 6. Has all input data been used correctly? Yts
: 7. Do the sources of inputs used meet current technical requirements as committed under the site license and regulatory commitments? Vcs
: 8. Are the Critical parameters designated for ease of verification? r/A Assumptions (Section 4.3.4.)
: 1. Have the assumptions necessary to perform the analysis been adequately documented? yes tom)
: 2. Is justification provided for all assumptions (except those based upon recognized engineering practice, physical constants or elementary scientific principals)? $(A
: 3. Are they reasonable for the Design Analysis? N to
: 4. Where necessary, are the assumptions identified for verification when the detailed design activities are completed. If so, have Action Tracking Items been established?
                                                                            ,IA References (Section 4.3.5.)
: 1. Are applicable codes, standards and regulatory requirements, including issues and addenda, employed in the Design Analysis properly identified and were their requirements met? ye S
: 2. Do the stated references reflect the appropriate revision? Y e-S
: 3. Are references that are not easily retrievable included as an attachment? N/N
: 4. Have supporting technical documents and references been reviewed when necessary?
4/21/11            1 es
 
Rc^ l^    A 14+c-c^^^+,
            'C,_ Oro - ^^ ^8d^1 CC-AA-309-1001 Revision 6 Page 41 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 2 of 3 Identification of Computer Programs (Section 4.3.6.)
: 1. Have the versions of the computer codes employed in the Design Analysis been certified for the application? Y<-5 Are              2.          the codes used identified along with source, computer type, inputs and outputs?
yes
: 3. Is the code suitable for present analysis? 7c-,
: 4. Does the computer model (modeling time steps, etc.) adequately represent the physical SySLWWWTi:l td(LU$U WIlUIUUIIS) (          t  es Is the output reasonable
: 5.                  when co iparedlto inputs? yes
: 6. Does the Computer Program conform with the NRC SER or similar document when applicable? z /A Method of Analysis (Section 4.3.7.)
Is the method    1.                      used appropriate considering the purpose and type of DesigAn alysis?
Is the method    2.                      in accordance with codes, standards, and regulatory requirements? if so are these codes, standards, and regulatory requirements referenced by number or title?
: 3. Has the method been employed elsewhere in industry or in license applications? Are these uses discussed? P 1A Are the methods4.used and recommendations given conservative relative to the design and safety limits?        Yes Numeric Analysis (Section 4.3.8.)
Are              1.        the equations used consistent with recognized engineering practice? YcS Is justification provided
: 2.            for any equations not in common use? Is the j ustification reasonable? N t A
: 3. Have the adjustment factors, uncertainties, and empirical correlations used in the analysis been correctly applied? yeS Is the result presented
: 4.          with proper units and tolerances? y S Are              5.          any of the results overly sensitive to small changes in input? Do tolerance and conclusions consider this sensitivity? h}O, 0 fA
: 6. What checking method was used to review the Design Analysis?
                    ^^-        Detailed Design Review.
                        --      Alternate Calculation Review.              'Reviews          3 Qualification Testing Review.


the site license and regulatory commitments? Vcs 8.Are the Critical parameters designated for ease of verification? r/A Assumptions (Section 4.3.4.)
1.Have the assumptions necessary to perform the analysis been adequately documented?
yes tom)2.Is justification provided for all assumptions (except those based upon recognized engineering practice, physical constants or elementary scientific principals)? $(A 3.Are they reasonable for the Design Analysis? N to 4.Where necessary, are the assumptions identified for verification when the detailed design activities are completed. If so, have Action Tracking Items been established?,IA References (Section 4.3.5.)1.Are applicable codes, standards and regulatory requirements, including issues and addenda, employed in the Design Analysis properly identified and were their requirements met?ye S 2.Do the stated references reflect the appropriate revision?Y e-S 3.Are references that are not easily retrievable included as an attachment? N/N 4.Have supporting technical documents and references been reviewed when necessary?
4/21/111 es
'C,_ Oro -
^^ ^8d^1 Rc^ l^A 14+c-c^^^+, CC-AA-309-1001 Revision 6 Page 41 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 2 of 3 Identification of Computer Programs (Section 4.3.6.)
1.Have the versions of the computer codes employed in the Design Analysis been certified for the application?
Y<-5 2. the codes used identified along with source, computer type, inputs and outputs?3.Is the code suitable for present analysis? 7c-, yes4.Does the computer model (modeling time steps, etc.) adequately represent the physical SySLWWWTi:l td(LU$U WIlUIUUIIS) (t es 5.
6.Does the Computer Program conform with the NRC SER or similar document when applicable?
z /A Method of Analysis (Section 4.3.7.)
1.
used appropriate considering the purpose and type of DesigAn alysis?2.
in accordance with codes, standards, and regulatory requirements? if so are these codes, standards, and regulatory requirements referenced by number or title?3.Has the method been employed elsewhere in industry or in license applications? Are these uses discussed?
P 1A 4.
and safety limits?Yes Numeric Analysis (Section 4.3.8.)
: 1. the equations used consistent with recognized engineering practice? YcS 2.
j ustification reasonable?
N t A 3.Have the adjustment factors, uncertainties, and empirical correlations used in the analysis been correctly applied?yeS 4.
: 5. any of the results overly sensitive to small changes in input? Do tolerance and conclusions consider this sensitivity?
h}O, 0 fA 6.What checking method was used to review the Design Analysis?
^^-Detailed Design Review.
--Alternate Calculation Review.'Reviews 3 Q ualification Testing Review.
Q tac - 0000 - - -
Q tac - 0000 - - -
13 H 0C B1 1i ATTACHMENT 4 Design Analysis Design Review Guidance Page 3 of 3 Results I Conclusion (Sections 4.3.9. and 4.3.10.)
13 H 0C B1 1i CC-AA- 309-1001 Revision 6 Page 42 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 3 of 3 Results I Conclusion (Sections 4.3.9. and 4.3.10.)
1.Is the magnitude of the results reasonable? Ye-5 2.Is the direction of trends reasonable?
: 1. Is the magnitude of the results reasonable? Ye-5
YfS 3.
: 2. Is the direction of trends reasonable?   YfS Are the conclusions
4.Have margin impacts been appropriately identified? a/A 5.Has an IR been initiated for any reduced margin that was identified during the preparation of the new/revised analysis? la /A 6.
: 3.     justifiable based on the results? Ye S
experience?
: 4. Have margin impacts been appropriately identified? a/A
y e.s 7.Has the effect on plant drawings, procedures, databases, and/or plant simulator been addressed?
: 5. Has an IR been initiated for any reduced margin that was identified during the preparation of the new/revised analysis? la /A Are the recommendations/results/conclusions
vi /A 8.Has the effect on other systems been addressed? i i / A 9.
: 6.                                         reasonable based on previous experience? y e.s
Specifications, M&TE selection sheets or procedures, COLR, etc.) been identified and tracked?g /A/Ei 10.When applicable, are the analysis results consistent with the proposed license amendment? k /11-11.Have other documents that have used the calculation as input been reviewed and
: 7. Has the effect on plant drawings, procedures, databases, and/or plant simulator been addressed? vi /A
: 8. Has the effect on other systems been addressed? i i / A Have any changes 9. in other controlled documents (e.g. UFSAR, Technical Specifications, M&TE selection sheets or procedures, COLR, etc.) been identified and tracked? g /A   /Ei
: 10. When applicable, are the analysis results consistent with the proposed license amendment? k /11-
: 11. Have other documents that have used the calculation as input been reviewed and revised as appropriate? to lPr General
: 1.                    Are applicable pages properly numbered and marked with the valid Design Analysis number? `ye-<,
Is all information legible and reproducible? kc_S
: 3. Have all cross-outs or overstrikes in the documentation been initialed and dated In ink?
I" Have all affected4.design analyses been documented on the Affected Documents List (ADL) for the associated Configuration Change?
FGP 3-+ % 652-levie                412f/ti


revised as appropriate? to lPr GeneralAre applicable pages properly numbered and marked with the valid Design Analysis number? `ye-<, Is all information legible and reproducible?
ATTACHMENT 9 Westinghouse Affidavit for Bundle Design Report and Reference Loading Pattern Report
kc_S3.Have all cross
 
-outs or overstrikes in the documentation been initialed and dated In ink?
Westinghouse                                                     Westinghouse Electric Company Nuclear Services 1 000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission                                Direct tel: (412) 374-4643 Document Control Desk                                            Direct fax: (724) 720-0754 11555 Rockville Pike                                                  e-mail: greshaja@ westinghouse.com Rockville, MD 20852                                              Proj letter: NF-BEX-11-142 CAW-11-3248 September 14, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
I" 4. (ADL) for the associated Configuration Change?
FGP 3-+ % 652-levie412f/tiCC-AA-309-1001 Revision 6 Page 42 of 73 ATTACHMENT 9 Westinghouse Affidavit for Bundle Design Report and Reference Loading Pattern Report Westinghouse Westinghouse Electric Company Nuclear Services 1 000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory CommissionDirect tel: (412) 374-4643 Document Control DeskDirect fax: (724) 720-0754 11555 Rockville Pikee-mail: greshaja@westinghouse.com Rockville, MD 20852Proj letter:
NF-BEX-11-142 CAW-11-3248 September 14, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE


==Subject:==
==Subject:==
Attachment to NF
Attachment to NF-BEX-10-162, Rev. 1 "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) and Attachment 1 to NF-BEX 184 "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary)
-BEX-10-162, Rev. 1 "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) and Attachment 1 to NF-BEX 184 "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary)
The proprietary information for which withholding is being requested in the above-referenced reports is further identified in Affidavit CAW-1 1-3248 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
The proprietary information for which withholding is being requested in the above
The subject documents were prepared and classified as Westinghouse Proprietary Class 2. Westinghouse requests that the documents be considered proprietary in their entirety. As such, non-proprietary versions will not be issued.
-referenced reports is further identified in Affidavit CAW-1 1-3248 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
The subject documents were prepared and classified as Westinghouse Proprietary Class 2.
Westinghouse requests that the documents be considered proprietary in their entirety. As such, non-proprietary versions will not be issued.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3248, and should be addressed to J.A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3248, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Very truly yours,J.A. Gresham, Manager Regulatory Compliance Enclosures CAW-11..3248 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
Very truly yours, J. A. Gresham, Manager Regulatory Compliance Enclosures
COUNTY OF BUTLER: Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
 
J.A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 141 day of September 2011 COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member.Pennsvivania Association of Notaries ss 2CAW-11-3248 (1)I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.
CAW-11..3248 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
(2)I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.
ss COUNTY OF BUTLER:
(3)I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:
(4)Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.(i)The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.(ii)The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
J. A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 141     day of September 2011 COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member. Pennsvivania Association of Notaries
Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
 
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a)The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of 3CAW-1 1-3248 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.(b)It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.(c)Its use 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 a similar product.(d)It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.(e)It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.(f)It contains patentable ideas, for which patent protection may be desirable.
2                                      CAW-11-3248 (1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.
There are sound policy reasons behind the Westinghouse system which include the following: (a)The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.(b)It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.(c)Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.
(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.
4CAW-11-3248 (d)Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.(e)Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.(f)The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.(iii)The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390; it is to be received in confidence by the Commission.(iv)The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.(v)The proprietary information sought to be withheld in this submittal is that which is contained in Attachment to NF
(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
-BEX-10-162, Rev. 1 "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) and Attachment 1 to NF
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
-BEX 184 "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary), for submittal to the Commission, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.
(i)     The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
This information is part of that which will enable Westinghouse to: (a)Support Exelon's use of Westinghouse Fuel at Quad Cities.
(ii)     The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
5CAW-11 3248 (b)Assist the customer to obtain license change.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
Further this information has substantial commercial value as follows: (a)Westinghouse can use this information to further enhance their licensing position with their competitors.(b)The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
(a)     The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for


licensing documentation without purchasing the right to use the information.
3                                      CAW-1 1-3248 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and
(b)    It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c)    Its use 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 a similar product.
(d)    It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)    It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)    It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a)    The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)    It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c)    Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.


the expenditure of a considerable sum of money.
4                                    CAW-11-3248 (d)    Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)    Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f)    The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390; it is to be received in confidence by the Commission.
(iv)  The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)  The proprietary information sought to be withheld in this submittal is that which is contained in Attachment to NF -BEX-10-162, Rev. 1 "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) and Attachment 1 to NF-BEX 184 "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary), for submittal to the Commission, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.
This information is part of that which will enable Westinghouse to:
(a)    Support Exelon's use of Westinghouse Fuel at Quad Cities.
 
5                                      CAW-11 3248 (b)    Assist the customer to obtain license change.
Further this information has substantial commercial value as follows:
(a)      Westinghouse can use this information to further enhance their licensing position with their competitors.
(b)      The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICE Transmitted herewith is the proprietary version of a document furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. The document is to be considered proprietary in its entirety.
PROPRIETARY INFORMATION NOTICE Transmitted herewith is the proprietary version of a document furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. The document is to be considered proprietary in its entirety.
COPYRIGHT NOTICE The report transmitted herewith bears a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in this report which is necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
COPYRIGHT NOTICE The report transmitted herewith bears a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in this report which is necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
ATTACHMENT 13 Westinghouse Affidavit for Quad Cities Unit 1 Cycle 22 SLMCPR Report Revision 2 Westinghouse Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory CommissionDirect tel:
 
(412) 374-4643 Document Control DeskDirect fax: (724) 720-0754 11555 Rockville Pikee-mail: greshaja@westinghouse.com Rockville, MD 20852Proj letter: NF-IBEX-11-3 R.ev. 2 CAW-1 1-3242 September 12, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE
ATTACHMENT 13 Westinghouse Affidavit for Quad Cities Unit 1 Cycle 22 SLMCPR Report Revision 2
 
Westinghouse                                                       Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission                                Direct tel: (412) 374-4643 Document Control Desk                                            Direct fax: (724) 720-0754 11555 Rockville Pike                                                e-mail: greshaja@westinghouse.com Rockville, MD 20852                                              Proj letter: NF-IBEX-11-3 R.ev. 2 CAW-1 1-3242 September 12, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE


==Subject:==
==Subject:==
NF-BEX-11-3 Rev.
NF-BEX-11-3 Rev. 2 P-Attachment, "Quad Cities Unit 1 Cycle 22 SLMCPR" (Proprietary)
2 P-Attachment, "Quad Cities Unit 1 Cycle 22 SLMCPR" (Proprietary)
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3242 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3242 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3242, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Very truly yours, J J. A. Gresham, Manager Regulatory Compliance Enclosures


on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.
CAW-11-3242 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.
ss COUNTY OF BUTLER:
Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3242, and should be addressed to J.A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.
Very truly yours, J J.A. Gresham, Manager Regulatory Compliance Enclosures CAW-11-3242 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:
COUNTY OF BUTLER:
Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief.
Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief.
J.A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me thisday of September 2011 COMMONWEALTH OF PENNSYLVANIA Notarial SealCynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member. Pennsvivania Assodation of Notaries ss 2CAW-11-3242 (1)I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.
J. A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this        day of September 2011 COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member. Pennsvivania Assodation of Notaries
(2)I am making this Affidavit in conformance with the provisions of 10 CFR Section 2,390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.
 
(3)I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
2                                      CAW-11-3242 (1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.
(4)Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.(i)The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.(ii)The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public.
(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2,390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.
Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows: (a)The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of 3CAW-11-3242 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.(b)It consists of supporting data, including test data, relative to a process (or
(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.
(i)     The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.
(ii)     The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.
Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:
(a)     The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of
 
3                                        CAW-11-3242 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.
(b)     It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.
(c)    Its use 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 a similar product.
(d)    It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.
(e)    It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.
(f)    It contains patentable ideas, for which patent protection may be desirable.
There are sound policy reasons behind the Westinghouse system which include the following:
(a)    The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.
(b)    It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.
(c)    Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense,
 
4                                    CAW- 11-3242 (d)    Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.
(e)    Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.
(f)    The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.
(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.
(iv)  The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.
(v)  The proprietary information sought to be withheld in this submittal is that which is appropriately marked in, NF-BEX-1 1-3 Rev. 2 P-Attachment, "Quad Cities Unit 1 Cycle 22 SLMCPR" (Proprietary) for review and approval, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.
This information is part of that which will enable Westinghouse to:
(a)      Support Exelon's use of Westinghouse Fuel at Quad Cities.


component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.(c)Its use 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 a similar product.(d)It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.(e)It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.(f)It contains patentable ideas, for which patent protection may be desirable.
5                                      CAW-11-3242 (b)     Assist the customer to obtain license change.
There are sound policy reasons behind the Westinghouse system which include the following: (a)The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.(b)It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.(c)Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense, 4CAW-11-3242 (d)Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.(e)Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.(f)The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.(iii)The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.(iv)The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.(v)The proprietary information sought to be withheld in this submittal is that which is appropriately marked in, NF-BEX-1 1-3 Rev. 2 P-Attachment, "Quad Cities Unit 1 Cycle 22 SLMCPR" (Proprietary) for review and approval, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.
Further this information has substantial commercial value as follows:
This information is part of that which will enable Westinghouse to: (a)Support Exelon's use of Westinghouse Fuel at Quad Cities.
(a)     Westinghouse can use this information to further enhance their licensing position with their competitors.
5CAW-11-3242 (b)Assist the customer to obtain license change.Further this information has substantial commercial value as follows: (a)Westinghouse can use this information to further enhance their licensing position with their competitors.(b)The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
(b)     The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.
Further the deponent sayeth not.
Further the deponent sayeth not.
PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.
PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.
In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).
In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).
COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding.
COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.
 
ATTACHMENT 14 Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Non-Proprietary)
ATTACHMENT 14 Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Non-Proprietary)
Westinghouse Non-Proprietary Class 3 Quad Cities Unit 1 Cycle 22 SLMCPR Westinghouse Electric Company Nuclear Fuel 1000 Westinghouse Drive Cranberry Township, PA 16066 C0 2011 Westinghouse Electric Company LLC, All Rights Reserved NF-BEX-11-3 Rev. 2 NP-Attachment 1.0Introduction This document contains a description of the Safety Limit Minimum Critical Power Ratio (SLMCPR) evaluation for Quad Cities Nuclear Power Station Unit 1 (QCNPS 1) Cycle 22.
 
Westinghouse Non-Proprietary Class 3 Quad Cities Unit 1 Cycle 22 SLMCPR Westinghouse Electric Company Nuclear Fuel 1000 Westinghouse Drive Cranberry Township, PA 16066 C0 2011 Westinghouse Electric Company LLC, All Rights Reserved NF-BEX-11-3 Rev. 2 NP-Attachment
 
1.0  Introduction This document contains a description of the Safety Limit Minimum Critical Power Ratio (SLMCPR) evaluation for Quad Cities Nuclear Power Station Unit 1 (QCNPS 1) Cycle 22.
Dual recirculation loop operation (DLO) and single recirculation loop operation (SLO)
Dual recirculation loop operation (DLO) and single recirculation loop operation (SLO)
SLMCPRs of 1.11 and 1.14, respectively, have been calculated for the Westinghouse SVEA-96 Optima2 assemblies in QCNPS 1 Cycle 22. Application of the Westinghouse methodology in Reference 1 requires modification of the QCNPSI Technical Specifications, Reference 3, to support DLO and SLO SLMCPRs of 1.11 and 1.14, respectively, for the SVEA-96 Optima2 fuel in Cycle 22. The SLMCPR values for QCNPSI Cycle 21 and Cycle 22 are shown in Table 1.For QCNPS 1 Cycle 22, Exelon Generation Company, LLC, (EGC) will load WestinghouseSVEA-96 Optima2 reload fuel for the third consecutive cycle.
SLMCPRs of 1.11 and 1.14, respectively, have been calculated for the Westinghouse SVEA-96 Optima2 assemblies in QCNPS 1 Cycle 22. Application of the Westinghouse methodology in Reference 1 requires modification of the QCNPSI Technical Specifications, Reference 3, to support DLO and SLO SLMCPRs of 1.11 and 1.14, respectively, for the SVEA-96 Optima2 fuel in Cycle 22. The SLMCPR values for QCNPSI Cycle 21 and Cycle 22 are shown in Table 1.
SVEA-96 Optima2 reload fuel was also loaded in QCNPSI Cycle 20 and Cycle
For QCNPS 1 Cycle 22, Exelon Generation Company, LLC, (EGC) will load Westinghouse SVEA-96 Optima2 reload fuel for the third consecutive cycle. SVEA-96 Optima2 reload fuel was also loaded in QCNPSI Cycle 20 and Cycle 21. In Cycle 22, all remaining legacy GE14 fuel shall be discharged, resulting in a full core loading of SVEA-96 Optimal assemblies.
: 21. In Cycle 22, all remaining legacy GE14 fuel shall be discharged, resulting in a full core loading of SVEA-96 Optimal assemblies.
Therefore, the Westinghouse NRC-approved methodology described in Reference 1, and further clarified in the response to Request for Additional Information (RAI) D13 of Reference 2, was used to determine the SLMCPRs for Cycle 22. Further clarification of the Westinghouse SLMCPR methodology was also provided to the NRC in support of the transition to SVEA-96 Optimal fuel in the Quad Cities and Dresden Units as follows:
Therefore, the Westinghouse NRC-approved methodology described in Reference 1, and further clarified in the response to Request for Additional Information (RAI) D13 of Reference 2, was used to determine the SLMCPRs for Cycle 22.
Further clarification of the Westinghouse SLMCPR methodology was also provided to the NRC in support of the transition to SVEA-96 Optimal fuel in the Quad Cities and Dresden Units as follows:
The response to NRC Request 19 in Reference 5 which supported the Licensing Amendment Request for transition to SVEA-96 Optima2 fuel in the Dresden and Quad Cities plants provided in Reference 4, The technical information supporting the Quad Cities Nuclear Power Station Unit 2 (QCNPS2) Technical Specification SLMCPR changes transmitted by Reference 6 as supplemented by the clarifying information in Reference 7.
The response to NRC Request 19 in Reference 5 which supported the Licensing Amendment Request for transition to SVEA-96 Optima2 fuel in the Dresden and Quad Cities plants provided in Reference 4, The technical information supporting the Quad Cities Nuclear Power Station Unit 2 (QCNPS2) Technical Specification SLMCPR changes transmitted by Reference 6 as supplemented by the clarifying information in Reference 7.
The same SLMCPR methodology described in these references was followed to establish appropriate SVEA-96 Optimal SLMCPRs for QCNPSI Cycle 22.
The same SLMCPR methodology described in these references was followed to establish appropriate SVEA-96 Optimal SLMCPRs for QCNPSI Cycle 22.
The EGC proposed license amendment to use the Westinghouse methodology for core reload evaluations at the Dresden and Quad Cities units was submitted to the NRC in Reference 4.
The EGC proposed license amendment to use the Westinghouse methodology for core reload evaluations at the Dresden and Quad Cities units was submitted to the NRC in Reference 4.
This submittal was approved by the NRC, and supported QCNPS2 Cycle 19, Dresden Nuclear Power Station Unit 3 (DNPS3) Cycle 20, QCNPS 1 Cycle 20, and Dresden Nuclear Power Station Unit 2 (DNPS2) Cycle 21, all of which are cores containing a reload of SVEA-96 Optima2 fuel, and QCNPS2 Cycle 20 and DNPS3 Cycle 21 which contain two reloads of SVEA-96 Optima2 fuel.
This submittal was approved by the NRC, and supported QCNPS2 Cycle 19, Dresden Nuclear Power Station Unit 3 (DNPS3) Cycle 20, QCNPS 1 Cycle 20, and Dresden Nuclear Power Station Unit 2 (DNPS2) Cycle 21, all of which are cores containing a reload of SVEA-96 Optima2 fuel, and QCNPS2 Cycle 20 and DNPS3 Cycle 21 which contain two reloads of SVEA-96 Optima2 fuel.
Condition 7 in the NRC safety evaluation for Reference I requires that the conservative factor be applied to the GE14 Operating Limit Minimum Critical Power Ratio (OLMCPR)and that this factor be identified in licensee applications. Since no legacy GE14 fuel assemblies are Page 2 of 17 NF-BEX-l 1-3 Rev. 2 NP-Attachment loaded into the QCNPS 1 Cycle 22 core, this factor is not relevant for this cycle and will not be included i n the Cycle 22 Core Operating Limits Report.
Condition 7 in the NRC safety evaluation for Reference I requires that the conservative factor be applied to the GE14 Operating Limit Minimum Critical Power Ratio (OLMCPR) and that this factor be identified in licensee applications. Since no legacy GE14 fuel assemblies are Page 2 of 17 NF-BEX-l 1-3 Rev. 2 NP-Attachment
I a,c 2.0SVEA-96 Optimal SLMCPR for Cycle 22 The SVEA-96 Optima2 SLMCPR for QCNPS1 Cycle 22 is based on a Reference Core design (SVEA-96 Optima2 bundle designs, core loading pattern and state point depletion strategy)


that represents realistic current plans for the Cycle 22 loading and operation. The Reference Core loading pattern for QCNPS 1 Cycle 22 is shown in Figure 1. The Reference Core design was generated via collaboration between EGC and Westinghouse based on EGC's cycle assumptions and design goals. The Reference Core was designed to meet the cycle energy requirements, to satisfy all licensing requirements, to provide adequate thermal margins and operational flexibility, and to meet other design and manufacturing criteria established by EGC and Westinghouse.
loaded into the QCNPS 1 Cycle 22 core, this factor is not relevant for this cycle and will not be included in the Cycle 22 Core Operating Limits Report.
a,c I
2.0  SVEA-96 Optimal SLMCPR for Cycle 22 The SVEA-96 Optima2 SLMCPR for QCNPS1 Cycle 22 is based on a Reference Core design (SVEA-96 Optima2 bundle designs, core loading pattern and state point depletion strategy) that represents realistic current plans for the Cycle 22 loading and operation. The Reference Core loading pattern for QCNPS 1 Cycle 22 is shown in Figure 1. The Reference Core design was generated via collaboration between EGC and Westinghouse based on EGC's cycle assumptions and design goals. The Reference Core was designed to meet the cycle energy requirements, to satisfy all licensing requirements, to provide adequate thermal margins and operational flexibility, and to meet other design and manufacturing criteria established by EGC and Westinghouse.
In general, the calculated SLMCPR is dominated by the flatness of the assembly CPR distribution across the core, and the flatness of the relative pin CPR distribution based on the pin-by-pin power/R-factor distribution in each bundle. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher SLMCPR.
In general, the calculated SLMCPR is dominated by the flatness of the assembly CPR distribution across the core, and the flatness of the relative pin CPR distribution based on the pin-by-pin power/R-factor distribution in each bundle. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher SLMCPR.
The calculation of the SLMCPR as a function of cycle exposure captures the interplay between the relative fuel assembly CPR and the bundle relative pin-by-pin CPR distributions established from the power/R-factor distributions and allows a determination of the maximum (limiting) SLMCPR for the entire cycle. This limiting SLMCPR is conservatively applied throughout the entire cycle, The SVEA-96 Optimal SLMCPR for QCNPS 1 Cycle 22 was determined as a function of cycle exposure based on radial assembly power distributions with about the same "flatness" as the Page 3 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment cycle exposure-dependent radial power distributions from [
The calculation of the SLMCPR as a function of cycle exposure captures the interplay between the relative fuel assembly CPR and the bundle relative pin-by-pin CPR distributions established from the power/R-factor distributions and allows a determination of the maximum (limiting) SLMCPR for the entire cycle. This limiting SLMCPR is conservatively applied throughout the entire cycle, The SVEA-96 Optimal SLMCPR for QCNPS 1 Cycle 22 was determined as a function of cycle exposure based on radial assembly power distributions with about the same "flatness" as the Page 3 of 17 NF-BEX-11-3 Rev. 2 NP- Attachment
 
cycle exposure-dependent radial power distributions from [
1 a,c Accordingly, the SVEA-96 Optimal SLMCPR for DLO was calculated at 100% power and 100% flow at various exposures throughout the cycle to assure that the limiting SLMCPR was identified. In addition, the DLO SLMCPRs were calculated at 100% power at the minimum allowed core flow at rated power (95.3% flow), as well as the maximum licensed core flow at rated power (108% flow) to confirm that a limiting SLMCPR had been established. Figure 3 shows a current QCNPS1 power-flow map which is applicable to Cycle 22. Consistent with Figure 3, a flow window of 95.3% to 108 % of rated core flow was analyzed.
1 a,c Accordingly, the SVEA-96 Optimal SLMCPR for DLO was calculated at 100% power and 100% flow at various exposures throughout the cycle to assure that the limiting SLMCPR was identified. In addition, the DLO SLMCPRs were calculated at 100% power at the minimum allowed core flow at rated power (95.3% flow), as well as the maximum licensed core flow at rated power (108% flow) to confirm that a limiting SLMCPR had been established. Figure 3 shows a current QCNPS1 power-flow map which is applicable to Cycle 22. Consistent with Figure 3, a flow window of 95.3% to 108 % of rated core flow was analyzed.
SLO SVEA-96 Optima2 SLMCPR calculations were also performed. These SLMCPR calculations were performed at [
SLO SVEA-96 Optima2 SLMCPR calculations were also performed. These SLMCPR calculations were performed at [
I a,c The SLO calculations used the same procedure as the DLO cases, except that the SLO cases applied a larger uncertainty for the core flow.The SLMCPR results for Cycle 22 are plotted in Figure 4. As shown in Figure 4, the DLO SLMCPR [I a,c Since the uncertainties at each DLO point are the same, this behavior is due to the interplay between the assembly relative CPRs and the relative fuel rod CPRs. In general, as the number of assemblies or fuel rods with CPRs in the vicinity of the assembly or fuel rods with the minimum CPR increases, the number of fuel rods with a potential for experiencing dryout increases.
a,c I    The SLO calculations used the same procedure as the DLO cases, except that the SLO cases applied a larger uncertainty for the core flow.
Therefore, a larger SLMCPR is required to assure that less than 0.1 % of the rods are in dryout.
The SLMCPR results for Cycle 22 are plotted in Figure 4. As shown in Figure 4, the DLO SLMCPR [
I a,c Since the uncertainties at each DLO point are the same, this behavior is due to the interplay between the assembly relative CPRs and the relative fuel rod CPRs. In general, as the number of assemblies or fuel rods with CPRs in the vicinity of the assembly or fuel rods with the minimum CPR increases, the number of fuel rods with a potential for experiencing dryout increases. Therefore, a larger SLMCPR is required to assure that less than 0.1 % of the rods are in dryout.
Experience has shown that the assembly CPR distributions tend to become [
Experience has shown that the assembly CPR distributions tend to become [
Ia,c Consequently, the peak SLMCPR tends to occur when the assembly CPR and rod CPR distributions combine to place the maximum number of fuel rod CPRs close to the minimum CPR.
Ia,c   Consequently, the peak SLMCPR tends to occur when the assembly CPR and rod CPR distributions combine to place the maximum number of fuel rod CPRs close to the minimum CPR.
This behavior is shown for the QCNPS 1 Cycle 22 SLMCPR by the relative assembly CPR histograms shown in Figures 5 through 7, respectively.
This behavior is shown for the QCNPS 1 Cycle 22 SLMCPR by the relative assembly CPR histograms shown in Figures 5 through 7, respectively.
Inspection of the DLO histograms in Figures 5 through 7 leads to the following observations, which explain the SLMCPR behavior in Figure 4:
Inspection of the DLO histograms in Figures 5 through 7 leads to the following observations, which explain the SLMCPR behavior in Figure 4:
NF-BEX-11-3 Rev. 2 NP-AttachmentI a,c Page 4 of 17 Ia,c Therefore, the DLO SLMCPR results at rated conditions in Figure 4 can be explained in terms of [1 a,c The adequacy of a DLO SLMCPR of 1.11 for the range of core flows at rated power permitted by Reference 3 was demonstrated by [I a,c Therefore, the results in Figure 4 confirm the adequacy of a DLO SLMCPR of 1.11 for a QCNPS1 Cycle 22 flow window at rated power conditions of 95.3% to 108 % flow.
a,c I
Page 4 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment
 
Ia,c Therefore, the DLO SLMCPR results at rated conditions in Figure 4 can be explained in terms of [
1 a,c The adequacy of a DLO SLMCPR of 1.11 for the range of core flows at rated power permitted by Reference 3 was demonstrated by [
I a,c Therefore, the results in Figure 4 confirm the adequacy of a DLO SLMCPR of 1.11 for a QCNPS1 Cycle 22 flow window at rated power conditions of 95.3% to 108 % flow.
The SLO results calculated at [
The SLO results calculated at [
SLMCPR of 1.14 for SLO.
I a,c Therefore, the results in Figure 4 confirm the adequacy of a SLMCPR of 1.14 for SLO.
I a,c Therefore, the results in Figure 4 confirm the adequacy of a The relative fuel rod CPRs in the SLMCPR calculations are [
The relative fuel rod CPRs in the SLMCPR calculations are [
I a,c Page 5 of 17 NF-BFX-11-3 Rev. 2 NP-Attachment la,c In addition to the strong dependence on assembly CPR and relative fuel rod CPR distributions, the SLMCPR is strongly dependent on the distribution of assembly and relative fuel pin CPRs about their mean values leading to an overall distribution of fuel rod CPRs relative to their mean values. The wider these distributions, the higher the SLMCPR must be to prevent 0.1%
a,c I
Page 5 of 17 NF-BFX-11-3 Rev. 2 NP-Attachment
 
la,c In addition to the strong dependence on assembly CPR and relative fuel rod CPR distributions, the SLMCPR is strongly dependent on the distribution of assembly and relative fuel pin CPRs about their mean values leading to an overall distribution of fuel rod CPRs relative to their mean values. The wider these distributions, the higher the SLMCPR must be to prevent 0.1%
of the fuel rods from experiencing boiling transition. The distributions of fuel rod CPRs relative to their mean values are determined by the uncertainties relative to the mean CPRs.
of the fuel rods from experiencing boiling transition. The distributions of fuel rod CPRs relative to their mean values are determined by the uncertainties relative to the mean CPRs.
Accordingly, the uncertainties used in establishing the SVEA-96 Optimal SLMCPR for Cycle 22 are shown in Table 2.
Accordingly, the uncertainties used in establishing the SVEA-96 Optimal SLMCPR for Cycle 22 are shown in Table 2.
3.0References 1.Licensing Topical Report, Reference Safety Report for Boiling Water Reactor Reload Fuel, CENPD-300-F-A, July 1996.
3.0    References
2.CENPD-3 89-P-A, I Ox] 0 SVEA Fuel Critical Power Experiments and CPR Correlations: SVEA-96+, August 1999.
: 1. Licensing Topical Report, Reference Safety Report for Boiling Water Reactor Reload Fuel, CENPD-300-F-A, July 1996.
3.Quad Cities Technical Specifications, Section 2.1.1.2 4.Letter, Patrick R. Simpson (Exelon Generation Company, LLC) to NRC, Request for License Amendment Regarding Transition to Westinghouse Fuel, dated June 15, 2005.
: 2. CENPD-3 89-P-A, I Ox] 0 SVEA Fuel Critical Power Experiments and CPR Correlations: SVEA-96+,
5.RS-06-009, Additional Information Supporting Request for License Amendment Regarding Transition to Westinghouse Fuel, January 26, 2006.
August 1999.
6.Letter from Patrick R. Simpson, Exelon Nuclear, to U.S. NRC,"Request for Technical Specifications Change for Minimum Critical Power Ratio Safety Limit ", QCNPS, Unit 2, December 15, 2005.
: 3. Quad Cities Technical Specifications, Section 2.1.1.2
7.RS-06-024,"Additional Information Supporting Request for Technical Specifications Change for Minimum Critical Power Ratio Safety Limit ", QCNPS, Unit 2, February 13, 2006.
: 4. Letter, Patrick R. Simpson (Exelon Generation Company, LLC) to NRC, Request for License Amendment Regarding Transition to Westinghouse Fuel, dated June 15, 2005.
Page 6 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment Table 1 Comparison of Cycle 21 and 22 Cores Descri p tion Cycle 21 Cycle 22 Number of Bundles in Core 724 724 Limiting Cycle Exposure Point Near EOC Near MOC Cycle Exposure at Limiting Point, EFPH 16809 EFPH 8357 EFPH Reload Fuel Type SVEA-96 O tima2 SVEA-96 Optima2 Reload Batch Average Weight % Enrichment 3.99 w/o 4.08 w/o Reload Batch Fraction (%)
: 5. RS-06-009, Additional Information Supporting Request for License Amendment Regarding Transition to Westinghouse Fuel, January 26, 2006.
37%30.4%Batch Fraction of SVEA-96 O tima2 Fuel 72.9%100.0%Batch Fraction of GNF GE14 Fuel 27.1%0.0%Core Average Weight % Enrichment 4.02 w/o 4.03 w/o Calculated Safety Limit MCPR (DLO) 1.11 for GE14, and SVEA-96 Optima2 1.11 SVEA-96 Optima2 Calculated Safety Limit MCPR (SLO)
: 6. Letter from Patrick R. Simpson, Exelon Nuclear, to U.S. NRC, "Request for Technical Specifications Change for Minimum Critical Power Ratio Safety Limit ", QCNPS, Unit 2, December 15, 2005.
: 1. 13 for GE 14, and SVEA-96 Optima2 1.14 SVEA-96 Optima2 Page 7 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment Table 2 - Uncertainties used in Quad Cities 1 Cycle 22 SVEA-96 Optimal SLMCPR Determination a,c Page 8 of 17 NF-BEX-1 1-3 Rev. 2 NP-Attachment 28 26 24 22 20 18 16 14 12 Figure 1 - Quad Cities 1 Cycle 22 - Reference Loading Pattern 0103050709111315171921232527 29 Legends QA20 0C20 QBZO 0B20 Q620 0620 Qcza 0020 0A20 0A20 0C20 0C20 0021 0021 0A20 0820 QF21 0122 0021 Q12201220122 0221 0H22 0021 QH22 0021 QH22 QF21 QH22 0821 QE21 QG22 QH22 0021 QD21 0821 0021 QG22 0F21 QH22 QG22 0221 QH22 0029 QG22 QE21 QG22 0021 QG22 0021 0221 0021 QG22 QE21 0221 0021 0022 QF2I 0021 0021 0022 0E21 QG22 0E21 QG22 0221 QG22 QE21 QG22 0E21 0021 0E21 QG22 QD21 0821 0021 0E21 QG22 0D21 QG22 0221 QG22 QE21 QG22 0021 0022 0821 QG22 0521 QF21 QD21 QG22 0221 QD21 0021 QF21 0021 0022 0E21 0322 0E21 0322 0021 QG22 QH22 QG22 QF21 QH22 002100210E21 0021 QG22 0221 0E21 0E21 QG22 QH22 QD21 01122 0021 01122 0221 01-122 QI22 QI22 QF21 QH22 0021 0820 0221 Q122 0021 0122 0C20 0020 0021 QD21 0A20 0820 0020 0C20 0A20 QA20 QA20 0020 0B20 0820 QB20 Center ASYTYP QA20 QB20 QA20 0620 QB20 QA20 QC20 QB20 QC20 QB20 QC20 QC20QA20QB20 QB20 QB20 Q021 QC20QB20QB20 QC20 QD21 QE21 011210122Q1220820QD21Q122QG22 0021QC20 08200122QH22 QE213021 0C20 QC20 0F21 Q122 QD21 0E21 QE21 QG22 QE21 01220F21QH22 0022 QD21 QD21QH22 QF21QH22 QG22 Q122Q021QH22 QD21QF21 QB20 0620 QA20 QD21 0122 QD21 QD21 QH22 QH22 QF21 Q021 QH22 QF21 QG22 QD21 QH22 QG22 QD21 Q122QF21QH22 QG22 Q021 0F210122Q021QE21QE21 QB200122QH22QE210021 0820QD21Q122QG22 QD21 QB20QD21Q12201220022 QB20 QC20 0021 QE21 0122 QC20 QC20QB20QB20 QE21 QA20 Q620QB20QB20 Q021 Q021 Q021QA20QC20QB20 QA20 QA20QB20QC20 QA20 QB20 QB20 Q620 Q620 QB20 QD21 Q122 QD21 Q122 Q122 QG22 QD21 QG22 QD21 QG22 QD21 0022 QH22 QG22 QB20 QB20 QB20 QB20 QE21 Q122 QG22 QD21 QD21 QA20 QF21 QH22 QD21 QG22 QD21 QG22 Q021 0C20 0C20 0820 QB20 QA20 QB20 QF21 QG22 QH22 QG22 QD21 QG22 Q122 Q122 Q022 QF-21 QG22 QD21 QG22 QD21 QG22 0122 NF-BEX-11-3 Rev. 2 NP-Attachment Page 9 of 17 Figure 1 - Quad Cities 1 Cycle 22 - Reference Loading Pattern 3133353739 0820 01320 0820 QC20 QA20 QA20 QA20 QC20 QC20 QB20 0A20 QD21 0D21 0C20 QC20 0122 0021 0122 QF21 QB20 0021 QH22 0F2101220122 QH22 0F21 QH22 0021 01122 0D21 QH22 QG22 QE21 QE21 GF21 QG22 QD21 0821 0021 0021 01122 0F21 QG22 QH22 QG22 0021 QG22 QE21 QG22 0E21 QG22 0021 0F21 0D21 QD21 0F21 QG22 QD21 0F21 QE21 QG22 QE21 QG22 QD21 QG22 0E21 RG22 QF21 QG22 QD21 QG22 0821 0021 QE21 0D21 0322 0E21 QD21 0821 QG22 QE21 QG22 QF21 QG22 0E21 QG22 QE21 QG22 0021 QD21 0F21 QG22 QD21 0F21 QE21 QG22 0D21 0F21 QD21 0322 0021 0022 QE21 QG22 0D21 QH22 QF21 QG22 QH22 0F21 0322 QD21 0821 0021 0021 01122 0322 QE21 0821 QH22 0F21 01-122 0021 QH22 0021 QH22 QF2101220122 0122 0021 0122 QF21 0820 RAID00210021 QC20 QC20 QA20 0A20 0C20 QC20 0620 0820 0820 0820 0C20 QA20 414345474951 53555759 Legends Center ASYTYP QA20 QB20 Q620 QB20 0A20 QB20 QC20 0A20 QD21 QB20 QB20 QC20 QC20 QC20 Q820 Q122 QD21 Q122 Q820 QA20 Q820 QB20 0122 QD21 QE21QC20QC20 QG22 QB20 Q920 Q122 QD21 QB20 QD21 QC20 QB20 QG22 QA20 QE21 Q122 QG22 QG22 0021 0122 QD21QB20 QB20 Q122 QG22 QG22 Q021 QD21 QA20 QD210B20 Q820 QE21 QH22 Q D21 Q122QB20 QG22 QH22 QC20 QB20 QA20 QE21 QG22 QE21QE21QD21 Q122QF21QC20QC20QC20 QF21 QF21Q122 QG22 Q021 Q021 QH22 QG22 Q021QG22QH22 QH22 QA20 QD21 QF21 QC20 Q620 Q620 QH22 QD21 0122 0022 QH22 QD21 QD21 0F21 QA20 0A20 0820 QF21 QG22 0021 OD21 Q021QI22 QH22 QD21 0A20 Q820 QF21 QH22 QG22 QH22 Q021 QD21 QH22 QD21 QB20 QA20 QA20 0F21 QG22 QH22 QF21Q122 QG22 QD21 QD21 QC20 QB20 QE21 QE21 0E21 Q122QF21 QG22 QD21 QE21 QG22 QH22 QD21QC20QC20 QB20 QC20 QA20 Q122QB20 QC20 QH22 0122 QG22 QG22 QD21 QD21 QA20 QD21QB20 Q820 Q122 QG22 QG22 QD21 Q122 QB20 0D21QB20 Q122 QD21 QD21 0C20 0820 Q022 QE21 GA20 0122 QG22 0E21 QB20 QC20 QC20 QB20 OB20 Q122 QD21 Q122 QB20 QB20 QA20 QB20 QD21 QC20 QD21 QC20 QB20 I QC20 0820 QB20 0820 0A20 QC20 0A20 QB20 QB20 0B20 QA20 60 58 56 32 30 28 26 24 22 NF-BEX-11-3 Rev. 2 NP-Attachment Page 10 of 17 Figure 2 - Quad Cities 1 Cycle 21 - Reference Loading Pattern 0103050709111315171921232527 2826 2825 2826 2825 2825 2825 2825 2825 2826 28252826QB20 0B20 0A20 QA200B200B20 0F21 QF21 QF21 QF21 QC20 0E21 QE21 QC200E210E21 0021 4820 QE21 0020 0021 0020 0E21 Q320 0021 0020 0021 0820 0021 0520 QD21 0B20 0021 QB20 0D21 0820 0D21 0820 QD21 Q620 0D21 0C20 0D21 0020 0D21 Q820 0021 0820 Q021 0C2000210820 0021 0A20 0021 0620 0021 0820 QD21 0620 Q021 0520 0021 0620 0820 0021 0A20 0021 0820 0021 0820 0021 0820 0021 0C20 0021 0620 0021 QA20 0021 0820 0021 0820 0021 0620 0021 0020 0021 0620 0021 0B20 0021 0820 QD21 QA20 0021 0620 0021 QB20 0021 0020 0029 0820 Q021 0020 0021 0020 0621 QB20 0621 062100210820 0621 0F21 0C20 0821 0021 QC20 0620 0820 QF210F210F21 2826 0820 0820 QA20 0A20 282528252825 2826 2825 2826 2825 282628252825 60 58 56 54 52 50 24 22 20 18 16 2826 2825 2826 2825 2825 2825 2825 2826 2825 2826 2825 2825 2826 2825 2826 2825 2825 2826 2825 2826 2825 2825 2825 2825 2826 2825 2826 QB20 4B20 QA20 QA20 0A20 QA20 QB20 QB20 2826 2825 2826 2825 2826 28252825 28252825 28262826 2825QA20 QA20 QF21 QB20 QF21 QB20 0C20 QF21QE21 QF21QE21 QF210C20 QF21QC20 QF21QE21 QF21QE21 QB20 0C20 QB20 0F21 QA20 0F21 2825QA20 28262826 2825 2825 2826 QB20 QA20 0F21 QE21 QE21 QE21 QD21 QB20 QE21 QE21 QB20 QD21 QE21 QE21 QE21 QF21 0A20 2825 2825 2825 QB20 2826 QF21 QC20 QE21 0C20 QD21 QC20 QE21 QB20 Q820 QE21 QC20 QD21 QC20 QE21 QC20 0F21 2825 1 28252825QB20 282528252826 28262825 2825 14 12 10 08 06 04 02 2825 2826 QB20 2825 2825 2826 2825 2826 2825 2825 QA20 QF21 0A20 QF21 QC20 QD21 QD21Q021QC20 QD21 QD21 QC20 QA20 0D21 QC20 QD21 0520 Q021 QB20 0021 QB20 Q021 Q520 Q021 QB20 QD21 QB20 QD21 QC20 QD21 QA20 Q021 QC20 QD21 QD21 QC20 0D21 QF21 QC20 QF21 QA20 2825 2826 QA20 2825 2825 2826 2825 2825 QA20 QF21 QE21 QE21 QG20 QD21 QB20 QD21 QB20 QD21 QB20 QD21 Q021 QB20 QD21 QB20 0021 QB20 QD21 QC20 QE21 QF21 QF21 QA20 2825 2825 Center ASYTYP 29 Legends NF-BEX- 11-3 Rev. 2 NP-Attachment Page 11 of 17 Figure 2 - Quad Cities 1 Cycle 21 -
: 7. RS-06-024, "Additional Information Supporting Request for Technical Specifications Change for Minimum Critical Power Ratio Safety Limit ", QCNPS, Unit 2, February 13, 2006.
Reference Loading Pattern 31 3335373941434547 2825 1 2825 1 282628252826 4951535557 59 Legends Center ASYTYP 600A200A200B20Q620 2826282528262825 2825 0F21QF210F21 Q820 0820 0A20 2825 2826 2825 0C20 0E21 0E21 0C200F21QF21 QA2028252825 0821 0520 00210E210E21 0E21 QF21 0A20 0620 QB20 0E21 QC20 0021 0C20 0E21 0C20 0F21 2826 0021 0B20 QD21 QC20 0021 0020 0021 0021 0F21 0520 QD21 0620 0021 QA20 QD21 QC20 0021 0C20 0021 0620 0021 0620 0021 0820 QD21 QC20 0E21 0820 0021 0C20QD21QB20QD21QA20 0021 0C20 QD21 QB20 QD21 0620 0021 0620 0021 0020 0021 0A20 0D21 0B20 0021 QC20 QD21 0520 0021 0C20 0021 0820 0021 0620 0021 0820 QD21 0820 0E21 0820 QD21 0A20 Q021 0820 QD21 0820 0021 Q620 Q820 0021 0A20 0021 0B20 QD21 062000210820 0021 0820 0021 QB20 0021 QB20 QD21 0620 0E21 QA20 0021 0620 QD21 QG20 QD21 0820 0021 0C20 0021 0B20 QD21 Q620 QD21 0520 QD21 0C20 0021 0820 0021 0C20 QD21 0820 QD21 0A20 0021 0C20 0021 0820 0021 0B20 0021 0820 QD21 0C20 QE21 0620 0021 0820 0021 0A20 0021 QC20 0021 0020 0021 0B20 0021 QC20 0021 0C20 0021 0021 QF21 0620 QE21 QC20 QD21 0C20 QE21 0C20 0F21 2826 0521 0520 0021 QE21 0621 QE21 QF21 0A20 0820 0C20 0621 QE21 QC20 QF21 0F21 0A20 2825 28250F210F21 QF21 0820 Q62D 0A20 282528262825 0A20 0A20 0820 082028262$25 282628252825 58 1 2825 1 2826 1 2825 2825 2825 2826 56 54 52 50 48 46 44 42 40 38 36 34 32 30 28 26 24 22 20 18 QC20QB20QB20 QC20 QF21QA20 QC20 QF21QA20 OC20QB20QB20 QA2028252826 QF21QA202825 QE21QF21QB20 QE21QF21QA20 QF2108202826 QF21QA202825 QA20 2826 2826 2825 2826 2825 QA20 QB20 QB20 2826 2825 2825 2825 2826 2825 2825 2825 2826 2825 2826 QF21 QE21 0F21 QE21 2626 QB20 0F21 2825 2825 2826 0020 QA20 QF21 0E21 QE21 QE21 QD21 QB20 0621 QE21 QB20 QD21 QE21 QE21 0E21 QF21 QA20 2825 2825 2826 2825 2825 2826 2825 2825 2826 2825 2825 2825 2826 2825 Page 12 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment Bundle Type Name Number Enrichment Cycle Loaded QA20 Opt2-3.99-15GZ8.00-3G6 00 56 3.99 20 QB20 Opt2-4.00-13GZ8.00-3G6.00 112 4.00 20 QC20 Opt2-4.05-12GZ7.00-2G6.00 68 4,05 20 QD21 Opt2-3.98-18GZ8.00 148 3.98 21 QE21 Opt2-3.99-16GZ8.00 64 3.99 21 QF21 Opt2-4.01-14GZ6.00 56 4.01 21 QG22 Opt2-4.07-19GZ7.50/5.50 104 4.07 22 QH22 Opt2-4.07-17GZ7.50/5.50 56 4.07 22 Q122 Opt2-4.12-12G5.50-2GZ5.50 60 4.12 22 Bundle Type Name Number Enrichment Cycle Loaded 2825 GE14-P1 ODNAB409-17GZ-1OOT-145-T6-2825 128 4.09 19 2826 GE14-PIODNAB408-15GZ-1GOT- 145-T6-2826 68 4,08 19 QA20 Opt2-3.99-15GZ8.00-3G6.00 56 3.99 20 QB20 Opt2-4.00-13GZ8.00-3G6.00 136 4.00 20 QC20 Opt2-4.05-12GZ7,00-2G6.00 68 4.05 20 QD21 Opt2-3.98-18GZ8.00 148 3.98 21 QE21 Opt2-3,99-16GZ8.00 64 3.99 21 QF21 Opt2-4.01-14GZ6.00 56 4.01 21 120 0003 ."PU Power- 2957 MW.,.
Page 6 of 17 NF-BEX-11-3 Rev. 2 NP- Attachment
100`% Core F10w- 98.0 Mi6/3r 1101 t 3200 E A:43.63 / 23.0 1 F 3:54.2 8 2 / 35.5 i F C: 1D0.0 6 1 / 95.3 4 F 0: 100.0100.0F 100.0 9 1 / 108.0 2 F 1 2800 F: 12400 27.0 z ? / 108.0 "_.F 18.8 +. ? / 36.6 04.9 : P / 88.5 8 c (7331 .?ower - 2511 tOW., IJrzq 11.002000 1600 T 400 10 Cavitation Interlock Line 0 60 50 40 30 20 0 708090100110120 010 Core Flow(%)
 
Page 13 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment Figure 4 Quad Cities 1 Cycle 22 SLMCPR Results for SVEA-96 Optima2 Fuel a,c Page 14 of 17 NF-BEX-l 1-3 Rev. 2 NP-Attachment Figure 5 - Assembly Histograms a,c NF-BEX-1 1-3 Rev. 2 NP-Attachment Page 15 of 17 Figure 6 -
Table 1 Comparison of Cycle 21 and 22 Cores Description                                                Cycle 21             Cycle 22 Number of Bundles in Core                                     724                   724 Limiting Cycle Exposure Point                           Near EOC               Near MOC Cycle Exposure at Limiting Point, EFPH                 16809 EFPH             8357 EFPH Reload Fuel Type                                     SVEA-96 O tima2         SVEA-96 Optima2 Reload Batch Average Weight % Enrichment                   3.99 w/o             4.08 w/o Reload Batch Fraction (%)                                   37%                 30.4%
Assembly Histograms a,c Page 16 of 17 NF-BBX-11-3 Rev. 2 NP-Attachment Figure 7 -
Batch Fraction of SVEA-96 O tima2 Fuel                     72.9%                 100.0%
Assembly Histograms a,c NF-BEX- l 1-3 Rev. 2 NP-Attachment Page 17 of 17}}
Batch Fraction of GNF GE14 Fuel                             27.1%                 0.0%
Core Average Weight % Enrichment                           4.02 w/o             4.03 w/o 1.11 for GE14,               1.11 Calculated Safety Limit MCPR (DLO) and SVEA-96 Optima2      SVEA-96 Optima2
: 1. 13 for GE 14,           1.14 Calculated Safety Limit MCPR (SLO) and SVEA-96 Optima2       SVEA-96 Optima2 Page 7 of 17 NF-BEX-11-3 Rev. 2 NP- Attachment
 
Table 2 - Uncertainties used in Quad Cities 1 Cycle 22 SVEA-96 Optimal SLMCPR Determination a,c Page 8 of 17 NF-BEX-1 1-3 Rev. 2 NP-Attachment
 
Figure 1 - Quad Cities 1 Cycle 22 - Reference Loading Pattern 01  03  05    07    09    11    13    15    17    19    21  23    25    27    29   Legends Center QA20 0C20 QBZO   0B20 Q620   ASYTYP QA20  0620 Qcza 0020   0A20 0A20 QB20  QA20  0620  QB20  0C20 0C20 0021   0021 0A20 QA20  QC20  QB20  Q620  QB20  0820 QF21 0122 0021 Q122 QC20  QB20  QC20  QC20  QD21  QD21  0122 0122 0221   0H22 0021 QA20  QB20  QB20  QB20  Q021  Q122  Q122  QH22 0021 QH22   QF21 QH22 QB20  QC20  QC20  QB20  QB20  QE21  Q122  QG22  0821 QE21 QG22   QH22 0021 QA20  QB20  QC20  QD21  QE21    Q122 QG22  QD21  QD21 0821 0021   QG22 0F21 QB20  Q620  01121  0122  Q122  QG22  QD21  QG22  QH22  QG22 0221   QH22 0029 0820  QD21  Q122 QG22  0021  0022  QD21  QG22 QE21 QG22   0021 QG22 QC20  0820  0122  QH22  QE21  3021  QH22  QG22  0021 0221 0021   QG22 QE21 0C20    QC20  0F21  Q122  QD21  0E21  QE21  QG22  QE21  0221 0021 0022   QF2I 0021 QD21  0122  0F21  QH22  0022  QD21  QF21  QG22  0021  0022 0E21   QG22 0E21 QD21  QD21  QH22  QF21  QH22  QG22  QH22  QD21  QG22 0221 QG22   QE21 QG22 QA20  Q122  Q021  QH22  QD21  QF21  QD21  QG22  0E21 0021 0E21   QG22 QD21 QB20    QA20  0122  QD21  QH22  Q021  QF21  QD21  QG22  0821 0021 0E21   QG22 0D21 28  0620    QD21  QD21  QH22  QF21  QH22  QG22  QH22  QD21  QG22 0221 QG22   QE21 QG22 26          Q021  Q122  QF21  QH22  QG22  Q021  QF21  Q022  0021 0022 0821   QG22 0521 24          0C20  0F21  0122  Q021  QE21  QE21  QG22  QF-21 QF21 QD21 QG22   0221 QD21 22          0C20  QB20  0122  QH22  QE21  0021 QH22  QG22  0021  QF21 0021  0022 0E21 20          0820  0820  QD21  Q122 QG22  QD21  QG22  QD21  0322 0E21 0322   0021 QG22 18          QB20  QB20  QD21  Q122  0122  0022  QD21  QG22  QH22 QG22 QF21   QH22 0021 16          QA20  QB20  QC20  0021  QE21    0122 QG22  QD21  0021  0E21 0021  QG22 0221 14          QB20  QC20  QC20  QB20  QB20  QE21  Q122  QG22  0E21 0E21 QG22   QH22 QD21 12                QA20  Q620  QB20  QB20  Q021  Q122  0122  01122 0021 01122 0221 01-122 Q021  Q021  QI22 QI22 QF21   QH22 0021 QA20  QC20  QB20  QB20  QB20  0820 0221 Q122 0021 0122 QA20  QB20  QB20  0C20 0020 0021   QD21 0A20 QA20  0820 0020 0C20   0A20 QA20 QA20 0020 0B20   0820 QB20 Page 9 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment
 
Figure 1 - Quad Cities 1 Cycle 22 - Reference Loading Pattern 31  33    35    37    39    41    43    45    47    49      51  53  55    57  59  Legends Center 60  0820 01320 0820  QC20 QA20                                                              ASYTYP 58  QA20 QA20  QC20  QC20  QB20 QA20 56  0A20 QD21 0D21  0C20  QC20  QB20  Q620  0A20  QB20 0122 0021  0122  QF21  QB20  QB20  QB20  QB20  QC20  0A20 0021  QH22  0F21  0122  0122  QD21 QD21   QC20  QC20  Q820    QC20 QH22 0F21  QH22   0021  01122 Q122  Q122  QD21 QB20  Q820    Q820 QA20 0D21  QH22 QG22   QE21  QE21  QG22 0122   QE21 QB20   QB20   QC20 QC20 Q920 GF21 QG22  QD21  0821  0021  QD21 QG22  Q122 QE21  QD21   QC20 QB20 QA20 0021  01122 0F21  QG22 QH22  QG22 0021  QG22   Q122  0122    QD21 QB20 QB20 QG22 0021  QG22  QE21  QG22 QD21 QG22   Q021 QG22   Q122   QD21 0B20 Q820  QA20 0E21 QG22 0021  0F21  0D21  QG22 QH22  Q D21 QE21  QH22    Q122 QB20 QC20  QB20 QA20 QD21 0F21  QG22  QD21  0F21  QE21  QG22  QE21  QE21  QD21    Q122 QF21 QC20 QC20 QC20 QE21 QG22  QE21  QG22  QD21 QG22  QF21  Q021  QG22  QH22    QF21 Q122 Q021  QC20 Q620 QG22  0E21  RG22  QF21 QG22  QD21  QH22   QG22  QH22   QF21    QH22 QD21 Q021  QA20 Q620 32 QD21  QG22 0821   0021 QE21  0022  QD21  0F21 QD21  QH22   QD21 0122 QA20  0A20 0820 30 0D21  0322  0E21  QD21  0821  QG22 0021   QF21  OD21  QH22    Q021 QI22 QA20  0A20 Q820 28 QG22 QE21 QG22   QF21  QG22 Q021  QH22  QG22  QH22  QF21    QH22 QD21 QD21  QA20 QB20 26  0E21 QG22 QE21   QG22 0021  QG22 0F21  QD21 QG22   QH22    QF21 Q122 QD21  QC20 QB20 24 QD21  0F21  QG22   QD21 0F21  QE21  QG22   QE21 0E21   QD21   Q122 QF21 QC20  QC20 QC20 22  QE21 QG22 0D21  0F21  QD21 QG22 QH22  QD21 QE21   QH22    Q122 QB20 QC20  QB20 QA20 0322 0021 0022   QE21 QG22 QD21  QG22   QD21 QG22  0122    QD21 QB20 Q820  QA20 0D21  QH22  QF21  QG22  QH22  QG22  QD21  QG22  Q122  Q122    0D21 QB20 QB20 0F21 0322  QD21  0821 0021 QD21  Q022  Q122  QE21  QD21    0C20 0820 GA20 0021 01122 0322   QE21 0821 QG22  0122  0E21  QB20  QB20    QC20 QC20 OB20 QH22  0F21  01-122 0021 QH22 Q122  Q122  QD21  QB20  QB20    QB20 QA20 0021 QH22 QF21  0122 0122 QD21  QD21  QC20 QC20   QB20 I QC20 0122 0021  0122  QF21  0820 0820 QB20  0820 QC20  0A20 RAID 0021  0021  QC20  QC20  QB20 QB20   0A20 0B20 QA20 0A20 0C20  QC20 0620  QA20 0820 0820  0820  0C20  QA20 Page 10 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment
 
Figure 2 - Quad Cities 1 Cycle 21 - Reference Loading Pattern 01  03  05    07    09    11    13    15    17    19    21    23    25    27  29  Legends Center 60                                                              2826  2825  2826  2825 2825 ASYTYP 58                                                        2825  2825  2825  2825  2826 2825 56                                      2825  2825  2826  2825  2826  QB20  0B20  0A20 QA20 54                                2825  2825  2826  2825  QA20  0B20  0B20  0F21  QF21 QF21 52                        2826  2825  2825  2825 QA20   QF21  QF21  QC20 0E21  QE21 QC20 50                  2825  2825  2826  QB20   0A20 QF21  QE21 0E21  0E21 0021  4820 QE21 2825  2825  2825  QB20   2826  QF21 QC20   QE21  0020  0021  0020  0E21 Q320 2825  2826  2826  QA20   QF21  QD21 QD21  QG20  0021  0020  0021  0820 0021 2826  2825  QA20  0F21  QC20  Q021 QC20  QD21 0520  QD21 0B20  0021 QB20 2825 2825  QA20 QF21  QE21  QE21  QC20 QD21   QB20 0D21  0820  0D21  0820 QD21 2826 2825 2826  QB20  QF21  QE21  0C20  QD21 QA20  QD21 Q620  0D21  0C20  0D21 0020 2825 2826 QB20  QB20  0C20   QE21 QD21  QC20 0D21  QB20 0D21  Q820  0021  0820 Q021 2826 2825 4B20  QF21  QE21  QD21   QC20  QD21 0520  QD21 0C20  0021  0820 0021 0A20 2825 2826 QA20  QF21  QE21  QB20 QE21  QB20 Q021  QB20  0021  0620  0021  0820 QD21 2825 2825 QA20   QF21  0C20  QE21  QB20  0021 QB20  QD21  0620  Q021  0520  0021 0620 2825 2825 0A20  QF21  QC20  QE21  Q820  Q021 Q520  Q021  0820  0021  0A20  0021 0820 2825 2826 QA20  QF21 QE21   QB20  QE21   QB20 Q021  QB20  0021 0820  0021 0820 0021 2826 2825 QB20  QF21  QE21  QD21  QC20  QD21 QB20   QD21 0C20 0021 0620 0021 QA20 24  2825 2826 QB20  QB20  0C20  QE21  QD21   QC20 QD21  QB20  0021 0820 0021 0820 0021 22  2826 2825 2826  QB20  0F21  QE21  QC20  QD21 QA20  0021 0620  0021 0020  0021 0620 20      2825 2825  QA20 0F21  QE21  QE21  QC20 Q021  QB20  0021 0B20  0021 0820 QD21 18          2826    2825 QA20  QF21  QC20  QD21 QC20  QD21  QA20 0021 0620 0021 QB20 16          2825    2826 2826  0A20  0F21  QD21 0D21  QC20  0021 0020 0029 0820 Q021 14          2825 1 2825  2825  QB20  2826  QF21 QC20  QE21  0020  0021  0020  0621 QB20 12                  2825  2825  2826  QB20  QA20  QF21  QF21  0621 0621 0021  0820 0621 10                        2826  2825  2825  2825  QA20  QF21  0F21 0C20 0821 0021 QC20 08                              2825  2825  2826  2825  QA20  0620 0820 QF21  0F21 0F21 06                                      2825  2825  2826  2825  2826 0820 0820 QA20 0A20 04                                                        2825   2825 2825 2825 2826 2825 02                                                              2826 2825 2826 2825 2825 Page 11 of 17 NF-BEX- 11-3 Rev. 2 NP-Attachment
 
Figure 2 - Quad Cities 1 Cycle 21 - Reference Loading Pattern 31    33      35  37    39    41    43    45    47    49    51  53    55    57  59  Legends Center 60    2825  1 2825 1 2826 2825 2826                                                           ASYTYP 58 1 2825 1 2826 1 2825  2825  2826  2825 56  0A20    0A20  0B20  Q620  2826  2825  2826  2825  2825 54  0F21    QF21   0F21  Q820  0820  0A20  2825  2826  2825  2825 52  0C20   0E21  0E21  0C20 0F21 QF21  QA20   2825  2825 2825 2826 50  0821    0520  0021  0E21  0E21  0E21  QF21   0A20   0620  2826  2825 2825 48    QB20   0E21  QC20 0021  0C20  0E21  0C20   0F21  2826 0020  2825 2825 46    0021    0B20  QD21 QC20 0021  0020  0021  0021  0F21 QA20  2826 2826 44    0520    QD21  0620  0021  QA20 QD21  QC20  0021  0C20 QF21   QA20 42    0021    0620  0021  0620  0021  0820  QD21   QC20   0E21 0E21  QF21 QA20 2825 40    0820    0021  0C20  QD21 QB20 QD21  QA20  0021   0C20 QE21  0F21 QB20 2626      2826 38  QD21   QB20   QD21 0620  0021  0620  0021  0020  0021 QE21  QC20 QB20  QB20 36    0A20    0D21  0B20  0021  QC20 QD21 0520  0021  0C20 QD21   QE21 QF21 QB20 34  0021    0820  0021  0620  0021  0820  QD21   0820  0E21 QB20   QE21 QF21  QA20 32    0820    QD21   0A20  Q021  0820  QD21 0820  0021  Q620 0621  QC20  QF21  QA20 30    Q820    0021  0A20  0021  0B20  QD21 0620  0021  0820 QE21  QC20  QF21  QA20 28    0021    0820  0021  QB20 0021 QB20 QD21   0620  0E21 QB20  QE21 QF21 QA20     2825 26    QA20    0021  0620  QD21  QG20  QD21  0820  0021  0C20 QD21  QE21 0F21  QB20      2826 24    0021    0B20  QD21  Q620  QD21  0520  QD21  0C20  0021 QE21  OC20  QB20  QB20 22    0820    0021  0C20  QD21  0820  QD21  0A20  0021  0C20  QE21  QF21 0820 2826 20    0021    0820  0021  0B20  0021  0820  QD21  0C20   QE21 0E21   QF21 QA20  2825 18  0620   0021  0820  0021 0A20  0021 QC20   0021 0020 QF21  QA20  2825  2826 0021   0B20  0021 QC20 0021 0C20 0021   0021   QF21 QA20  2826 0620   QE21  QC20  QD21 0C20  QE21  0C20   0F21  2826 QB20   2825 2825 0521    0520  0021 QE21  0621  QE21  QF21  0A20   0820  2826  2825 2825 0C20    0621  QE21  QC20 QF21  0F21  0A20    2825  2825  2825  2826 0F21    0F21  QF21  0820 Q62D  0A20  2825  2826  2825  2825 0A20    0A20  0820 0820 2826  2$25  2826  2825  2825 2825  2826    2825 2825  2825  2825 2825  2825    2826 2825  2826 Page 12 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment
 
Bundle                                                                                                                  Cycle Name                                    Number      Enrichment Type                                                                                                                  Loaded QA20                               Opt2-3.99-15GZ8.00-3G6 00                                  56        3.99            20 QB20                              Opt2-4.00-13GZ8.00-3G6.00                                  112        4.00            20 QC20                               Opt2-4.05-12GZ7.00-2G6.00                                  68        4,05            20 QD21                                     Opt2-3.98-18GZ8.00                                  148        3.98            21 QE21                                     Opt2-3.99-16GZ8.00                                  64        3.99            21 QF21                                     Opt2-4.01-14GZ6.00                                  56        4.01            21 QG22                                  Opt2-4.07-19GZ7.50/5.50                                  104        4.07            22 QH22                                  Opt2-4.07-17GZ7.50/5.50                                  56        4.07            22 Q122                              Opt2-4.12-12G5.50-2GZ5.50                                  60        4.12            22 Bundle                                                                                                                  Cycle Name                                    Number      Enrichment Type                                                                                                                  Loaded 2825                   GE14-P1 ODNAB409-17GZ-1OOT-145-T6-2825                                 128        4.09            19 2826                   GE14-PIODNAB408-15GZ-1GOT- 145-T6-2826                                  68        4,08            19 QA20                              Opt2-3.99-15GZ8.00-3G6.00                                   56         3.99           20 QB20                               Opt2-4.00-13GZ8.00-3G6.00                                   136        4.00           20 QC20                               Opt2-4.05-12GZ7,00-2G6.00                                   68         4.05           20 QD21                                     Opt2-3.98-18GZ8.00                                   148       3.98           21 QE21                                     Opt2-3,99-16GZ8.00                                   64         3.99           21 QF21                                     Opt2-4.01-14GZ6.00                                   56         4.01           21 120 0003 ."PU Power      - 2957 MW.,.
1101  100`% Core F10w      - 98.0 Mi6/3r                                                                        t 3200 A:    43.6  3 / 23.0    1 F                                                              E 3:    54.2 8 2 / 35.5    i F C:  1D0.0 6 1 / 95.3      4 F 2800 0:  100.0          100.0 100.0 9 1 / 108.0 F
2 F                                                                      1 F:    27.0 z ? / 108.0 "_. F 18.8 +. ? / 36.6                                                                            12400 04.9 : P / 88.5 8 c (7331 . Jrzq 11.00   ?ower - 2511  tOW., I                                                                    2000 1600 T 400 10                                                                    Cavitation Interlock Line 0                                                                                                                0 0          10          20           30      40  50      60      70          80            90 100      110 120 Core Flow(%)
Page 13 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment
 
Figure 4 Quad Cities 1 Cycle 22 SLMCPR Results for SVEA-96 Optima2 Fuel a,c Page 14 of 17 NF-BEX-l 1-3 Rev. 2 NP-Attachment
 
Figure 5 - Assembly Histograms              a,c Page 15 of 17 NF-BEX-1 1-3 Rev. 2 NP-Attachment
 
Figure 6 - Assembly Histograms a,c Page 16 of 17 NF-BBX-11-3 Rev. 2 NP-Attachment
 
Figure 7 - Assembly Histograms             a,c Page 17 of 17 NF-BEX- l 1-3 Rev. 2 NP-Attachment}}

Latest revision as of 23:57, 6 February 2020

Additional Information Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit
ML112650386
Person / Time
Site: Quad Cities Constellation icon.png
Issue date: 09/21/2011
From: Gullott D
Exelon Nuclear, Exelon Corp, Exelon Generation Co
To:
Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML112650398 List:
References
RS-11-155, TAC ME6383
Download: ML112650386 (192)
v • d • e


Text

Exelon Generation Company, LLC www.exeloncorp.com 4300 Winfield Road clear Warrenville, IL 60555 10 CFR 50.90 RS-11-155 September 21, 2011 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Unit 1 Renewed Facility Operating License No. DPR-29 NRC Docket No. 50-254

Subject:

Additional Information Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit

References:

1. Letter from Mr. Jeffrey L. Hansen (Exelon Generation Company, LLC) to U. S. NRC, "Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit," dated June 7, 2011
2. Letter from U. S. NRC to Mr. Michael J. Pacilio (Exelon Nuclear),

"Quad Cities Nuclear Power Station, Unit 1 - Request for Additional Information Related to Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAG No. ME6383)," dated August 22, 2011 In Reference 1, Exelon Generation Company, LLC (EGC) requested an amendment to Appendix A, Technical Specifications (TS), of Renewed Facility Operating License No.

DPR-29 for Quad Cities Nuclear Power Station (QCNPS), Unit 1. The proposed change revises the value of the single recirculation loop operation (SLO) safety limit minimum critical power ratio (SLMCPR) in TS Section 2.1.1, "Reactor Core SLs." Specifically, the proposed change would replace the current SLO SLMCPR requirement for QCNPS Unit 1 with a new SLMCPR requirement. This proposed change does not affect the QCNPS Unit 1 two recirculation loop operation (TLO) SLMCPR or either of the SLMCPR values for Unit 2. This change is needed to support the current cycle of operation (i.e., Cycle

22) for QCNPS Unit 1 for cycle exposure greater than 4000 MWd/MT, which is currently scheduled to occur in November 2011.

In Reference 2, the NRC requested that EGC provide additional information in support of their review of Reference 1. The NRC request for additional information (RAI) and the specific EGC responses are provided in Attachments 1 and 2 to this letter,

September 21, 2011 U. S. Nuclear Regulatory Commission Page 2 provides the response to RAI 1 and Attachment 2 to this letter provides the responses to RAIs 2 through 6. Attachment 2 contains information proprietary to Westinghouse Electric Company, LLC that is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit, provided in Attachment 3, sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390, "Public inspections, exemptions, requests for withholding." Accordingly, it is requested that the information in Attachment 2 be withheld from public disclosure in accordance with 10 CFR 2.390. Attachment 4 to this letter provides a non-proprietary version of the RAI responses provided in Attachment 2.

In support of the RAI responses provided in Attachments 1 and 2, several technical reports have been provided as attachments to this letter. Attachments 5 and 6 provide copies of EGC procedures used in the reload control and the bundle and core design process, respectively. Attachment 7 provides a copy of the Cycle Design Inputs and Requirements (i.e., CDIR) document for Quad Cities Unit 1, Cycle 22. Attachment 8 provides EGC Calculation QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan." 0 provides the Quad Cities Unit 1, Cycle 22 Bundle Design Report and 1 provides the Reference Loading Pattern Report for Quad Cities Unit 1, Cycle 22. Westinghouse Electric Company, LLC considers the information provided in these two reports to be proprietary. Therefore, an affidavit signed by Westinghouse, the owner of the information, is provided in Attachment 9. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390.

Accordingly, it is requested that the information in Attachments 10 and 11 be withheld from public disclosure in accordance with 10 CFR 2.390. Since the documents provided in Attachments 10 and 11 to this letter are considered to be proprietary in their entirety, non-proprietary versions are not provided.

Review of the original Quad Cities Unit 1 Cycle 22 SLMCPR report provided as to Reference 1 indicated several corrections were required. The SLO and DLO SLMCPR values in the report had incorrectly been identified as proprietary. Since these values are included in QCNPS TS 2.1.1, the values should not be withheld as proprietary. Therefore, the report has been revised to remove the proprietary brackets from around the SLMCPR values. Secondly, it was noted that two minor typographical errors needed correction. In the fifth paragraph on page 5 of 17, the SLMCPR at 95.3%

and 108% flow were updated to 1.1097 and 1.1029, respectively. The correction of these typographical errors has no impact on the results of the SLMCPR analysis. 2 to this letter provides Revision 2 of the SLMCPR report. Westinghouse Electric Company, LLC considers information provided in this report to be proprietary.

Therefore, an affidavit signed by Westinghouse, the owner of the information, is provided in Attachment 13. This affidavit sets forth the basis on which the information may be withheld from public disclosure by the NRC and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR 2.390. Accordingly, it is requested

September 21, 2011 U. S. Nuclear Regulatory Commission Page 3 that the information in Attachment 12 be withheld from public disclosure in accordance with 10 CFR 2.390. A non - proprietary version of this report is provided in Attachment 14.

EGC has reviewed the information supporting a finding of no significant hazards consideration that was provided to the NRC in Reference 1. The additional information provided in this submittal does not affect the bases for concluding that the proposed license amendment does not involve a significant hazards consideration. No new regulatory commitments are established by this submittal.

If you have any questions concerning this letter, please contact Mr. Timothy A. Byam at (630) 657-2804.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 21St day of September 2011.

Respectfully, David M. Gullott Manager - Licensing Exelon Generation Company, LLC Attachments:

1. Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit
2. Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Proprietary)
3. Westinghouse Affidavit for RAI 2 through 6 Responses
4. Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Non -

Proprietary)

5. EGC Procedure NF-AA-100, "Reload Control Procedure"
6. EGC Procedure NF-AB-1 10, "Bundle and Core Design (BWR)"
7. EGC Transmittal of Design Information document NF1000236 Rev. 1
8. EGC Calculation QDC-0000-N-1 804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan"
9. Westinghouse Affidavit for Bundle Design Report and Reference Loading Pattern Report
10. Attachment to NF-BEX-1 0-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary)
11. Attachment 1 to NF-BEX-1 0-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary)
12. Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Proprietary)

September 21, 2011 U. S. Nuclear Regulatory Commission Page 4

13. Westinghouse Affidavit for Quad Cities Unit 1 Cycle 22 SLMCPR Report Revision 2
14. Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Non-Proprietary)

ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit

ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit In reviewing the Exelon Generation Company's (Exelon's) submittal dated June 7, 2011, related to the request to replace the current single recirculation loop operation (SLO) safety limit minimum critical power ratio (SLMCPR) requirement, for the Quad Cities Nuclear Power Station (QCNPS), Unit 1, the NRC staff has determined that the following information is needed in order to complete its review:

RAI-01

Provide:

a. The details of the QCNPS, Unit 1, Cycle 22, analysis performed to obtain the final core loading pattern including procedure, guideline, criteria, and approved methodologies used for this analysis, and
b. The design document for the QCNPS, Unit 1, Cycle 22, core loading pattern.

Response to RAI-01 a The procedures and guidelines used for the QCNPS, Unit 1, Cycle 22 core loading pattern development are Exelon Generation Company, LLC (EGC) procedures NF-AA-100, "Reload Control Procedure" (Reference 1-1) and NF-AB-1 10, "Bundle and Core Design (BWR)," (Reference 1-2). EGC procedures documented in References 1-1 and 1-2 direct the bundle design and core reload process. The criteria are defined in the Cycle Design Inputs and Requirements (CDIR) document (Reference 1-3), which sets the design criteria for cycle energy, thermal margins, and other design constraints for the cycle. Following completion of the core loading pattern development, Westinghouse provides to EGC the Bundle Design Report (Reference 1-4) and the Reference Loading Pattern Report (Reference 1-5) which describe in detail the bundle design and loading pattern. In addition, Reference 1-5 includes an explanation of how the CDIR design requirements are satisfied.

The approved Westinghouse methodologies used to evaluate the core loading are documented in Reference 1-7. The results of implementing these methods are used to generate References 1-4 and 1-5.

EGC Procedures NF-AA-100, NF-AB-1 10, and the EGC Transmittal of Design Information document NF1000236 Rev. 1 (i.e., Q1C22 CDIR) are provided as Attachments 5, 6, and 7, respectively.

Response to RAI-01 b Westinghouse documentation of the final core loading pattern is provided in References 1-4 and 1-5. The EGC documentation of the final as-loaded core loading pattern is documented in Calculation QDC-0000-N-1 804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" (Reference 1-6). Reference 1-6 documents the as-loaded Q1C22 core as well as associated technical information about the core loading used for EGC internal activities.

Page 1 of 2

ATTACHMENT 1 Response to NRC RAI-01 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit Westinghouse proprietary References 1-4 and 1-5 are provided as Attachments 10 and 11 to this letter, respectively. EGC Calculation QDC-0000-N-1 804 is also provided as to this letter.

References RAI-01 1-1 NF-AA-100, "Reload Control Procedure," Revision 13 1-2 NF-AB-1 10, "Bundle and Core Design (BWR)," Revision 10 1-3 Cycle Design Inputs and Requirements for Quad Cities Unit 1, Cycle 22, dated November 30, 2010 1-4 Attachment to NF-BEX-10-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22," dated November 2010 1-5 Attachment 1 to NF-BEX-1 0-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22," dated December 2010 1-6 QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan" 1-7 CENPD-300-P-A, Revision 0, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996 Page 2 of 2

ATTACHMENT 3 Westinghouse Affidavit for RAI 2 through 6 Responses

Westinghouse Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja @ westinghouse.com Rockville, MD 20852 Proj letter: NF-BEX-11-142 CAW-1 1-3247 September 15, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

USBWR- 11-25 -P-Attachment, "Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383)" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3247 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3247, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, J. A. Gresham, Manager Regulatory Compliance Enclosures

CAW- 11-3247 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 145 ^4 day of September 2011 Notary Public COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member, Pennsylvania Association of Notaries

2 CAW-11-3247

{1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse, (2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit, (3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse, (ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-11-3247 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use 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 a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-11-3247 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage, (iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in USBWR-11-25-P-Attachment, "Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383)" (Proprietary), for submittal to the Commission, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.

This information is part of that which will enable Westinghouse to:

(a) Support Exelon's use of Westinghouse Fuel at Quad Cities.

5 CAW-11-3247 (b) Assist the customer to obtain license change.

Further this information has substantial commercial value as follows:

(a) Westinghouse can use this information to further enhance their licensing position with their competitors.

(b) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

Proprietary Information Notice Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant -specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

Copyright Notice The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

ATTACHMENT 4 Response to NRC RAI-02 through RAI-06 Supporting the Request for Technical Specification Change for Minimum Critical Power Ratio Safety Limit (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 USBWR 11-25-NP-Attachment Responses to NRC Request for Additional Information on Quad Cities Technical Specification Change for Minimum Critical Power Ratio Safety Limit (TAC No. ME6383) (Non-Proprietary)

Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry Township, PA 16066 J 2011 Westinghouse Electric Company LLC All Rights Reserved

Page 2 of 10 USB WR-11-25-NP-Attachment

RAI-02

Provide the fuel bundle critical power ratio distribution in the core for the limiting point in the cycle.

Response to RAI-02 Figure RAI-02-1 shows the fuel assembly minimum critical power ratio values for the most limiting point in the cycle according to Figure 4 of Attachment 3 of the Exelon submittal. The values are a,c shown [ I

Page 3 of 10 USB WR-11-25-NP-Attachment Figure RAI-02 QCNPS, Unit 1, Cycle 22 - Minimum CPR distribution for the limiting point in the cycle a,c

Page 4 of 10 US B W R- 11-2 5-NP-Attachment Figure RAI-02 QCNPS, Unit 1, Cycle 22 - Minimum CPR distribution for the limiting point in the cycle a, c

Page 5 of 10 USBWR-1 1-25-NP-Attachment

RAI-03

Provide:

a. Verification that the power to flow map bounds QCNPS, Unit 1, Cycle 22, operation including stability Option III features of scram region and controlled entry region for back up stability projection, and
b. A list of approved methodologies used to perform the above stability calculations.

Response to RAI-03a Figures RAI-03-1 and RAI-03-2 show the QCNPS, Unit 1, Cycle 22 BSP Stability results for nominal and reduced feedwater heating respectively. These figures show that the Cycle 22 stability results are bounded by the backup stability protection scram and controlled entry regions.

Figure RAI-03-1 QCNPS, Unit1, Cycle 22 BSP Stability Results - Nominal FWT Band

Page 6 of 10 U S B W R-11-25-NP-Attachment Figure RAI-03-2 QCNPS, Unit 1, Cycle 22 BSP Stability Results - Reduced FWT Band Response to RAI-03b The methodologies used to perform the Stability Analysis for QCNPS, Unit 1, Cycle 22 are included in the following references.

3-1 CENPD-295-P-A, "Thermal-Hydraulic Stability Methodology for Boiling Water Reactors",

July 1996.

3-2 GE-NE-0000-0028-9741-R1, "Plant-Specific Regional Mode DIVOM Procedure Guideline",

June 2005.

Page 7 of 10 U SB WR-11-25-NP-Attachment

RAI-04

Provide the rationale why a 30.4 percent reload batch fraction for SVEA-96 Optimal fuel caused the proposed SLMCPR increment of 0.01 for SLO and no change for TLO for the proposed loading pattern in Figure 1 of Attachments 3 and 5 of the Exelon submittal.

Response to RAI-04 The batch change in the reload fraction for SVEA-96 Optimal did not cause the change in the result for SLMCPR for SLO and no change for TLO. The reason for the change in the result for SLMCPR is described in detail in the response to Question RAI-05c.

Page 8 of 10 USB WR-11-25-NP-Attachment

RAI-05

Provide:

a. Or make available for audit, copy of the McSLAP computer code including theory and user's manual;
b. A description of the relationship between the McSLAP computer code and CENPD-300-P-A;
c. Details of two errors found in Westinghouse's McSLAP computer code including applicable portions of the methodology described in CENPD-300-P-A, and their impact on the SLMCPR calculations including those at exposure beyond 4000 MWd/MTU; and
d. Or make available for audit, a flow chart consisting of approved methodologies used for this SLMCPR calculation with respect to their input to each sub-routine calculation.

Response to RAI-05a and RAI-05d Westinghouse will make available for audit in the Westinghouse Rockville office a copy of the McSLAP computer code including the theory / user's manual. The documents contain a general overview flow chart used for SLMCPR calculation.

If the NRC needs to perform code executions, Westinghouse needs to be informed in advance in order to coordinate sending personnel to the Rockville office.

Response to RAI-05b The McSLAP code uses a statistical approach for sensitivity and uncertainty analysis of MCPR using a Monte Carlo technique. The McSLAP code integrates the Monte Carlo approach described in Section 5.3.2 of Reference 5-1.

An example of the SLMCPR evaluation is provided in Appendix D section D.4.4 of CENPD-300-P-A (Reference 5-1). The NRC request for additional information for Reference 5-1, questions F 11 and F 12, contain further clarification of the methodology, Response to RAI-05c The SLMCPR calculations for QC 1 C22 are impacted by these issues from the beginning of cycle to the end of cycle, not just beyond the 4000 MWd/MTU exposure point. The exposure point is mentioned in the transmittal to provide a time in cycle by which the revised SLMCPR values need to be implemented. Prior to the 4000 MWd/MTU exposure point, even with the issues described in the response to Question 5c, the current Technical Specification SLMCPR values of 1.11 and 1.13 for SLO and TLO respectively would remain bounding.

The following three issues have been reported against the McSLAP methodology.

McSLAP Issue 1

Description:

a,c

Page 9 of 10 USB WR-11-25-NP-Attachment Ia,c McSLAP Issue 2

Description:

I a,c McSLAP Issue 3

Description:

I a,c I

Mc SLAP Issues Impact:

J a,c References RAI-05:

5-1 CENPD-300-P-A, Revision 0, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996

Page 10 of 10 U SB WR-11-25-NP-Attachment

RAI-06

Describe any Part 21 issues relating to the fuel design applied to the QCNPS, Unit 1, Cycle 22, fuel assemblies. Also, with respect to the Part 21 issues, identify all affected factors, and quantify their impact on the parameters shown in Table 2 of Attachment 3.

Response to RAI-06 There are no Part 21 reportable issues related to the fuel design applied to the QCNPS, Unit 1, Cycle 22 fuel assemblies. With respect to the issue associated with this Technical Specification change request, it was determined not to be reportable tinder IOCFR21 due to the existence of offsetting conservatisms.

ATTACHMENT 5 EGC Procedure NF-AA-100, "Reload Control Procedure"

Exelon, NF-AA-100 Revision 13 Page 1 of 54 Nuclear Level 3 - Information Use RELOAD CONTROL PROCEDURE

1. PURPOSE 1.1. This procedure defines the steps to be followed for the effective specification, design, review, approval and implementation of reload fuel assemblies, core designs and related services (including core component management) for Exelon nuclear reactors. (Reference NF-AA-10) 1.2. This procedure defines the requirements for documenting Core Reload and Cycle Management configuration changes. There are T&RMs that detail the processing of Fuel Change Package (FCP) type Engineering Changes (EC - Passport) or Engineering Change Requests (ECR - PIMS) for Core Reload and Cycle Management configuration changes. (Reference NF-AA-100-1000) 1.3. This document is not a detailed specification of the reload procedures to be followed for each unit. That detail is provided by other corporate and site specific procedures and T& RMs. Nor does this procedure repeat the detail of applicable general topics such as: task control; documentation control; engineering reviews; etc., which are the subject of other procedures.

1.4. This procedure is also applicable for mid-cycle outages that implement core re-designs (e.g. replacement of failed fuel). Although not all steps within this procedure may be required to be executed, each step should be evaluated for applicability. Those steps that assure safe, economic and reliable plant operation and those steps that ensure communication of the redesign details and their impact on subsequent plant operations are essential.

2. TERMS AND DEFINITIONS 2.1. 50. 59 Review - a review performed in accordance with 1 OCFR50.59 to determine if a proposed change, test or experiment requires prior NRC approval via license amendment under 1 OCFR50.90. (Reference LS-AA-1 04) 2.2. Advance Work Authorization (AWA) - authorization to proceed at risk with installation work activities in the field without issuance of a completely approved configuration change. The advance work does not affect any in-service equipment and the equipment is not placed in service and is not declared operable, and has no impact on operating or in-service equipment until the configuration change is approved.

2.3. Affected Document List (ADL) - an electronic database list that identifies the controlled documents affected by configuration change packages.

2.4. Beginning of Cycle (BOC) - the timeframe when an operating fuel cycle is initiated.

NF-AA-100 Revision 13 Page 2 of 54 2.5. Compensatory Actions - a real commitment of effort or material to reduce the probability or consequences of a risk; part of a Risk Management Assessment.

2.6. Configuration Change a change to the design basis, design documentation, or physical plant configuration 2.7. Contingency Plans - the action in response to a "risk trigger" (without necessarily committing to the actual effort or material to execute that plan); part of a Risk Management Assessment.

2.8. Core Components - components that are not intimately incorporated into fuel assemblies. They are either not located in the fuel assembly or could be re-used many times in different fuel assemblies. Therefore their procurement/ use is not intimately linked to the reload fuel. Examples of core components are: (PWR):

control rods; thimble plug assemblies; primary sources; secondary sources (BWR):

control blades.

2.9. Core Loading Plan CLP) - a document that contains the full core layout that indicates core-loading positions of all reload assemblies/bundles using the vendor-supplied identifiers and supporting information.

2.10. Core O peratin g Limits Re portort Ct7LR - a document that contains the core thermal and reactivity limits required by Technical Specifications for an operating fuel cycle.

2.11. Core Reload - the process of performing the engineering work necessary to refuel a reactor core. This work includes the specification, calculations, documentation, reviews, testing, and any updates to computer codes, databases, Technical Specifications, UFSAR, procedures and training necessary to operate the fuel cycle under design.

2.12. Cycle Design Inputs and Requirements (CDIR) - a formal document(s) containing cycle design targets and the inputs and acceptable outputs that define the boundaries of the cycle design and licensing analyses. The document(s) may be identified by vendor-specific naming conventions (see Attachment 2).

2.13. Cycle Management - the process of performing the engineering work necessary to implement changes to the cycle specific configurations. This work includes the specification, calculations, documentation, reviews, testing, and any updates to computer codes, databases, Technical Specifications, UFSAR, procedures and training necessary to implement changes to the current operating fuel cycle design.

2.14. Design Technical Review (DTR) - a technical review of the PFCD and Risk Management Assessment performed by the Reload Design Overview Team.

2.15. End of Cycle (EOC) - the timeframe when an operating fuel cycle is concluded.

NF-AA-1 00 Revision 13 Page 3 of 54 2.16. End of Rated (EOR) - the cycle exposure at which there is insufficient reactivity to achieve rated thermal power without using cycle extension maneuvers.

2.17. Energy Utilization Plan (EUP) - the approved schedule of core thermal energy requirements (i.e., cycle length, power level, effective full power days, and EOC extension options such as coastdown) and outage dates input to the reload design.

2.18. Final Fuel Cycle Design (FFCD) - the approved reload design that will be used for an operating fuel cycle.

2.19. Fuel - the assemblies (PWR) or bundles (BWR) of fuel rods assembled in a retaining structure. In this document, references to reload fuel should be taken to encompass both the reload-specific fuel and fuel related components.

2.20. Fuel Change Package (FCP) - a type of EC/ECR used to document, approve and implement core reload and cycle management activities.

2.21. Fuel Related Components - components that are intimately related to a single fuel assembly, i.e., their entire in-core lifetime is usually associated with installation in only one fuel assembly. Therefore their procurement/use is intimately linked to the reload fuel. Examples of fuel related components are (PWR): removable burnable poison rod assemblies (various designs such as Removable Burnable Poison Assemblies (RBPAs), Burnable Poison Rod Assemblies (BPRAs), Vibration Suppressors, or Wet Annular Burnable Absorbers (WABAs)); (BWR): channels, channel fasteners. Design and process changes to Fuel Related Components are evaluated as changes to Fuel Assemblies.

2.22. Lead Test Assembly (LTA) - a fuel assembly or core component with no preexisting in-reactor operating experience.

2.23. Lead Use Assembly (LUA) - a fuel assembly or core component with pre-existing in-reactor experience but may be its first application at an Exelon reactor.

2.24. Like-for-Like Replacement I Identical Replacement Item - the replacement item is considered like-for-like if there is a high level of confidence that there have been no changes in the design, material, or manufacturing process since the procurement of the item being replaced.

2.25. Monitoring Plan - a compensatory action to actively look for Risk Triggers and initiate contingency plans.

2.26. Nuclear Fuels (NF) - the Exelon Nuclear department responsible for fuel related engineering activities.

2.27. Nuclear Safety Review Board (NSRB) - a committee that oversees the management and operating practices of Exelon Nuclear.

NF-AA-100 Revision 13 Page 4 of 54 2.28. Operational Experience (OE) - nuclear industry information based on plant performance issues, lessons learned, regulatory reports, internal Exelon reports or via industry information exchange forums, such as OPEX from INPO.

2.29. Passport Engineering Change / PIMS Engineering Change Request (EC/ECR) -

approved design packages that govern configuration changes to the plants.

2.30. Plant Operations Review Committee (PORC) - a multi-disciplined committee responsible for review of activities that have potential to affect nuclear safety.

2.31. Preliminary Fuel Cycle Design (PFCD) - the preliminary reload design that is presented to the RDOT for a design technical review and to the RRB for NF management approval prior to finalizing the reload design for licensing.

2.32. Re-insert Fuel - fuel used in the reload cycle that has been previously irradiated and was stored in the spent fuel pool during the previous cycle.

2.33. Re-use Fuel - fuel used in the reload cycle that was operating in the reactor core in the previous cycle.

2.34. Reload- Desi g n Review Criteria - Criteria relating to Core Reload design and operating strategy changes and their impacts on: safety, operations, performance, monitoring, modeling, training, procedures and licensing, against which all reload designs must be reviewed. These criteria are generally vendor/unit specific and are specified in lower-level procedures and/or guidelines (see Attachment 3).

2.35. Risk Management Assessment - the end deliverable from the Risk Management Process: a document that identifies key risk management parameters and associated actions.

2.36. Risk Trigger - A deviation from a nominal range of acceptable values that will activate a contingency plan.

2.37. Senior Management Design Initialization (SMDI) - a meeting held early in the Reload Design process to ensure that senior site management actively participate in the decision making process for significant changes in the reload design, core operating strategy, fuel mechanical design, and fuel cost associated with the reload.

2.38. Updated Final Safety Analysis Report (UFSAR) - a document that contains all the changes necessary to reflect information and analyses submitted to the NRC per 10 CFR 50.71(e) since the submission of the original FSAR. UFSAR description includes text, tables, diagrams, etc., that provide an understanding of the design bases, safety analyses and facility operation under conditions of normal operation, anticipated operational occurrences, design basis accidents, external events, and natural phenomena for which the plant is designed to function.

NF-AA-100 Revision 13 Page 5 of 54

3. RESPONSIBILITIES NOTE: Other personnel may perform activities allocated to the Core Designer within this procedure as determined by lower-level procedures or as appointed by the applicable NF manager. However, the Core Designer retains responsibility for overall cognizance and coordination of the reload process within NF.

3.1. Chemistry RDOT Member - the individual in the site Chemistry department allocated responsibility for ensuring that the historical and projected chemistry parameters used as input into the RDOT are correct, all planned chemistry changes over the time period reviewed by the RDOT are accurately described and the possible impacts are reviewed, and relevant chemistry industry operating experience is appropriately incorporated into the RDOT review.

3.2. Core Designer - the individuals in the NF organization allocated responsibility for the bundle/core design(s) and oversight of the reload design process. This individual is responsible for ensuring the Core Reload and Cycle Management activities are documented on the appropriate FCP EC/ECR. This individual communicates Core Reload and Cycle Management related issues to management, appropriate site organizations, fuel vendor(s) and core component vendor(s).

3.3. Director of NF - Provides authorization of AWAs that affect the functional design, design criteria of the plant or the 1 OCFR50.59 Evaluation, if applicable.

3.4. EC Coordinator (Passport only) - performs administrative review of the FCP.

Ensures all documents and attachments are in the FCP package. Ensures that the ADL and the install attributes are correctly populated.

3.5. Fuel Buyer - procures nuclear fuel and fuel related components.

3.6. Fuel Reliability Engineer (FRE) - initiates detailed reviews of new or changed fuel assemblies and core component design and fabrication process changes.

Responsible for determining the impact of chemistry changes identified at RDOT meetings on fuel performance.

3.7. Manager, BWR/PWR Design - a manager within NF with responsibility for Core Reload and Cycle Management support to the unit under consideration. This individual assigns the Core Designer, and any delegates, as well as necessary reviewers and verifiers. Provides authorization for AWAs that do not affect the functional design, design criteria of the plant or the 1 OCFR50.59 Evaluation, if applicable.

NF-AA-100 Revision 13 Page 6 of 54 3.8. NF Data Bank (NFDB) Analyst - updates the NFDB for each site based on fuel assemblies discharged and fresh fuel assemblies introduced by the reload 3.9. NF Engineer for Fuel Fabrication Oversight - coordinates the monitoring of the nuclear fuel and core component fabrication process.

3.10. NF Vice President - The NF VP has final approval authority for the EUP, fuel design changes, and the final fuel cycle design. Chairs the Reload Review Board.

3.11. Nuclear Oversight Vendor Auditor - conducts assessments and surveillances of vendors, as appropriate, to evaluate design, licensing, and fabrication activities for fuel and other related components and services.

3.12. Outage Services -- the organization that receives and installs nuclear fuel and core components.

3.13. Reactor Engineer (RE) - the individuals in the site organization who are the primary interface with the Core Designer during the Core Reload and Cycle Management activities.

3.14. Reload Design Overview Team (RDOT) - a multi-discipline team made up of NF and site personnel that provides planning, input, and review of reload activities. It assures both that appropriate boundaries for the reload design are established, and that any changes to established boundaries required by the design are acceptable (see Attachment 1).

3.15. Reload Review Board (RRB) - an internal NF board that provides management challenge, oversight and review of all Core Reload Designs. The RRB oversees the reload strategy, fuel cycle economics and business issues, impacts of design decisions, and communication of reload plans outside of NF. The RRB meets as required to assure that each reload receives the appropriate level of oversight.

Various steps of this procedure require the NF VP to obtain RRB concurrence of the approach to be adopted (see Attachment 4) 3.16. Site Vice President - The Site VP provides final site concurrence for core reloads and for significant fuel design changes.

3.17. Supp ly Management Bu yer - procures core components.

NF-AA-100 Revision 13 Page 7 of 54

4. MAIN BODY NOTE: Tasks in this procedure related to FCP preparation and independent review by NF personnel are associated with Certification Guide N-AN-ENG-CERT-NF05. Certification Guide requirements for other tasks are defined in the applicable governing procedures/T&RMs, as necessary.

NOTE: The procedure steps in Section 4 include many tasks that may be performed concurrently or in a different sequence.

Essential sequencing is explicitly stated where necessary.

Detailed schedules are specified in vendor-specific guidelines. Refer to NF-AB-100-4000, NF-AB-100-5000, NF-AB-100-6000, NF-AP-100-7000 and NF-AP-100-8000.

NOTE: Steps 4.1 and 4.2 deal with long lead-time items that are typically addressed well in advance of the bulk of the reload process.

4.1. Core Component Management and Replacement 4.1.1. TRACK and PROJECT the exposure and/or lifetime history of core components.

This responsibility is typically divided between NF and Plant / Reactor Engineering as follows:

BWR control blades; PWR in-core detectors, neutron sources, and control rod assemblies. (Core Designer or Plant / Reactor Engineering)

BWR Local Power Range Monitors (LPRM), Wide Range Neutron Monitors (WRNM), and core support plate plugs. (Plant / Reactor Engineering) 4.1.2. PERFORM or COMMISSION any technical analysis required to determine which core components need replacement for the reload cycle, due to lifetime issues or other constraints. (Core Designer or Plant / Reactor Engineering) 4.1.3. DEVELOP replacement and core shuffling strategies for the above core components. (Core Designer or Plant / Reactor Engineering)

1. REVIEW any changes in design attributes of the core components.
2. NOTIFY the Core Designer of any design changes in components that will impact the reload cycle design inputs or requirements.
3. PROVIDE replacement requirements to the site organizations responsible for budgeting the items and preparing purchase requisitions.
4. ENSURE any core components listed in 4.1.1 that are to be replaced in the upcoming refueling outage are included in the outage work scope. Refer to OU-AA-101.

NF-AA-100 Revision 13 Page 8 of 54 4.1.4. PREPARE configuration change packages for core component changes. Refer to CC-AA-1 03 and NF-AA-1 01. (FRE or designated individual) 4.1.5. PROCURE any necessary core components from approved suppliers. (Supply Management Buyer)

1. REVIEW current fuel and/or core component contract(s) for applicability. A unique contract (including a bid specification, bid evaluation, contract award process, etc.) may be required.

4.1.6. INSTALL the core components and VERIFY their adequate operation prior to reload cycle operation. (Outage Services or Site Maintenance) 4.2. Fuel Design Selection NOTE: The following review of fuel designs and fuel related component designs pertains to fuel product lines or hardware options that are mechanically different (e.g.,

GE14 to GNF2, advanced cladding material, channel material, mixing vanes, etc.) from that used previously for the station in question. The review of neutronically different bundle designs is considered part of the Core Reload design process.

NOTE: For significant fuel design changes, a Special SMDI meeting should be conducted at a time early enough to accommodate changes in strategy.

4.2.1. REVIEW available fuel designs and fuel related component designs (e.g., channels, WABAs, BPRAs) to identify designs that have potential to be used in the reload.

(Core Designer and FRE) 1 CONSIDER the following items when reviewing fuel designs:

suitability of the current fuel design product lines available under current fuel vendor contracts product lines available from other fuel vendors degree of confidence in and experience with available products changes necessitated by Fuel Reliability or operating experience (OE) issues changes necessary to achieve unit objectives such as the EUP, enhanced economics, performance goals, or operational goals applicability of current in-core and out-of-core methods (CM-1)

NF-AA-100 Revision 13 Page 9 of 54 significance of any design changes relative to the current fuel design mechanical/nuclear compatibility with dry storage/transport system deployed at the site, as applicable 4.2.2. RECOMMEND the fuel design to be used for the Core Reload design. (Manager, BWR/PWR Design)

1. SUPPORT the recommendation with initial evaluations relating to the items listed in 4.2.1.1. Detailed evaluations that are required prior to fuel design implementation may be completed later (see Step 4.2.4).
2. RECOMMEND taking part in a new fuel design evaluation or test loading at this stage, if applicable.

4.2.3. CLASSIFY the fuel design selected for the Core Reload. Refer to NF-AA-101. (FRE)

1. If the fuel design selected is "Like-for-Like" to the fuel design currently in use at the unit, then DOCUMENT this in the Core Reload FCP and SKIP to Step 4.3. Otherwise PROCEED to Step 4.2.4.

4.2.4. PREPARE a configuration change package for the new or changed fuel design.

Refer to CC-AA-103 and NF-AA- 101. (FRE or designated individual)

1. IDENTIFY all new supporting reviews or analyses that are needed based on the results of the NF-AA-1 01 review.
2. ENSURE that Core Reload design inputs or requirements that are impacted by fuel design changes are addressed in the CDIR (Step 4.3.8).

4.2.5. APPROVE the selection of a new/changed fuel design before Core Reload design work begins. This approval shall have the concurrence of the Reload Review Board (Reference Attachment 4 for conduct of RRB meeting). (NF Vice President)

1. If the change includes a change in fuel vendor, then OBTAIN the written concurrence of the President and Chief Nuclear Officer.
2. If the change is not approved, then DIRECT the Core Designer to review alternate fuel design options (i.e., RETURN to Step 4.2.1).

4.2.6. PRESENT significant fuel design changes to senior site management via a special Senior Management Design Initialization (SMDI) meeting (Step 4.4). (Core Designer or FRE) 4.2.7. If LTAs/LUAs from a different fuel vendor are being introduced, then CONSIDER expanding the duration of the reload schedule to allow for transfer of analysis results between the two fuel vendors. (Core Designer)

NF-AA-100 Revision 13 Page 10 of 54 NOTE: During the Core Reload design process, numerous design analyses (internal and/or external) will be generated to support the reload. These design analyses shall be prepared, reviewed, approved and retained in accordance with CC-AA-309. Processing of these design analyses shall be done using a Core Reload FCP per this procedure and NF-AA-100-1000.

4.3. Core Reload Design Inputs, Assumptions, and Requirements 4.3.1. INITIATE and PREPARE a Core Reload FCP type EC/ECR to control the reload configuration change. Refer to NF-AA-100-1000. (Core Designer) 4.3.2. PERFORM a review of historic (previous 2 to 3 years) and current industry operating experience information to assure that applicable experiences are considered when preparing or reviewing reload products in order to prevent similar problems. Refer to NF-AA-100-1010. (Core Designer) 4.3.3. PERFORM a Technical Pre-Job Brief for the Core Reload design effort. Refer to HU-AA- 1212. (Core Designer and Manager, BWR/PWR Design)

1. DISCUSS Nuclear Fuels, Inter-Department and potential Independent Third Party reviews. The Pre-Job Brief is expected to be a face-to-face discussion with the Manager, BWR/PWR Design, taking the appropriate questioning/challenging role.

4.3.4. IDENTIFY members of the Reload Design Overview Team (RDOT). (Core Designer)

1. RDOT members and responsibilities are described in Attachment 1.

4.3.5. DEFINE the reload Energy Utilization Plan (EUP). Refer to NF-AA-105-1000. (Core Designer)

1. CONSULT the unit business plans and Exelon Planned Outage Schedule to ensure that targets for core energy output (i.e., cycle length, power level, effective full power days, and end-of-cycle extension options such as coastdown) are met;
2. If the plant is generator-limited or if the reactor operating power level changes due to seasonal limitations (e.g., river temperature, grid voltage), then OBTAIN a projected reactor power level load profile for the design cycle from the Station;
3. DETERMINE the expected energy requirements for the reload cycle plus additional forecast cycles as specified in NF-AA-105-1000;

NF-AA-100 Revision 13 Page 11 of 54

4. DOCUMENT the EUP in the Core Reload FCP and the appropriate vendor transmittals.

4.3.6. SCHEDULE and CONVENE a Reload Design Kickoff Meeting with the fuel vendor.

Members and responsibilities are described in Attachment 5. (Manager, BWRIPWR Design) 1 If required, then ENSURE a tracking item is generated to process/evaluate any identified change(s), such as methodology changes, product line updates, shipping requirements, and manufacturing changes.

2. IDENTIFY any long lead-time items, such as Technical Specification Change Requests or other licensing actions that are expected to be required.
3. IDENTIFY impacts to the training simulator that may result from fuel design changes or changes in the fuel design codes. Changing fuel design codes may result in the inability to perform a cycle specific simulator core update because the simulator uses output from the design codes.

A. If simulator updates are impacted, COORDINATE with Site Training and IT actions necessary to maintain cycle specific simulator core updates.

B. BE AWARE that this impact may have budget and long lead-time implications.

4.3.7. PRESENT the initial EUP to the Reload Review Board before the Fuel Cycle Design activities begin (Refer to Attachment 4 for conduct of RRB meeting). (Core Designer)

1. PREPARE the RRB presentation in accordance with the NF standard template. Refer to NF-AA-100-2000.
2. INCORPORATE changes required by the RRB and DOCUMENT RRB concurrence with the EUP, If the EUP includes a change in rated power level or significant change in operating strategy, then OBTAIN the written concurrence of the President and Chief Nuclear Officer.
4. If necessary, then REVISE the EUP with the approval of the appropriate Manager, BWR/PWR Design. For significant changes to the EUP due to cycle design actualities (e.g., when achieving the EUP does not offer optimum economics, when an acceptable core design cannot be achieved, or when a significant change in outage/cycle length occurs), OBTAIN the concurrence of the NF VP.

N F-AA-1 00 Revision 13 Page 12 of 54 4.3.8. DEFINE the content of the Cycle Design Inputs and Requirements (CDIR) for the reload cycle. The format of the CDIR may be unit-specific and is described by lower level procedures and/or guidelines. Typical items to be included are listed in Attachment 2. The CDIR may consist of multiple documents that are required at different phases during the reload process (e.g., cycle design phase, reload licensing phase, etc.). (RDOT) (CM-3)

1. COORDINATE the preparation of the CDIR. (Core Designer)
2. INCLUDE the EUP specified by the Reload Review Board in Step 4.3.7.

(Core Designer)

3. INCLUDE applicable design considerations and impacts using the Design Attribute Review (DAR). Refer to CC-AA-1 02. (Core Designer)
4. PROVIDE a draft copy of the CDIR to RDOT members early enough (approximately 2 weeks) to allow sufficient review time. (Core Designer) 4.3.9. PERFORM a Risk Management Assessment of the Reload. (Core Designer)
1. The scope of the assessment must CONSIDER all risks from initial fuel receipt of the reload batch through the end of the first cycle (including the offload and shuffle at the end of the cycle), and INCLUDE risks associated with significant changes to the core-operating regime or fuel design. The assessment scope may also be extended to risks during the design process and subsequent cycles.

NOTE: The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per regulatory processes such as 50.59, 50.90, 50.46, etc.

2. CONSIDER in all risk management assessments any changes to the vendor or Exelon methods that support core operation.

ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.

ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.

- DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).

NF-AA-100 Revision 13 Page 13 of 54

3. PERFORM the following required elements in the Risk Management Assessment: (CM-12)

- IDENTIFY Risk Triggers

- DETERMINE Compensatory Actions

- IDENTIFY Contingency Plans

- IDENTIFY Monitoring Plans

4. DETERMINE if an Independent Third Party Review is required. Refer to HU-AA-1212 and NF-AA-100-1600.
5. COMMUNICATE the resulting Risk Management Assessment with NF and Station management through the RRB and SMDI processes described in subsequent steps.

4.3.10. REVIEW and CONCUR with completed CDIR documentation and the Risk Management Assessment before they are used in definitive design analyses.

(RDOT)

1. Appropriate NF and site representatives should PERFORM this review to ensure that the personnel responsible for said items approve. This may include representatives on the RDOT and others as necessary.

4.3.11. APPROVE the Risk Management Assessment before the reload design is released for definitive design analyses. (Manager, BWR/PWR Design) 4.4. Senior Management Review and Approval NOTE: The SMDI review of the cycle design may be performed after completion of the Preliminary Fuel Cycle Design (PFCD) in Step 4.5.

4.4.1. PRESENT the key information for the proposed cycle design (e.g., EUP, fuel design, operating strategy, modifications, training requirements, reload schedule, Risk Management Assessment, recent fuel vendor audit/assessment results, near term future audit/assessment goals, etc.) to senior site management via the Senior Management Design Initialization (SMDI) meeting. The scope and focus of the SMDI meeting and the personnel required to attend are discussed in Attachment 4.

(Core Designer) (CM-12)

1. PREPARE the SMDI presentation in accordance with the NF standard template. Refer to NF-AA-100-2000.

4.4.2. If the cycle design plans are acceptable, then DOCUMENT SMDI concurrence.

(Core Designer)

NF-AA-100 Revision 13 Page 14 of 54 4.4.3. If the cycle design plans are not acceptable, then DEFINE requirements to be satisfied for concurrence. (SMD!)

1. ADD the resulting requirements to the reload support work scope, CDIR, and/or schedule. (Core Designer)
2. If the SMDI is held after the PFCD is developed in Step 4. 5 and the SMDI requirements require a significant change to the PFCD and/or its characteristics, then RETURN to earlier analysis and approval Steps of this procedure. (Core Designer) 4.5. Preliminary Fuel Cycle Design (PFCD)

NOTE: Subject to the election of appropriate commercial options in the fuel contract, either the fuel vendor or NF staff may perform the bundle/cycle design activities discussed in this section.

4.5.1. DEVELOP Scoping Fuel Bundle/Cycle Design(s) to be considered as the first step leading to the identification of the PFCD. Criteria for developing alternate designs may include operating margin, core limits, economics, EUP, operating strategy, etc.

(Core Designer)

1. DETERMINE the range of design options to be considered. (Manager, BWRIPWR Design)

A. SPECIFY the level of detail required, which need not be that required for formal calculations or analyses, and SPECIFY design comparison criteria.

B. SPECIFY the number of future operating cycles to be considered for each design option.

2. USE the design inputs specified in Step 4.3 of this procedure to design Scoping Fuel Bundle/Cycle Design(s). Some flexibility in the design inputs may be allowed to fully evaluate options. (Core Designer)

A. COLLABORATE with the fuel vendor, as needed, in developing bundle/cycle designs.

B. TRANSMIT the EUP and CDIR to the vendor. Refer to CC-AA-310.

This transmittal may be delayed if the vendor is not involved in the PFCD.

NF-AA-1 00 Revision 13 Page 15 of 54 4.5.2. REVIEW the proposed Scoping Fuel Bundle/Cycle Design alternatives and SELECT the design to be taken forward to the more detailed PFCD design stage. (Manager, BWRIPWR Design)

1. CONSULT the Reload Review Board, if necessary (e.g., if the EUP is revised significantly or a significant change in operating strategy is proposed). (Core Designer)

OBTAIN confirmation from the fuel vendor that the selected reload fuel assembly/bundle design(s) can be fabricated, licensed, and shipped under the terms of the current contract. (Core Designer) (CM-2)

A. IDENTIFY any potential manufacturing problems or constraints and CONSIDER vendor requests for changes to the design that will assist in the manufacturing campaign.

OBTAIN vendor experience or feedback on any other aspect of the fuel designs and IDENTIFY potential improvements.

4.5.3. FINALIZE the Preliminary Fuel Cycle Design (PFCD) based on the Scoping Fuel Cycle Design selected. (Core Designer) (CM-9)

1. IMPLEMENT the approved Core Design Inputs and Requirements, including both the approved EUP and the approved Fuel Design, into the PFCD.
2. MODEL the reload cycle plus two additional cycles to provide long-term fuel cycle optimization and economics information. The additional cycles may be modeled with existing or proposed fuel designs.
3. REVIEW the Risk Management Assessment, and MODIFY if necessary.

A. If the Risk Management Assessment (RMA) is significantly modified, COMMUNICATE the new RMA to NF and Station management (SMDI membership), per NF-AA-100-1600, if they have already received the RMA at this point in the reload process. Non-significant/editorial changes do not require senior management review.

4.5.4. CONVENE the RDOT to perform a Design Technical Review (DTR) of the PFCD.

(Core Designer) (CM-4)

1. PRESENT the PFCD and the Risk Management Assessment and EVALUATE the proposed cycle design against the CDIR and the Reload Design Review Criteria specified in lower level procedures and/or guidelines.

Criteria to be considered are listed in Attachment 3. (Assigned NF Engineers)

NF-AA-100 Revision 13 Page 16 of 54 A. ENSURE that the core loading pattern clearly identifies the locations of the various fuel batches in the core (e.g., using batch designators, color coding, etc.)

2. ENSURE that all inputs, requirements and impacts relating to their area of expertise are being addressed. (RDOT)
3. RAISE any concerns that may impede successful implementation of the PFCD. (RDOT)

A. RESOLVE RDOT concerns or ASSURE that concerns can be resolved by future reload work scope (Core Designer)

B. ADDRESS new design considerations and impacts in the FCP.

(Core Designer)

1. If there is an unresolved concern that may jeopardize the PFCD, then RETURN to Step 4.5.1, 4.5.2, or 4.5.3, as necessary.

C. Promptly NOTIFY the Manager, BWR/PWR Design, NF VP, and site management of significant issues or concerns. (Core Designer)

4. DOCUMENT the minutes of the DTR. (Core Designer)

A. If needed, then IDENTIFY all work scope necessary to support design implementation including any significant design changes or new fuel design implementation requirements.

4.5.5. PRESENT the PFCD and the Risk Management Assessment to the Reload Review Board (Refer to Attachment 4 for conduct of RRB meeting). (Core Designer)

1. PREPARE the RRB presentation in accordance with the NF standard template and NOTE changes from the initial RRB meeting. Refer to NF-AA-100-2000.
2. INCORPORATE any changes required by the RRB and DOCUMENT RRB concurrence with the PFCD and the Risk Management Assessment.

A. If Significant changes occur and if needed, then RETURN to earlier Steps of this procedure.

NF-AA-100 Revision 13 Page 17 of 54 B. If the Risk Management Assessment (RMA) is significantly modified per the RRB, COMMUNICATE the new RMA, per NF-AA-100-1600, to any NF and Station management (SMDI membership) who have already received the RMA at this point in the reload process. Non-significant/editorial changes do not require senior management review.

4.5.6. PROVIDE the multi-cycle PFCD prediction to the Fuel Buyer for the basis of long-term nuclear materials, conversion and enrichment requirements forecasts. (Core Designer)

1. If later changes associated with the Final Fuel Cycle Design or the as-loaded core significantly affect this information, then PROVIDE updated data.

4.5.7. PROVIDE the multi-cycle PFCD prediction to the fuel accounting organization, to form the basis of fuel accounting and economics forecasts. (Fuel Buyer)

1. If later changes associated with the Final Fuel Cycle Design or the as-loaded core significantly affect this information, then PROVIDE updated data.

4.5.8. OBTAIN fuel vendor review of the PFCD, if included in the scope of the applicable fuel contract. (Core Designer)

1. IDENTIFY any potential manufacturing problems or constraints and CONSIDER vendor requests for changes to the design that will assist in the manufacturing campaign.
2. OBTAIN vendor experience or feedback on any other aspect of the design.
3. COLLABORATE with the vendor to identify improvements to the PFCD that can be incorporated into the Final Fuel Cycle Design.
4. ADDRESS new design considerations and impacts in the FCP.

AGREE on the preliminary schedule for final fuel delivery and all other company and vendor deliverables required under the fuel fabrication contract.

4.5.9. For certain vendor-specific reload plans, the PFCD may be used as the basis for the fuel order. If the fuel order is placed based on the approved PFCD, then PERFORM Steps 4.7.1 thru 4.7.4 at this point. (Core Designer)

NF-AA-100 Revision 13 Page 18 of 54 4.6. Final Fuel Cycle Design NOTE: Subject to the election of appropriate commercial options in the fuel contract, either the fuel vendor or NF staff may perform the cycle design activities discussed in this section.

NOTE: The FFCD need only model the upcoming Core Reload cycle. Multi-cycle modeling is not required.

4.6.1. DEVELOP the Final Fuel Cycle Design (FFCD). (Core Designer)

1. INCLUDE any changes on which RDOT and/or SMDI concurrence are based.

(Core Designer)

2. COLLABORATE with the fuel vendor and INCORPORATE any vendor-recommended design changes agreed by the Manager, BWR/PWR Design.

(Core Designer)

3. VERIFY the material condition of any re-insert fuel (previously irradiated fuel that is resident in the fuel pool during the current operating cycle) utilized in the FFCD. (Core Designer)

A. OBTAIN material condition information from Reactor Engineer and/or Fuel Reliability Engineer. (Core Designer)

B. DEVELOP an action plan (e.g., FFCD redesign, fuel inspection, fuel repair, etc.) if the material condition of any re-insert fuel is questionable or needs repair. (Core Designer, Reactor Engineer and NF Technical Support Engineer)

4. ASSESS the impact of any changes between the PFCD and FFCD, versus the CDIR, the Risk Management Assessment, and the Reload Design Review Criteria. Significant changes or additional information not available during the PFCD phase may require a second design technical review per Step 4.5.4. (RDOT)

A. If required by Manager, BWR/PWR Design, then REQUEST vendor review of the FFCD. (Core Designer)

5. Promptly NOTIFY the Manager, BWR/PWR Design, Fuel Buyer and the RDOT; plus the NF VP and site management as necessary, of significant changes. Also, NOTIFY these and other stakeholders of minor changes at the earliest convenient opportunity. (Core Designer)

A. If significant changes occur, then OBTAIN RRB concurrence.

NF-AA-100 Revision 13 Page 19 of 54

6. VERIFY the design per applicable procedures and DOCUMENT the FFCD in the FCP. (Core Designer) 4.6.2. PROVIDE the affected departments (typically the RDOT members) with a summary of the FFCD for use in early identification of procedure and training requirements for the reload. (Core Designer)
1. IDENTIFY anticipated procedure changes, deletions, and additions. (RDOT)
2. IDENTIFY training changes, deletions, and additions, and FORWARD to the Training department to support identification and completion of training requirements. (RDOT)

NOTE: The PFCD may have been used as the basis for the fuel order instead of the FFCD per Step 4.5.9.

4.6.3. FREEZE the FFCD. The FFCD is to be used as the basis for the fuel order, licensing analyses and reload preparations. (Core Designer)

1. If needed, then EXECUTE small changes to the Reload Core design and new fuel requirements may be made subsequent to the FFCD, to better optimize the reload. Such changes must be acceptable to the fuel vendor under the terms of the fuel contract and of sufficiently minor impact not to invalidate any reload licensing work in progress.
2. If more significant perturbations occur, then CONSIDER these changes as a redesign and HANDLE as described in Step 4.9.5.

4.7. Fuel Procurement, Fabrication and Receipt 4.7.1. VERIFY that the following prerequisites/inputs have been addressed. (Core Designer)

- All necessary prior RRB approvals plus President and Chief Nuclear Officer concurrence, when necessary

- SMDI and RDOT concurrence

- FFCD (or PFCD) has been reviewed by the fuel vendor

- FFCD (or PFCD) has been reviewed against the CDIR 4.7.2. PROVIDE (or CONFIRM vendor provides) the finalized FFCD (or PFCD) enriched material requirements to the Fuel Buyer for procurement processing. (Core Designer)

NF-AA-1 00 Revision 13 Page 20 of 54

1. CONFIRM the enriched material delivery date. (Fuel Buyer) 4.7.3. ORDER fuel and fuel-related components based on the requirements of the approved FFCD (or PFCD). (Fuel Buyer)
1. VERIFY significant design changes/new fuel designs have been evaluated sufficiently and that unverified portions of changes are controlled. Refer to CC-AA-1 03 and NF-AA-1 01.
2. OBTAIN the necessary order information (e.g., quantities, designs, specifications, etc.) from the Core Designer.

A. ADJUST the Fuel order to account for inventory items to be either supplied or utilized.

3. OBTAIN required delivery schedules from the Site, Outage Services or per the relevant contract.

ORDER fuel and fuel-related components in compliance with all conditions (such as lead-time) of the relevant contract(s).

4.7.4. COORDINATE monitoring of the fabrication process. (NF Engineer for Fuel Fabrication Oversight)

1. MONITOR the progress of the fabrication campaign. (NF Engineer for Fuel Fabrication Oversight)
2. MONITOR generic issues associated with the fuel vendor and its fabrication facility. (NF Engineer for Fuel Fabrication Oversight)
3. CONDUCT assessments and surveillances of vendors, as appropriate, to evaluate design, licensing, and fabrication activities for fuel and other related components and services. (Nuclear Oversight Vendor Auditor) (CM-14)

NF shall SUPPLY technical specialists, as requested, to assist in the assessments of the fuel design, licensing, and fabrication activities.

NF will PROVIDE a summary report of the results of the most recent fuel vendor audits/assessments and actions for future SMDI meetings.

4.7.5. ESTABLISH final fuel and component delivery schedules and associated shipping and security instructions with the vendor(s). (Fuel Buyer and/or Site) 4.7.6. CONFIRM the reload design has been determined to be acceptable by the successful completion of all previous steps of this procedure. (Core Designer)

1. If required, then INCLUDE review of applicable CC-AA-103 and NF-AA-101 activities/analyses.

NF-AA-100 Revision 13 Page 21 of 54

2. If any previous Step remains incomplete just prior to scheduled Fuel Receipt, then NF shall NOTIFY both the Site and Fuel Vendor of any restrictions which must be applied pending that Step's successful completion.
3. If required, then ENSURE that all design analyses required to receive and store new fuel designs on-site have been completed and documented in the applicable DCP per NF-AA-101.
4. VERIFY that the reactivity requirements to receive and store the fresh reload batch on-site have been met.

4.7.7. RECEIVE fuel according to procedures that require a physical inspection of the fuel, plus a review of supporting vendor certifications and as-built data. (Outage Services)

1. ENSURE that the delivered fuel meets pre-defined acceptance criteria.
2. Any fuel that fails these inspections shall either: be RECTIFIED or REPLACED by the fuel vendor; or be subject to a full assessment to determine its suitability for use.

4.8. Reload Licensing and Core Loading Plan 4.8.1. If the PFCD was used for fuel procurement, then DEVELOP the FFCD per step 4.6.1 prior to Reload Licensing and Core Loading Plan activities. (Core Designer) 4.8.2. INITIATE the scope of work required for the successful licensing of the FFCD. (Core Designer)

1. DOCUMENT the Licensing and Core Loading information in the Core Reload FCP.
2. COLLABORATE with all stakeholders and the fuel vendor to agree on the associated responsibilities, acceptance criteria, specific deliverables, and schedule.
3. ENSURE the resolution of all remaining concerns associated with the reload Design Technical Review are included in the work scope including any design/licensing basis changes required by significant design changes or new fuel design. Refer to NF-AA-101.
4. ENSURE all analysis inputs, assumptions, methods, outputs and acceptance criteria for the reload analysis are either specified or accepted prior to the start of work. (CM-2)

NF-AA-100 Revision 13 Page 22 of 54 4.8.3. If required, then ENSURE a 50.59 Review is performed for any identified changes, such as methodology changes, product line updates, shipping requirements and manufacturing changes, which support the Core Reload FCP. (Refer to LS-AA-104)

(Core Designer) 4.8.4. INITIATE any licensing actions required for the reload, for instance changes to the unit Technical Specifications. (Licensing and Regulatory Affairs)

1. INITIATE licensing actions as soon as practicable to ensure sufficient lead-time for regulatory submittal and review.
2. Where licensing changes or threats are identified later in the reload process, promptly NOTIFY all stakeholders so that action plans can be developed.

4.8.5. MANAGE completion of reload analysis and licensing support. (Core Designer)

1. COORDINATE the scheduling, monitoring, completion and review of all reload support work.
2. ENSURE that all identified reload design and licensing scope is being addressed by scheduled work activities.
3. ENSURE that incomplete analyses are expected to be resolved in a timely manner, or have sufficient contingencies.
4. ADAPT the reload support work scope and schedule, as agreed with key stakeholders, to account for any analysis problems.
5. COORDINATE with site to ensure that the outage work program includes all fuel-related activities.
6. PROVIDE reload core design data to the Operations Training Department for simulator testing. Refer to TQ-AA-303. The ability to supply this data may be impacted by the actions identified in Step 4.3.6.3.
7. ENSURE that all source documentation is produced by appropriate parties (fuel vendor, contractors, NF, etc.) in a timely manner, reviewed, then distributed to appropriate recipients at site.

A. DOCUMENT the review and acceptance of fuel vendor and contractor documentation in the Core Reload FCP.

8. DOCUMENT the applicable design analyses, UFSAR updates, ADL changes, etc. in the Core Reload FCP.

NF-AA-100 Revision 13 Page 23 of 54

9. If significant changes are made to the inputs, assumptions, or methods used in reload analysis and/or licensing, WRITE an IR to document the changes and DISCUSS the changes during future Reload Kickoff meetings, per Attachment 5.

4.8.6. PROVIDE details of any "new" or "non-standard" testing requirements and acceptance test criteria for the reload in the FCP. (Core Designer)

1. Promptly NOTIFY Reactor Engineering of the "new" or "non-standard" testing once its existence becomes known.

4.8.7. PREPARE the reload COLR, based on the results collated at Step 4.8.5. (Core Designer) 4.8.8. PREPARE the reload IOCFR50.59 screening/ evaluation per LS-AA-104. (Core Designer) 4.8.9. PROVIDE information to Site Training Department to support development of training materials for the Core Reload design. (Core Designer)

1. SOLICIT Reactor Engineering and Training input on content of information.

Items to be considered are listed in Attachment 3.

2. PROVIDE information early enough to support completion of personnel training before aspects of the new reload affect duties.
3. IDENTIFY continuing training needs during the cycle if the design results in significant changes during varying stages of core life.
4. DOCUMENT any "new" or "non-standard" training requirements in the Core Reload FCP.

NOTE: A Core Loading Plan consistent with the expected Core Reload design may be provided earlier in the reload process to allow early development of the fuel shuffle sequence. However, the final version used to load and verify the as-loaded core must be consistent with the final core following any redesign.

4.8.10. PROVIDE a full-core Core Loading Plan to Reactor Engineering that indicates core-loading locations of all the reload cycle's assemblies/bundles using their vendor-supplied identifiers. (Core Designer)

1. PROVIDE core locations of affected control rod assemblies/control blades, burnable poison assemblies and neutron sources using their vendor-supplied identifiers.

NF-AA-100 Revision 13 Page 24 of 54

2. OBTAIN documentation from the fuel vendor detailing the as-built characteristics of the reload fuel versus the new assembly/bundle identifiers.

A. PROVIDE this information, as necessary, to appropriate personnel to update SNM and fuel movement software.

3. ENSURE the Core Loading Plan is consistent with the final core design.
4. VERIFY that the Core Loading Plan satisfies any optimization targets specified (e.g., at Step 4.3.8), for instance on as-built new fuel variations; component lifetime management; or fuel move optimization,
5. REVIEW the Core Loading Plan and as-built data prior to implementation.
6. TRANSMIT the Core Loading Plan via a TODI. Refer to CC-AA-310.

4.8.11. PROVIDE the following information to the NFDB analyst:

For BWRs, nominal enrichment for each reload bundle type For PWRs, nominal enrichment for the high enrichment central region of the fuel, excluding axial blankets, for each reload sub-batch. Include U-234 and U-236 enrichment, if applicable.

List of serial numbers for fuel to be discharged List of serial numbers for fuel to be re-inserted from the spent fuel pool, if applicable Rod type map(s) when introducing a new fuel type into the unit, if applicable.

(e.g., GE14 to GNF2, LTAs, etc.)

Rod type maps for all reload sub-batches containing IFBA rods 4.8.12. PROVIDE the computer files required by the core shuffle simulator (e.g.,

SHUFFLEWORKS) to Reactor Engineering. (Core Designer) 4.8.13. PROVIDE the FCP review package, reload 50.59 documentation, COLR, and vendor documentation summarizing the acceptability of the reload design to the appropriate members of the RDOT and CONVENE the RDOT, if necessary. (Core Designer) 4.8,14. PROVIDE the affected departments (typically the RDOT members) with an FCP review package for use in performing the interfacing review of, and input to the FCP.

(Core Designer)

1. OBTAIN affected Inter-Department reviews of the FCP. Refer to CC-AA-102 Attachment 9 and 10 Review Checklists. At a minimum, impact reviews should be obtained from all "Required" RDOT members (see Attachment 1) in

NF-AA-100 Revision 13 Page 25 of 54 departments other than Nuclear Fuels, as well as from departments that may be affected by unique items in a given reload.

2. IDENTIFY the procedure changes, deletions, and additions. In Passport, enter affected procedures on the ADL. In PIMS, list the affected procedures in the ECR text or Attachments. If a procedure is required to perform calibration or testing for the FCP, but is only partially implemented, then ASSURE that the procedure changes required are separate or interim procedure revisions. Refer to CC-AA-102, Attachment 9.

IDENTIFY 3. training changes, deletions, and additions required by Operations in the Special Instructions, and FORWARD to the Training department to support identification and completion of training requirements. Refer to CC-AA-102, Attachment 9.

4.8.15. After the receipt of the CC-AA-102 Attachment 9 and 10 Review Checklists, ASSEMBLE the contents of the FCP, SIGN the FCP and PROVIDE it to the assigned reviewer. (Core Designer) 4.8.16. REVIEW and APPROVE the Core Reload FCP. Refer to NF-AA-100-1000. This initial revision of the FCP will be the basis of the fuel load and approval is required prior to reloading the core for the reload cycle. (NF Engineer and Manager, BWRIPWR Design) 4.8.17. PROCESS the FCP package. (Core Designer)

For PIMS 1. ECRs, TRANSMIT the Approved FCP, including FCP Attachments, to the appropriate site Records Management.

2. For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-102 Attachment forms, PROVIDE a copy to the appropriate Site EC Coordinator for review, and RETAIN for future transmittal to Records Management. Only the final FCP package including all revisions is transmitted to Records Management.

4.8.18. CONFIRM that required activities are done, then BRIEF the Operations Representative on the Core Reload FCP scope and status of the package with respect to its readiness for Ops Acceptance. Required activities include the completion of Work Orders related to the FCP revision, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance. (Reactor Engineer and Site EC Coordinator)

1. In Passport, CHANGE the status of the FCP to MODIFIED.

2.

Do not CHANGE the status of the Cycle Management FCP to CLOSED -

Passport, COMPLT - PIMS.

NF-AA-100 Revision 13 Page 26 of 54 4.8.19. PROCESS the Core Reload COLR (Refer to AD-AA-102, Station Qualified Reviewer, and NF-AB-120-3600, COLR Generation - BWR Units) and 50.59 documentation (Refer to LS-AA-104), as required. The Core Designer should ensure the following actions are completed.

1. TRANSMIT the COLR to Station Reactor Engineering (Core Designer)
2. PERFORM SQR review of the COLR (Station Reactor Engineering)
3. APPROVE the COLR (SEAM)
4. TRANSMIT the COLR to Station Document Services (Station Reactor Engineering)
5. DISTRIBUTE the COLR to appropriate station personnel upon Station Reactor Engineering notification of issuance prior to startup (Station Document Services)
6. TRANSMIT the COLR to the NRC (Station Regulatory Assurance) 4.8.20. ENSURE all appropriate analyses/activities supporting the fuel load are complete prior to the loading of fuel with significant design changes or new fuel designs.

(Core Designer)

1. If Refueling and Core Verification activities are required prior to the Reload Licensing and/or Core Loading Plan approval in the Core Reload FCP, then PROCESS an Advance Work Authorization (AWA) prior to the activities.

Refer to NF-AA-1 00-1000.

4.9. Refueling and Core Verification 4.9.1. PERFORM all fuel-specific work items, as agreed in the outage work program or necessitated by emergent issues, according to identified procedures. (Site)

1. PREPARE procedures for fuel activities including offload, repair, movement, reload, component shuffling, etc. to assure that Fuel Reliability, Reactivity Management, vendor handling and SNM tracking requirements are met.
2. Procedures shall ASSURE that no changes to fuel or core configurations shall be performed until appropriate prerequisites are satisfied.

4.9.2. INITIATE any surveillance required to identify fuel failures or diagnose other fuel performance issues (e.g., foreign material, cladding performance, spacer grid damage, channel bow, control rod/blade condition, etc.) in re-use, re-insert or other offloaded fuel. (Site)

NF-AA-100 Revision 13 Page 27 of 54 PERFORM additional surveillance to either justify the re-insertion of particular (re-use or re-insert) fuel, or support company assessments on the suitability of the fuel design, as required.

4.9.3. LOAD the core according to the final Core Loading Plan (including core component loading requirements). (Site andlor Outage Services)

1. PERFORM an independent visual verification of the loaded locations/orientations and correct physical seating of all fuel and control components as prescribed by fuel handling/SNM procedures.

4.9.4. INITIATE a revision to the Core Reload FCP to incorporate final configuration changes (Cycle Specific Physics Data, Core Monitoring Database, Cycle Management Report, etc). (Core Designer)

NOTE: Prioritization of objectives for a redesign and significant changes resulting from a redesign shall be subject to the approvals consistent with the earlier steps of this Procedure. The amount of re-work shall depend on the perturbation from the FFCD, which should be minimized. If possible, redesigns should be done such that the initial FFCD licensing analyses remain bounding.

4.9.5. If a core redesign is necessary due to unanticipated reasons (e.g., a large shift in the outage schedule, fuel failure/damage, etc.), then INITIATE redesign activities.

(Core Designer)

1. DEVELOP a redesign plan and schedule.
2. ENSURE that the implemented reload design satisfies the requirements of this Procedure and other applicable procedures, prior to releasing redesign information to site.
3. Following the completion of all prior Steps of the redesign work, DOCUMENT the changes in the Core Reload FCP and REPEAT the appropriate Steps of this procedure affected by the redesign as a consistent set.

4.9.6. Prior to start- up, REVIEW the applicability of all assumptions and analyses on which the acceptability of the reload core is based. (Core Designer and Reactor Engineer)

1. CHECK both that the plant state is as assumed in the reload analyses and that the as-loaded core is consistent with all analyses.
2. VERIFY all activities required by the CDIR or by NF-AA-101, if required, have been completed.

NF-AA-100 Revision 13 Page 28 of 54

3. CONVENE review meetings as necessary to complete this step.
4. If a redesign has taken place, then ASSURE that all necessary steps of this procedure have been repeated.
5. CONFIRM that the 10CFR50.59 screening/evaluation has been completed, and that reload information has been presented to and reviewed by the site PORC and NSRB as required by applicable procedures. Refer to LS-AA-1 04 and LS-AA-106.
6. REVIEW the Risk Management Assessment, and MODIFY if necessary. The RDOT, RRB, and SMDI shall approve of significant changes to the Risk Management Assessment as directed by NF-AA-1 00-1600.

4.9.7. PROVIDE the departments affected by the FCP Revision with an FCP review package for use in performing the interfacing review of, and input to the FCP Revision. (Core Designer)

1. OBTAIN affected Inter-Department reviews of the FCP by utilizing the CC-AA-102 Attachment 10 Review Checklists. At a minimum, impact reviews should be obtained from Reactor Engineering and Operations.

4.9.8. After the receipt of the CC-AA-1 02 Attachment 10 Review Checklists, ASSEMBLE the contents of the FCP, SIGN the FCP and PROVIDE it to the assigned reviewer.

(Core Designer) 4.9.9. REVIEW and APPROVE the Core Reload FCP Revision. Refer to NF-AA-1 00-1000. This revision is intended to be the final revision of the FCP and approval is required prior to cycle start- up. (NF Engineer and Manager, BWRIPWR Design) 4.9.10. PROCESS the FCP package. (Core Designer)

1. For PIMS ECRs, TRANSMIT the Approved FCP Revision, including FCP Attachments, to the appropriate site Records Management.
2. For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-1 02 Attachment forms, for this revision, PROVIDE a copy to the appropriate Site EC Coordinator for review, and PROVIDE the final FCP package including printouts of all revisions (i.e.,

dispositions, Passport screen printouts, FCP Attachments, CC-AA-1 02 Attachments, etc. for each revision) to the Site EC Coordinator for processing.

4.9.11. VERIFY that all necessary documentation has been supplied and that updates are prepared, ready for start-up. (Site) This includes:

All site manuals and procedures (including TS and COLR)

NF-AA-100 Revision 13 Page 29 of 54 Incorporation of cycle-specific physics data into site procedures or databook Incorporation of reload design acceptance criteria into site startup and core management procedures The Core Monitoring System Database The training simulator and other training materials Design Basis Database (e.g., ATLAS, DBdb, etc.)

4.9.12. IMPLEMENT the updates of site material (documentation, procedures, computer databases, etc.) including updates necessary for any redesign impact. (Site) 4.9.13. PROVIDE all necessary pre-startup information to USNRC. (Licensing and Regulatory Affairs) 4.9.14. CONFIRM that required activities are done, then BRIEF the Operations Representative on the Core Reload FCP scope, completion, and final status of the package with respect to its readiness for Ops Acceptance. Required activities include the completion of Work Orders related to the FCP, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance. (Reactor Engineer and Site EC Coordinator) 4.9.15. If Operations has been satisfied that the Core Reload FCP scope is acceptable, then CHANGE the status of the Core Reload FCP to reflect that all changes and associated fuel moves are complete (CLOSED -- Passport, COMPLT - RIMS). (Core Designer or Site EC Coordinator) 4.9.16. PERFORM a Post-Job Brief for the Core Reload after the start-up of the refueled unit. Refer to HU-AA- 1212. (Core Designer and Manager, BWRIPWR Design)

1. ENSURE participation by appropriate RDOT members and the vendor.

4.10. Core Operation and Monitoring NOTE: When significant changes to the core design have been made, reload-specific tests or surveillances may be specified.

4.10.1. PERFORM any BOC tests or surveillances required to demonstrate that the Core Reload behaves as predicted (and therefore that the reload analyses are applicable to the as-loaded core). (Site)

NF-AA-100 Revision 13 Page 30 of 54

1. If any test or surveillance criteria are violated, then REFER to the applicable procedures, which shall clearly prescribe the actions to be taken and define whether continued operation is acceptable.

4.10.2. PROVIDE all necessary post-start-up reload documentation to the USNRC, for example: COLR (if not provided at Step 4.9.13); start-up report, 10CFR50.46 documentation, etc. (Licensing and Regulatory Affairs) 4.10.3. If required by station procedures, then INITIATE an independent offsite review of the reload design and 10CFR50.59 screening/evaluation following the start-up of the cycle. (NF) 4.10.4. OPERATE the core according to procedures that ensure that NF core management guidance, required to preserve analysis assumptions or to support fuel performance objectives, is followed. (Site) (CM-11)

1. MONITOR the core performance against defined acceptance criteria.

Monitored information should include but not be limited to: (Site and/or NF)

Power distributions versus prediction and evaluation limits

- Core reactivity versus prediction and evaluation limits

- Fuel exposure versus prediction and evaluation limits

- Margin to thermal limits (e.g., LHGR, APLHGR, MCPR, etc.) and operational limits (setpoints, rod limits, etc.)

- Accuracy of maneuver predictions

- Coolant chemistry control

- Fuel integrity/coolant activity

- Reactivity management items

- Trends which erode margin to operating goals

- Anomalies

- Any additional cycle-specific items determined under this process

2. DOCUMENT core monitoring results according to unit procedures. (Reactor Engineering)

A. HIGHLIGHT all anomalous conditions and trends that threaten operating goals or safety margins and COMMUNICATE to key stakeholders. Refer to LS-AA-125.

3. DOCUMENT significant problems identified according to the corporate corrective action process. (Site and Core Designer)

NF-AA-100 Revision 13 Page 31 of 54 A. Where appropriate, SHARE anomaly data and experience with the fuel vendor and DISSEMINATE as industry Operating Experience (CM-16) 4.10.5. USE relevant results of the monitoring program as inputs to the following cycle Reload design. (Core Designer)

NOTE: It is not necessary to open a Cycle Management FCP for a given operating cycle unless a configuration change is to be incorporated.

4.10.6. DOCUMENT Cycle Management configuration changes (revisions to COLR, 50.59 screening/evaluation, Cycle Management Report, etc.) on a Cycle Management FCP. (Core Designer)

1. INITIATE an FCP type EC/ECR to control the Cycle Management configuration changes. Refer to NF-AA-100-1000. (Core Designer)
2. PERFORM a Technical Pre-Job Brief for the Cycle Management design effort. Refer to HU-AA-1212. (Core Designer and Manager, BWR/PWR Design)

A. DISCUSS NF, Inter-Department and potential Independent Third Party reviews. The Pre-Job Brief is expected to be a face -to-face discussion with the Manager, BWRIPWR Design, taking the appropriate questioning/challenging role.

3. INCLUDE applicable design considerations and impacts using the Design Attribute Review (DAR). Refer to CC-AA-1 02 Attachment 1. (Core Designer)

NOTE: The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per a regulatory processes such as 50.59, 50.90, 50.46, etc.

4. REVIEW the Risk Management Assessment for the cycle of interest and update as necessary based on the Cycle Management changes at hand.

(Core Designer)

CONSIDER in the assessment any changes to the vendor or Exelon methods that support core operation.

ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.

ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.

NF-AA-100 Revision 13 Page 32 of 54

- DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).

5. If issues arise during cycle operation which cause significant changes to the inputs, assumptions, or methods used in reload analysis and/or licensing, WRITE an IR to document the changes and DISCUSS the changes during future Reload Kickoff meetings, per Attachment 5. (Core Designer)

REVISE necessary documentation, as applicable, and PERFORM the appropriate steps defined in NF and other procedures for the Cycle Management changes at hand. (Core Designer)

7. PROVIDE the departments affected by the Cycle Management FCP Revision with an FCP review package for use in performing the interfacing review of, and input to the FCP Revision. (Core Designer)

A. IDENTIFY the procedure changes, deletions, and additions. In Passport, enter affected procedures on the ADL. In PIMS, list the affected procedures in the ECR text or Attachments. Refer to CC-AA-102, Attachment 9.

B. IDENTIFY training changes, deletions, and additions required by Operations in the Special Instructions, and FORWARD to the Training department to support identification and completion of training requirements. Refer to CC-AA-1 02, Attachment 9.

C. OBTAIN affected Inter-Department reviews of the Cycle Management FCP. Refer to CC-AA-102 Attachment 10 Review Checklists. At a minimum, impact reviews should be obtained from Reactor Engineering and Operations.

8. After the receipt of the CC-AA-1 02 Attachment 9 and 10 Review Checklists, ASSEMBLE the contents of the Cycle Management FCP, SIGN the FCP and PROVIDE it to the assigned reviewer. (Core Designer)
9. REVIEW and APPROVE the Cycle Management FCP. Refer to NF-AA-1 00 -

1000. The FCP will be the basis for the current Cycle Management configuration changes. (NF Engineer and Manager, BWRIPWR Design)

10. PROCESS the FCP package. (Core Designer)

A. For PIMS ECRs, TRANSMIT the Approved FCP, including FCP Attachments, to the appropriate site Records Management.

B. For Passport ECs, PRINT the Approved FCP, including FCP Attachments, Passport screens, and electronic CC-AA-102 Attachment forms, PROVIDE a copy to the appropriate Site EC Coordinator for

NF-AA-100 Revision 13 Page 33 of 54 review, and RETAIN for transmittal of the final FCP package to Records Management.

C. For Passport ECs, TRANSMIT the final FCP package including printouts of all revisions to the appropriate Site EC Coordinator for processing.

NOTE: Cycle Management FCPs can consist of documentation changes (e.g., core model changes, 50.46 reports, UFSAR changes), operating procedure changes (e.g., COLR changes) and physical changes to the plant (e.g., mid-cycle re-design, core monitoring system software/database change).

Ops Acceptance of the Cycle Management FCP is only required when operating procedure changes or physical changes to the plant are made. Required activities include the completion of Work Orders related to the FCP revision, FCP sign-offs, and issuance of procedures and documents on the ADL that are required prior to Ops Acceptance.

11. If Ops Acceptance is required for the FCP revision, CONFIRM that required activities are done, then BRIEF the Operations Representative on the Cycle Management FCP scope and status of the package with respect to its readiness for Ops Acceptance. (Reactor Engineer and Site EC Coordinator)

A. In Passport, CHANGE the status of the FCP to MODIFIED.

B. Do not CHANGE the status of the Cycle Management FCP to CLOSED - Passport, COMPLT - PIMS.

12. PERFORM a Post-Job Brief for the Cycle Management changes after the changes have been implemented. Refer to HU-AA-1212. (Core Designer and Manager, BWRIPWR Design)
13. REVISE the Cycle Management FCP, as required, for subsequent changes throughout the balance of the cycle. (Core Designer)
14. At the end of the operating cycle, CHANGE the status of the Cycle Management FCP to reflect that all Cycle Management configuration changes are complete (CLOSED - Passport, COMPLT - PIMS). Ops Acceptance is not required to close out the Cycle Management FCP at the end of the operating cycle. (Core Designer or Site EC Coordinator)

NF-AA-100 Revision 13 Page 34 of 54 4.11. FCP Revisions 4.11.1. INITIATE a revision to the FCP in Passport or PIMS, as appropriate. (Core Designer)

NOTE: NF Director authorization is required for an Advanced Work Authorization (AWA) when the functional design, design criteria of the plant, or the 10CFR50.59 Evaluation are affected by the FCP revision. An NF Manager can authorize other AWAs.

4.11.2. If Advanced Work Authorization (AWA) to proceed at risk with installation work activities in the field without issuance of a completely approved configuration change is required and the responsible NF Manager is sure that the advance work does not affect any in-service equipment and the equipment will not be placed in service and will not be declared operable, and will have no impact on operating or in-service equipment until the configuration change is approved, then: (Core Designer)

1. DOCUMENT the following information in the Revision Summary area of the FCP (Topic Notes Panel - Passport, Page 2 - PIMS)

A statement that an AWA is being used Scope of Authorized Work, including markups as needed Justification the advance work does not affect any in-service equipment and the equipment will not be placed in service and will not be declared operable, and will have no impact on operating or in -

service equipment until the configuration change is approved Special Instructions Name of Core Designer that is giving the AWA Name of the Recipient (Reactor Engineer, etc.) of the AWA Name of NF Management personnel that is authorizing the AWA Date of the AWA

2. INDICATE in the FCP that AWA is going to be used:

A. ADD an AWA Milestone for Passport or ADD a Y in the AWA field at the top of Page 2 for PIMS.

NF-AA-100 Revision 13 Page 35 of 54 4.11.3. SUMMARIZE revision changes in the "REVISION

SUMMARY

". (Core Designer)

1. IDENTIFY any affected documents (i.e., documents on the ADL) in the "REVISION

SUMMARY

" for the respective revision in which they were changed.

4.11.4. REVISE the portions of the FCP that are affected, including disposition notes, package attributes (from use of CC-AA-102), package milestones, ADL, etc. as needed to reflect the change. (Core Designer)

1. DISCUSS change with the interfacing departments that may be affected. The DAR (CC-AA-102 Attachment 1) is used to determine which groups are affected.

4.11.5. If affected, then DOCUMENT the review and acceptability of the revision by providing CC-AA-102 Attachment 10 Forms or electronically signing the FCP revision. (Affected Interfacing Departments)

NOTE: The risk management assessment process is to be used to identify and mitigate significant risks to the reload design or cycle operation processes. It is not in any way meant to replace the review of a change per a regulatory processes such as 50.59, 50.90, 50.46, etc.

4.11.6. REVIEW the Risk Management Assessment for the cycle of interest and update as necessary based on the changes at hand. (Core Designer)

- CONSIDER in the assessment any changes to the vendor or Exelon methods that support core operation.

- ENSURE that sufficient relevant industry and station data exists to mitigate risk associated with use of the new method.

ENSURE any data that is to be used for risk mitigation has been appropriately validated; otherwise, do not UTILIZE the data.

- DOCUMENT risks associated with method changes supporting core operation in the risk management assessment and provide the justification for use of the new method (i.e. proper data and validation).

4.11.7. REVIEW, APPROVE and PROCESS the FCP in accordance with appropriate sections above for the revision scope. This may require different reviews than those required in previous FCP revisions. (Core Designer)

NF-AA-100 Revision 13 Page 36 of 54 4.12. Reload And Cycle Management Decision Making NOTE: The Nuclear Fuels Department many times participates in decision making processes that have been initiated by outside sources, especially the nuclear stations. When this is the case, Step 4.12.1 may be credited as complete via NF participation in the larger process.

4.12.1. APPLY the appropriate decision making process when a technical decision that impacts plant operation is to be made during the reload or cycle management processes.

1 USE a graded approach to decision making that is dependent on the decision's: (Core Designer / Manager, BWRIPWR Design)

Risk - possible consequences to the plant or company Number and Level of the stakeholders Distribution of departments represented by the stakeholders (e.g.

station operations vs. corporate accounting)

2. SCREEN -OUT lower level decisions which should not require a formal decision making process to ensure resources are used as optimally as possible. (Core Designer / Manager, BWRIPWR Design)
3. When available, UTILIZE an existing decision making process such as (but not limited to): (Core Designer)

- OTDM - Operational technical decision making process as documented in OP-AA-106-101-1006.

- A technical evaluation as documented in accordance with CC-AA-309-101.

- The Kepner-Tregoe problem solving and decision making process (trained facilitators and individual users available within Exelon).

4.12.2. DOCUMENT the use of all formal decision making processes in the reload and cycle management processes in the associated FCP(s) This does not include those items screened-out of the process in Step 4.12.1.2 - though they may need to be included in the FCP per a different procedural requirement. (Core Designer) 4.12.3. ENSURE that all decisions made and documented in the reload and cycle management FCPs have received the proper level of communication and review commensurate with the importance of the decision. (Manager, BWR/PWR Design) 4.12.4. REVIEW all technical decisions made using one of the formal methods above for possible inclusion in the proper reload and/or cycle management risk management assessment(s) per NF-AA-1 00-1600. (Core Designer)

NF-AA-100 Revision 13 Page 37 of 54

5. DOCUMENTATION NOTE: Detailed documentation requirements for documents that are attached to or referenced by Core Reload and Cycle Management FCPs are contained in lower level procedures and T&RMs.

5.1. TRANSMIT the Core Reload and Cycle Management FCP EC/ECR to Records Management. (Reference RM-AA-101) SRRS# 3B. 107 (Core Designer)

6. REFERENCES 6.1. Writer References 6.1.1. 1 OCFR50 Appendix B, Criterion III, Design Control 6.1.2. ASME NQA-1, Quality Assurance Program Requirements for Nuclear Facilities 6.1.3. INPO 90-009 (Guideline), March 1990, Guidelines for the Conduct of Design Engineering 6.1.4. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores 6.1.5. INPO SOER 03-02, Managing Core. Design Changes 6.1.6. NF-AA-1, Exelon Nuclear Policy, Fuel Management 6.1.7. NF-AA-3, Exelon Nuclear Policy, Fuel Reliability 6.1.8. NF-AA-10, Reload Control Process Description 6.2. User References 6.2.1. AD-AA-102, Station Qualified Review 6.2.2. CC-AA-102, Design Input and Configuration Change Impact Screening 6.2.3. CC-AA-103, Configuration Change Control for Permanent Physical Plant Changes 6.2.4. CC-AA-309, Control of Design Analyses 6.2.5. CC-AA-310, Transmittal of Design Information 6.2.6. HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Brief 6.2.7. LS-AA-104, Exelon 50.59 Review Process

NF-AA-100 Revision 13 Page 38 of 54 6.2.8. LS-AA-106, Plant Operations Review Committee 6.2.9. LS-AA-125, Corrective Action Program (CAP) Procedure 6.2.10. NF-AA-101, Nuclear Fuel Assembly and Core Component Design and Fabrication Process Changes 6.2.11. NF-AA-100-1000, Core Reload and Cycle Management Configuration Changes 6.2.12. NF-AA-100-1010, Use of Operating Experience Information for Nuclear Fuels Work Products 6.2.13. NF-AA-100-1600, Reload and Cycle Operation Risk Management Assessment Instructions 6.2.14. NF-AA-100-2000, Standard RRB and SMDI Presentation Templates 6.2.15. NF-AA-105-1000, Energy Utilization Plan Development 6.2.16. NF-AB-100-4000, GNF Reload Control Implementation 6.2.17. NF-AB-100-5000, Westinghouse Reload Control Implementation for BWRs 6.2.18. NF-AB-100-6000, Framatome Advanced Nuclear Fuel (FANP) Reload Control For BWRs 6.2.19. NF-AB-120-3600, Core Operating Limits Report Generation - BWR Units 6.2.20. NF-AP-100-7000, Westinghouse NSSS Reload Design Control Implementation 6.2.21. NF-AP-100-8000, AREVA PWR Reload Control Implementation 6.2.22. OU-AA-101, Refuel Outage Management 6.2.23. RM-AA-101, Records Management Program 6.2.24. TQ-AA-303, Controlling Simulator Core Updates and Thermal-Hydraulic Model Updates 6.3. Station Commitments 6.3.1. These commitments apply to all Exelon nuclear generating stations

1. CM-1 INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03980, Passport 00353329-01) (Step 4.2.1, Attachment 3)
2. CM-2 INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03977, Passport 00353329-11) (Step 4.5.2, 4.8.2)

NF-AA-100 Revision 13 Page 39 of 54 CM-3

3. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03976, Passport 00353329-10) (Step 4.3.8)

CM-4

4. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03975, Passport 00353329-09) (Step 4.5.4, Attachment 3)

CM-5

5. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03972, Passport 00353329-07) (Attachment 3)

CM-6

6. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03971, Passport 00353329-06) (Attachment 3)
7. CM-7 INPO SOER 96-02, Design and Operating Considerations for

. Reactor Cores (PIMS T03970, Passport 00353329-05) (Attachment 3)

CM-8

8. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03969, Passport 00353329-04) (Attachment 3)

CM-9

9. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03964, Passport 00353329-03) (Step 4.5.3, Attachment 2)

CM-10

10. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03973, Passport 00353329-08) (Attachment 3)

CM-11

11. INPO SOER 96-02, Design and Operating Considerations for Reactor Cores (PIMS T03981, Passport 00353329-02) (Step 4.10.4)
12. CM-12 INPO SOER 03-02, Managing Core Design Changes (PIMS T04588, Passport 00353329-12) (Step 4.3.9, 4.4.1)
13. CM-13 INPO SOER 03-02, Managing Core Design Changes (PIMS T04590, Passport 00353329-14) (Attachment 1 and Attachment 3)
14. CM-14 INPO SOER 03-02, Managing Core Design Changes (PIMS T04591, Passport 00353329-15) (Step 4.7.4 and Attachment 3)
15. CM-15 INPO SOER 03-02, Managing Core Design Changes (PIMS T04592, Passport 00353329-16) (Attachment 3)
16. CM-16 INPO SOER 03-02, Managing Core Design Changes (PIMS T04589, Passport 00353329-13) (Step 4.10.4)

NF-AA-100 Revision 13 Page 40 of 54

7. ATTACHMENTS 7.1. Attachment 1 - Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities 7.2. Attachment 2 - Cycle Design Inputs and Requirements 7.3. Attachment 3 - Reload Design Review Criteria 7.4. Attachment 4 - Conduct of Reload Review Board and SMDI Meetings 7.5. Attachment 5 - Conduct of Reload Design Kickoff Meetings And Team Responsibilities

NF-AA-100 Revision 13 Page 41 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities Page 1 of 3 Reload Design Overview Team (RDOT)

The RDOT is a multi-discipline team made up of NF and site personnel that provides planning, input, and review of reload activities. It assures both that appropriate boundaries for the reload design are established, and that any changes to established boundaries required by the design are acceptable.

RDOT members consist of one representative from each of the following departments.

Attendance by these members is required for the RDOT meetings in which the CDIR is defined, and the Design Technical Review of the reload is performed. Additional required attendees are identified on a case-by-case basis, considering functional areas that may be affected by significant changes proposed for a specific reload. Depending on the agenda, supplemental RDOT meetings may not require full RDOT attendance. The Core Designer and PWR/BWR Design Manager are responsible for making this determination.

- NF (the unit Core Designer, who shall chair) [required]

- Fuel Reliability Engineer [required]

- Engineering Safety Analysis [required]

- Reactor Engineering (site) [required]

- Chemistry (site) (CM-13) [required]

- Radiation Protection (site) [required]

- Operations (site) [required]

Training (site) [required]

Other contributors (e.g., Plant Engineering, Plant Maintenance, Licensing and Regulatory Affairs, Outage Management, Outage Services, Emergency Procedure Guidelines owner, vendor representative) shall participate as required.

NF-AA-100 Revision 13 Page 42 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities Page 2 of 3 The conduct of RDOT meetings is as follows:

1. ORGANIZE RDOT meetings required by this procedure. (Core Designer)
2. PREPARE the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers)
3. SCHEDULE RDOT meetings required by this procedure and ENSURE adequate representation at each RDOT meeting commensurate with the items to be discussed.

(RDOT Chair)

a. IDENTIFY required attendees in meeting invitation.
b. IDENTIFY optional attendees in meeting invitation.
c. ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
4. CONVENE RDOT meetings and PRESENT prepared information. (Core Designer and other responsible NF engineers)
a. If meeting is cancelled prior to scheduled meeting time then RESCHEDULE (Core Designer)
b. If meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (RDOT Chairman)

- If meeting is cancelled then RESCHEDULE. (Core Designer)

- If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s) and SOLICIT comments (Core Designer).

- WRITE an Issue Report documenting inadequate attendance (RDOT members not in attendance, RDOT Chairman or designee).

5. DOCUMENT required RDOT concurrence in minutes to the RDOT meetings. (Core Designer)
a. If concurrence is not obtained, then ESCALATE differences of opinion to Senior Managers from Nuclear Fuels and other areas, as applicable. (Core Designer)
6. DOCUMENT action items in minutes to the RDOT meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
7. DOCUMENT meeting attendance in minutes to the RDOT meetings. (Core Designer)

NF-AA-100 Revision 13 Page 43 of 54 ATTACHMENT 1 Conduct of Reload Design Overview Team (RDOT) Meetings And Team Responsibilities Page 3 of 3 The RDOT shall:

EVALUATE 1. and AGREE upon the Cycle Design Inputs and Requirements (as defined by this and other guidelines/procedures);

ASSURE 2. that all design inputs are approved by both NF and unit representatives before they are used in definitive analyses; SUPPORT NF 3. technical reviews of proposed fuel and core designs; ASSURE the4.completeness of the work scope scheduled to support reload design implementation; AGREE 5. on the reload design and licensing schedule and responsibilities; EVALUATE 6.changes to, or problems with, the reload design and/or licensing schedule, and DETERMINE corrective actions; CONFIRM 7. the applicability of the reload analyses prior to criticality; MEET 8. as necessary to fulfill the above goals.

To support the above requirements, each RDOT representative shall:

1. PARTICIPATE as required in RDOT meetings or activities; ENSURE 2. that the Cycle Design Inputs and Requirements reflect all relevant plant and procedure changes in their functional area; REVIEW 3. historic and current, unit and industry, operational experience and assure that applicable experience is considered; PROVIDE 4. or CONCUR with inputs for design and licensing analyses; REVIEW 5. proposed fuel and reload designs for potential impacts related to their functional area; 6.

IDENTIFY changes to programs, procedures, UFSAR or training in their functional area necessitated by the reload design; 7.

IDENTIFY performance monitoring needs relating to fuel, core, or operating strategies for which the station has little experience.

NF-AA-100 Revision 13 Page 44 of 54 ATTACHMENT 2 Cycle Design Inputs and Requirements Page 1 of I The CDIR should assure clear and consistent understanding between the company, fuel vendor and other organizations performing reload work. The items to be documented may include:

1. Cycle design inputs such as: the EUP; the reload fuel design; the cycle operating regime; estimated previous cycle EOC; (CM-9)
2. Other cycle design objectives such as component lifetimes; fuel utilization targets; reload batch size; fuel shuffle constraints; specific use of particular fuel or components (reinserts; LTAs);
3. Constraints or objectives for the BOC and EOC refueling outages;
4. Vendor fuel design limits such as maximum enrichment, burnup capability; incore residence; maneuvering rates; chemistry; LHGR;
5. Safety analysis inputs such as RCS characteristics, reactivity parameters, boron concentrations; etc.;
6. Limiting values for safety parameters, e.g., moderator temperature coefficients; peaking factors; shutdown margin; source term;
7. Bounding operating envelopes and/or conditions;
8. Other plant characteristics (and particularly changes) which affect the assumed bases for the reload design;
9. Requirements for particular analyses or methodologies;
10. Targets for ease/flexibility of operation; margin to setpoints; etc.;
11. Relevant inputs from: dialogue with the fuel vendor or contractors; Operating Experience; unit core monitoring programs;
12. An initial schedule of reload work, deliverables and responsibilities

NF-AA-100 Revision 13 Page 45 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 1 of 3 The following list of items should be considered when developing specific Reload Design Review Criteria for lower level procedures or guidelines.

Evaluations of new core designs and operating strategies should include (but are not limited to) the considerations tabulated below. The effects of each on: limiting safety analysis envelopes, actual steady-state plant operation and `routine' maneuvers and transients should all be considered.

1 Operational Strategy. Will the operational strategy be different to previous cycles? Are changes to o eratin rocedures required?

2 Operational considerations:

- Can the core be operated at the intended power level for the full cycle with sufficient flexibility and maneuverability?

- Is there any change in margin to heat rate and other safety limits? Will this significantly affect required operator actions in steady-state or transient operation?

- Will core response or controllability be affected (changes in rod worth, reactor period, temperature coefficients of reactivity, shutdown margin, potential for xenon oscillations, power oscillations or instability, shutdown boron requirements, etc.)?

- Equipment Out-of-Service (EOOS). Is the scope of the licensed EOOS analyses sufficient to address current plant operational issues? Are additional EOOS analyses required/warranted?

3 Reactivity Management. Are the reactivity control mechanisms during fuel handling or for the loaded core different from previous cycles? Are there any changes to reactivity envelopes or control assumptions?

4 Core Monitoring. Could core monitoring and instrument indications be affected by significant changes in loading pattern, power distribution, core response, core pressure drop, shielding, etc.?

5 Changes in Key Parameters. Are significant changes expected at any time in cycle for:

power distribution; reactivity coefficients; reactivity worths; reactor response? Will this affect the values assumed in safety analyses? How will this be a pparent to the operators?

6 Does the cycle design si nificantl affect instrument response or calibration, etc.?

7 Does the cycle design affect other items of plant, e.g., flows or temperatures to Steam Generators, reactor vessel fluence, etc.? Have these been discussed with cognizant 1 en ineers? CM-4)

NF-AA-1 00 Revision 13 Page 46 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 2 of 3 Chemistry. Will coolant chemistry be affected by core design changes or will the chemistry regime be changed? Will planned chemistry changes affect core or fuel performance (e.g., zinc, hydrogen, noble metals, pH control, etc.)? Will known abnormal chemistry conditions affect core or fuel performance (e.g., condenser tube leaks, chemical intrusions, etc.)? Are empirical failed fuel indications expected to differ from p revious cycles? CM-13 4

9 ! Modeling. Does the core design involve novel characteristics that may affect modeling accuracy vs. Previous cycles? How will this be accounted for in the design and testing acceptance criteria? Should procedural guidance or training be provided to address any limitations associated with the predictive capability of core performance prediction tools?

Is an independent, backup code required or recommended? (CM-7) (CM-15 0 Methodologies. Are reload methodologies changing? Are new methodologies approved by NRC? What is the comparative experience between new and old methodologies?

Are increased uncertainties or extra conservatism required? (CM-1)

NOTE: A change in methodology may require a 50.59 Review (Refer to LS-AA-1 04 Start-Up. Are any special considerations or testing necessary for initial start-up, due to use of markedly different fuel or core desians?

12 Fuel, Fuel Related- and Core- Component Loading. Are fuel enrichments, loading pattern, poison loadings, burnable absorber design/loading, core component design/composition, etc. all within previous experience or proven envelopes?

13 Coolant Conditions. Will coolant flows and/or temperatures be affected by changes in fuel desi n, physical modifications to the RCS or changes in operating strategy 14 Vendor Precautions. Are any changes to the operating regime necessary to comply with vendor-provided cautions or limitations for the operation of fuel, fuel related- and core-components?

Reload. Does the reload batch size, fuel design, outage offload or reload sequence, etc.

place any new requirements on fuel storage, handling or monitoring, both in- and ex-core? (examples: heat load, criticality, shielding, storage capacity, etc) (CM 5) (CM-6 Cumulative Changes. Each new core should be evaluated to ensure that cumulative small changes over multiple cycles do not constitute a significant unreviewed change.

Examples: mix of fuel designs; erosion of operating margin; increase in poison concentrations; accuracy of core monitoringg.__

NF-AA-1 00 Revision 13 Page 47 of 54 ATTACHMENT 3 Reload Design Review Criteria Page 3 of 3 Core monitoring. Does the new cycle introduce any features which are worthy of 17 increased monitoring? Are changes to core monitoring procedures required? Are any new tests necessary?

Design Complexity. Does either the fuel cycle design or individual assembly designs 18 (loading variations, etc.) introduce unnecessary complexity to the manufacturing and/or reload process? Are increased vendor and/or reload QA measures warranted? (CM-14)-

NOTE: Acceptance criteria for simulator fidelity may be more limiting if the simulator is used to qualify operators on reactivity manipulations.

Training. Is operator training and/or a replica simulator update required? (CM-10) Does the reload increase PCI (pellet cladding interaction) vulnerability that requires additional 19 operator awareness training?

Is the training simulator impacted as a result of fuel design changes or changes in the fuel design codes? Changing fuel design codes may result in the inability to perform a cycle specific simulator core update because the simulator uses output from the design codes.

Plant Conditions. Have any changes in plant conditions invalidated the assumptions 1 20 made in the fuel c cle deli nand reload licensin ?

Procedures. Do changes associated with the new core design require changes to plant or fuel suDoort organization procedures?

LTAs. Does the cycle contain an LTA program? Have all necessary plant and licensing 22' issues regardinghe t LTAs been identified?

Cycle Flexibility/Contingency. Does the cycle support sufficient contingency for a change in operating strategy, increased cycle energy or a BOG redesign due to either: (indicated or potential) fuel failures; handling damage; or previous cycle under /overrun? Does the cycle support sensible options for future c Iles?

Cycle design objectives. Does the design meet the defined cycle design objectives and requirements. Is the balance between economics and operational flexibility appropriate?

Regulatory compliance. Are any changes to TS, UFSAR or other configuration or 25 licensing documentation required? Are changes to Design basis documents or databases needed?

Are there any mixed core issues to address? (CM-8)

Source Term. Is the source term used for radiological analyses still appropriate? Is an 27 update to the radiolo g ical anal sis re aired?

NF-AA-100 Revision 13 Page 48 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page I of 4 RELOAD REVIEW BOARD MEETINGS The Reload Review Board (RRB) is an internal Nuclear Fuels (NF) board that provides oversight and review of all Reload Designs. The RRB oversees the reload strategy, fuel cycle economics and business issues, impacts of design decisions, and communication of reload plans outside of NF. The RRB meets as required by this procedure to assure that each reload receives the appropriate level of oversight.

The RRB is made up of the following members, or their nominated delegates:

Vice President, NF (Chairperson)

- NF Director (Vice-Chair)

- Manager, BWR/PWR Design

- Manager, Engineering Safety Analysis

- Manager, Fuel Reliability and Projects

- Fuel Supply Director

- Spent Fuel Director Fuel Vendor Representative The conduct of RRB meetings is as follows:

NOTIFY RRB members and ORGANIZE RRB meetings required by this procedure. (Core Designer)

2. PREPARE and PRESENT the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers)
a. ENSURE that commercially sensitive data is not distributed to individuals outside of Exelon.
3. ENSURE adequate representation at each RRB meeting commensurate with the items to be discussed. (NF VP or designee)
4. DOCUMENT required RRB concurrence in minutes to the RRB meetings. (Core Designer)
5. DOCUMENT action items in minutes to the RRB meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
6. DOCUMENT meeting attendance in minutes to the RRB meetings. (Core Designer)

NF-AA-100 Revision 13 Page 49 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 2 of 4 Senior Management Design Initialization (SMDI) Meetings A meeting held early in the Reload Design process to ensure that senior site management actively participate in the decision making process for significant changes in Reload Design process, core operating strategy, fuel mechanical design, and fuel cost associated with the reload. The purpose of the meeting is to inform and gain concurrence from site senior management on the reload plans (including cycle operating capacity factor and operating margins) and design options before the reload design is fully initialized. A summary of the results of the most recent fuel vendor audits/assessments, as well as near term future audit/assessment goals, shall also be provided to the SMDI.

For major changes with broad impacts (e.g., the proposed use of a new fuel design), a Special SMDI meeting should be conducted to obtain written concurrence from site senior management at a time early enough to accommodate changes in strategy.

The following personnel, or their nominated delegates, are recommended for attendance at SMDI meetings. Attendees required for a quorum are identified below. Identification of additional required attendees shall be made on a case-by-case basis, depending on the functional areas that may be affected by significant changes proposed for a specific reload.

Optional attendees, while not required for a quorum, should be encouraged to attend.

Required SMDI Meeting Attendees

- NF Vice President (Chairperson)

- Site Vice President (Vice-Chair)

- Station Manager

- Engineering Director

- Operations Director

- Training Director

- Reactor Engineering Manager

- Site Chemistry Manager

- Site Radiation Protection Manager

-- NF Core Designer

--- Manager, Engineering Safety Analysis

N F-AA-1 00 Revision 13 Page 50 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 3 of 4 Optional SMDI Meeting Attendees Outage Director

- Sr Manager of Design Engineering Sr Manager of System Engineering

- Licensing and Regulatory Affairs Manager Other Site Managers (e.g., Engineering, Plant Maintenance, Health Physics, Outage Management)

- Reactor Services representative (particularly for fuel design changes or lead assemblies)

- NF Director NF representatives (Manager PWR/BWR Design, FRE, other individual contributors)

Fuel Vendor representative

- RDOT representatives

- Corporate Operations Fleet Outage Scheduler

NF-AA-100 Revision 13 Page 51 of 54 ATTACHMENT 4 Conduct of Reload Review Board and SMDI Meetings Page 4 of 4 The conduct of SMDI meetings is as follows:

1. ORGANIZE SMDI meetings required by this procedure. (Core Designer)
2. PREPARE the information required by applicable steps in this procedure. (Core Designer and other responsible NF engineers)
3. SCHEDULE SMDI meetings required by this procedure and ENSURE adequate representation at each SMDI meeting commensurate with the items to be discussed.

(Core Designer)

a. INVITE all required and optional members listed above.
b. IDENTIFY required attendees in meeting invitation.
c. IDENTIFY optional attendees in meeting invitation.
d. ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
4. CONVENE SMDI meetings and PRESENT prepared information. (Core Designer and other responsible NF engineers)
a. DISCUSS the procedural purpose and expectations of the SMDI meeting including active participation by all required organizations. (Core Designer)
b. CONFIRM all required procedural requirements are satisfied (Core Designer)
c. If meeting is cancelled prior to scheduled meeting time then RESCHEDULE (Core Designer)
d. If meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (NF VP or designee)

- If meeting is cancelled then RESCHEDULE. (Core Designer)

- If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member( s) and SOLICIT comments (Core Designer).

- WRITE an Issue Report documenting inadequate attendance. (SMDI members not in attendance, NF VP or designee).

5. DOCUMENT required Site VP concurrence in minutes to the SMDI meetings. (Core Designer)
6. DOCUMENT action items in minutes to the SMDI meetings and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
7. DOCUMENT meeting attendance in minutes to the SMDI meetings. (Core Designer)

N F-AA-100 Revision 13 Page 52 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 1 of 3 Reload Design Kickoff Meeting The Reload Design Kickoff consists of a team of NF and Fuel Vendor personnel that provides up front planning, input, and review of reload activities. It identifies responsible individuals for various reload design and licensing tasks, potential non-routine activities that may necessitate additional resources, and an overview of expected reload scope and schedule. This meeting provides an opportunity for the fuel vendor to seek clarification and understanding of the reload expectations.

The Reload Design Kickoff meeting should be attended by the following personnel or their nominated delegates. Additional required attendees are identified on a case-by-case basis, considering functional areas that may be affected by significant changes proposed for a specific reload. The Core Designer and BWR/PWR Design Manager are responsible for making this determination.

- NF PWR/BWR Design Manager (Chairperson - required)

- NF Unit Core Designer (required)

- Engineering Safety Analysis Engineer (required)

- Fuel Vendor Project Manager (required)

- Fuel Vendor Fuel/Core representative (required)

- Fuel Vendor Safety Analysis representative (required)

- Fuel Reliability Engineer (optional)

- NF Vendor Oversight Engineer (optional)

- Manager, Engineering Safety Analysis (optional)

- NF Fuel Supply Representative (optional)

- Other Fuel Vendor Representatives (e.g., Licensing representatives, Manufacturing representatives, etc.) (as needed)

- Other contributors (e.g., Reactor Engineering) shall participate as required.

NF-AA-100 Revision 13 Page 53 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 2 of 3 The conduct of Reload Design Kickoff meetings is as follows:

1. ORGANIZE and PREPARE for the meeting. (Core Designer and other responsible NF engineers, and/or Fuel Vendor representatives)
2. SCHEDULE the meeting and ENSURE adequate representation for the items to be discussed. (Core Designer and Fuel Vendor representative)
a. INVITE all required and optional members listed above.
b. IDENTIFY required attendees in meeting invitation.
c. IDENTIFY optional attendees in meeting invitation.
d. ENSURE meeting invitation clearly states that attendance is mandatory for the required attendees (or their designees).
3. CONVENE the meeting and DISCUSS required topics (Core Designer and other responsible NF engineers, and/or Fuel Vendor representatives). Topics that should be considered include the following items:

- fuel product lines and optional features

- core design approach (ILLCD, CCC, low leakage design, etc.)

- fuel channel designs

- control blades/control assemblies

- burnable absorbers

- operating domains plant modifications industry OPEX vendor and/or Exelon reload process/procedure changes reload methodology changes shipping containers reload schedule manufacturing schedule manufacturing product/process/procedure changes lessons learned from previous reload post job briefs collaboration kickoff meeting action items significant changes that were made to the inputs, assumptions, or methods used for reload analysis and/or licensing for prior reloads during either the reload analysis and licensing phase or during cycle operation

NF-AA-100 Revision 13 Page 54 of 54 ATTACHMENT 5 Conduct of Reload Design Kickoff Meetings And Team Responsibilities Page 3 of 3

a. If the meeting is cancelled prior to scheduled meeting time then RESCHEDULE.

(Core Designer)

b. If the meeting is convened, and all required personnel (or their designees) are not in attendance, then DECIDE if there is adequate representation to conduct meeting. (Chair)

- If meeting is cancelled then RESCHEDULE. (Core Designer)

- If meeting is held then PROVIDE meeting presentation materials and minutes to absent required member(s) and SOLICIT comments. (Core Designer)

- WRITE an Issue Report documenting inadequate attendance (Chair or designee).

4. DOCUMENT action items in meeting minutes and VERIFY that action items are added to the appropriate task tracking system. (Core Designer)
a. DOCUMENT meeting attendance in meeting minutes to the Reload Design Kickoff meetings. (Core Designer)
b. INCLUDE assignments to track resolution of any impacts to simulator cycle specific core update ability as referenced in Step 4.3.6. (Core Designer)

ATTACHMENT 6 EGC Procedure NF-AB-1 10, "Bundle and Core Design (BWR)"

Exekn SM NF-AB-110 Revision 10 Page 1 of 13 Nuclear Level 3 - Information Use BUNDLE AND CORE DESIGN (BWR)

1. PURPOSE 1.1. This procedure describes and outlines the processes and specific analyses require d

to generate reload bundle nuclear design(s) and core designs for Exelon BWR reactors. This includes specification of both the preliminary and final bundle/core designs, if needed, based upon the specific vendor/Exelon process map.

1.2. The bundle and core design processes are dependent upon each other. A reload bundle design must be shown to work within a specific core design in order to be acceptable for use. (CM-1)

2. TERMS AND DEFINITIONS 2.1. Bundle and Core Design - The bundle and core design specifies the type and the number of fuel bundles to be ordered for Design Cycle, It also specifies the core loading pattern for use in the reload licensing analysis.

2.2. Cycle Design Inputs and Requirements (CDIR) -- A formal document(s) containing cycle design targets and the inputs and acceptable outputs that define the boundaries of the cycle design and licensing analyses.

2.3. Current Cycle - The current operating cycle.

2.4. Design Cycle - The cycle in which the fuel assembly designs, core reload design and reload-licensing work is being performed.

2.5. Energy Utilization Plan (EUP) - The approved schedule of core thermal energy requirements (i.e., cycle length, power level, EFPD and end of cycle extension options such as coastdown) and outage dates input to the reload design.

2.6. Fuel Change Package (FCP) - A type of EC/ECR used to document, approve and implement core reload and cycle management activities.

NF-AB-110 Revision 10 Page 2 of 13

3. RESPONSIBILITIES 3.1. Core Designer - The Nuclear Fuels individual who is qualified in the reload bundle and core design processes and assigned by the Manager BWR Design to complete the work in the reload bundle and core design schedule. The Core Designer also provides guidance to the Verifier regarding the scope of verification and satisfactorily resolves the Verifiers comments.
4. MAIN BODY NOTE: Tasks in this Procedure have no associated Certification Guides except those called out in the referenced T&RMs.

NOTE: The process of determining the bundle and core design is iterative in nature. Each design step does not need to be performed in a given iteration (i.e. unacceptable results at one step negate the need to perform the remaining steps).

The steps in this procedure may be performed in an order different from that listed below.

NOTE: All of the tasks in this T&RM are the responsibility of the Core Designer.

NOTE: The Technical Task Pre-job briefing may be performed by the Core Designer as a reverse pre job brief, per HU-AA-1212, at the discretion of the Manager, BWR Design.

4.1. PERFORM a Technical Task Pre job briefing and risk assessment for the bundle and core design activities per HU-AA-1212.

4.2. ENSURE that the Design Cycle EUP and CDIR are available for use, as developed per NF-AA-105-1000 and NF-AB-105, respectively, and that the CDIR contains the Design Cycle hot and cold critical eigenvalue targets, as developed per NF-AB-1 10-3020 and NF-AB-1 10-3025, respectively.

NOTE: NF-AA-101 allows the bundle specification to be issued prior to full completion of the new fuel product line evaluation.

4.3. If a new fuel product line is being introduced for the Design Cycle, then INTERACT with the Fuel Reliability Engineer who is working with NF-AA-101 either prior to or in parallel with this procedure. (CM-1)

NF-AB-1 10 Revision 10 Page 3 of 13 4.4. DETERMINE the bundle designs and core design for the Design Cycle.

4.4.1. OBTAIN the cell friction model database inputs or parameters (from the fuel vendor if necessary) to support model validation for the Design Cycle core design. The inputs include, but are not limited to, inch-days, exposure, channel types (thick/thin etc.), channel material (Zr-2, Zr-4, NSF, etc.).

4.4.2. REVIEW the bundle and core design constraints and design goals described in the CDIR, and CONSIDER these criteria when determining appropriate candidate reload bundle nuclear design(s) for the Design Cycle.

NOTE: Consider the use of previous bundle designs for use in the Design Cycle.

4.4.3. DEVELOP a candidate reload bundle nuclear design(s) for the Design Cycle using either NF-AB-1 10-2000 or NF-AB-1 10-3005 to generate the nuclear physics cross section data for the reload bundle design.

4.4.4. DEVELOP a candidate core loading pattern for the Design Cycle per NF-AB-1 10-2210 using the candidate reload bundle nuclear designs from Step 4.4.3. (CM-3) 4.4.5. REVISE the bundle and/or core designs to meet the CDIR design criteria.

4.4.6. ENSURE that the core design is within the cell friction database. If the core design does not fall within the bounds of the database, then DETERMINE additional actions to be taken (e.g., monitoring, re-channeling, additional design margin).

NOTE: The Core Maneuvering Characteristics determination is normally performed based on the preliminary fuel cycle design (PFCD) as described in NF-AA-100.

4.4.7. EVALUATE the Design Cycle Core Maneuvering Characteristics determination per NF-AB-110-4010.

4.4.8. PERFORM a preliminary fuel storage reactivity evaluation and fuel shipping reactivity evaluation for the reload bundle nuclear design(s) using NF-AB-1 10-2070 as guidance and based on the acceptance criteria listed in the CDIR. Transmittal of the preliminary results is not required. (CM-2)

If the fuel storage reactivity criteria are not met, either RETURN to Step 4.4.3 and DEVELOP a new reload bundle design(s), or CONTACT the fuel vendor to obtain a revised criticality evaluation for the fuel bundle design(s) being considered.

4.4.9. ENSURE that the final bundle and core designs are acceptable and that they produce satisfactory results and comply with all design constraints and design goals listed in the CDIR. (CM-7)

NF-AB-1 10 Revision 10 Page 4 of 13

1. CONFIRM that the bundle designs meet the manufacturing, fuel storage/fuel shipment criticality and other Exelon or vendor-specific constraints and design criteria. (CM-5)
2. CONFIRM that the core design meets the hot excess reactivity, shutdown margin, margin to thermal limits, margin to fuel exposure limits, cycle energy requirements, axial power shape, fuel shuffling criteria, fuel channel distortion related requirements, control rod inventory requirements, control rod sequence exchanges and any other design criteria. (CM-4)(CM-6)
3. For Westinghouse Optima2 fuel bundles, CHECK the sub-bundle R-factors contained in the Westinghouse fuel bundle design report (s) and ENSURE that the sub-bundle R-factors are within the valid range of the Westinghouse critical power ratio (CPR) correlation as defined in the CDIR.
4. If necessary, RETURN to the appropriate previous steps of this procedure and ADJUST the core loading pattern, depletion rod patterns, and/or the bundle designs to produce satisfactory results and to comply with the design constraints and goals listed in the CDIR.
5. CONFIRM that the final bundle and core design complies with each line item in the CDIR.

A. ASSIGN responsibility for confirming each line item in the CDIR to the appropriate personnel. Individual CDIR line items may be assigned to the NF Core Design engineer, Engineering Safety Analysis engineer, fuel vendor engineer, or station engineer, as appropriate.

B. OBTAIN confirmation for each line item in the CDIR from the assigned personnel.

C. DOCUMENT confirmation of each line item in the CDIR as part of the owner acceptance review of the appropriate design documents.

5. DOCUMENTATION 5.1. TRANSMIT the final bundle designs and quantities to the fuel vendor and to Records Management with a copy to Fuel Supply via a TODI. (Reference CC-AA-310) (SRRS#3A.130) This transmittal should contain a picture of the bundles/lattices with pin-by-pin designations.

5.2. OBTAIN the Design Cycle fuel bundle design reports from the fuel vendor.

NOTE: The unique identifier of a fuel bundle design report is typically the name and number of the report as assigned by the fuel vendor.

NF-AB-110 Revision 10 Page 5 of 13 5.2.1. SPECIFY the unique identifier of each of the Design Cycle fuel bundle design reports in the FCP and DOCUMENT the fuel bundle design reports in the disposition section of the appropriate "BWR Core Reload" FCP for the unit/cycle of interest.

Per NF-AA-100-1000, the disposition section of the FCP is entitled "Design Analyses, UFSAR, Documents, and Meetings".

5.2.2. PERFORM an acceptance review of the fuel bundle design reports obtained from the fuel vendor.

1. PERFORM a technical task pre-job brief for the acceptance review in accordance with HU-AA-1212.
2. DOCUMENT the acceptance review in an action tracking item.
3. REFERENCE the action tracking item number in the appropriate section of the FCP. Per NF-AA-1 00-1000, this section of the FCP is entitled "Supporting Documents Preparation and/or Review".

5.23. PERFORM a fuel storage reactivity evaluation and fuel shipping reactivity evaluation for the reload bundle nuclear design(s) using NF-AB-110-2070 and based on the acceptance criteria listed in the CDIR. (CM-2) 5.2.4. DOCUMENT the Design Cycle fuel bundle design reports in the applicable FCP by performing either of the following steps:

1. ATTACH the fuel bundle design reports to the FCP, or
2. TRANSMIT the fuel bundle design reports to the station and to Records Management via a TODI (Reference CC-AA-310 and SRRS# 3A.130) or as a Record (SRRS# 3B.107).

5.3. OBTAIN a Multi-Cycle Analysis report from the fuel vendor. This analysis should utilize the final bundle designs documented in Steps 5.1 and 5.2.

NF-AB-110 Revision 10 Page 6 of 13 5.3.1. PERFORM an acceptance review of the Multi-Cycle Analysis report obtained from the fuel vendor.

1. PERFORM a technical task pre-job brief for the acceptance review in accordance with HU-AA-1212.
2. COMPLETE Attachment 1, Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review.
3. REFERENCE the completed Attachment 1 DVG and the Multi-Cycle Analysis report in the appropriate section of the FCP.

5.3.2. DOCUMENT the Multi-Cycle Analysis report in the applicable FCP by perform ing either of the following steps:

1. ATTACH the Multi-Cycle Analysis report and completed Attachment 1 DVG to the FCP, or
2. ENTER the Multi-Cycle Analysis report and completed Attachment 1 DVG in ERMS (SRRS## 3B.107).

5.4. CONSIDER documenting in the FCP (or in other less formal ways) any additio nal information that describes the development of the Design Cycle bundle and core designs, if that information will provide valuable reference material in the future for individuals who were not directly involved with the design.

6. REFERENCES 6.1. Station Commitments 6.1.1 Limerick and Peach Bottom
1. CM-1 T03969: SOER 96-02 Recommendation 1; "Evaluate Mixed Cores During Design" (Steps 1.2, 4.3)
2. CM-2 T03971: SOER 96-02 Recommendation 1; "Fuel Storage Reactivity" (Steps 4.4.8, 5.2.3)
3. CM-3 T03964: SOER 96-02 Recommendation 1; "3D Nodal Simulator Model" (Steps 4.4.4)
4. CM-4 T03980: SOER 96-02 Recommendation 2; "Use of Additional Design Margins" (Steps 4.4.9.2)
5. CM-5 T03981: SOER 96-02 Recommendation 3; "Perform Monitoring of Core Design" (Step 4.4.9.1)

NF-AB-110 Revision 10 Page 7 of 13

6. CM-6 T03970: SOER 96-02 Recommendation 1; "Design Should Account for Operating Data" (Step 4.4.9.2)

CM-7 T03976: SOER 96-02 Recommendation 2; "Control of Design Inputs and Calculation Methods" (Step 4.4.9) 6.2. Cross References 6.2.1. CC-AA-310, Transmittal of Design Information 6.2.2. HU-AA-1212, Technical Task Risk/Rigor Assessment, Pre-Job Brief, Independent Third Party Review, and Post-Job Review 6.2.3. NF-AA-100, Reload Control Procedure 6.2.4. NF-AA-100-1000, Core Reload and Cycle Management Configuration Changes 6.2.5. NF-AA-101, Nuclear Fuel Assembly and Core Component Design and Fabrication Process Changes 6.2.6. NF-AA-105-1000, Energy Utilization Plan Development 6.2.7. NF-AB-105, Managing Cycle Design Inputs and Requirements 6.2.8. NF-AB-110-2000, Bundle Design - GENIE 6.2.9. NF-AB-110-2070, Fuel Storage Reactivity 6.2.10. NF-AB-110-2210, Core Loading Pattern Development 6.2.11. NF-AB-1 10-3005, CASMO-4 Lattice Physics Calculations 6.2.12. NF-AB-110-3020, Hot Operating K-effective (BWR) 6.2.13. NF-AB-110-3025, Cold Critical K-effective (BWR) 6.2.14. NF-AB-110-4010, Core Maneuvering Characteristics Determination

7. ATTACHMENTS 7.1. Attachment 1, Design Verification Guide - MULTI-CYCLE ANALYSIS Accepta nce Review

NF-AB-110 Revision 10 Page 8 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 1 of 6

1. DOCUMENT NAME REV NO.

II. LIST OF PROCEDURES AND T&RM USED The following is a list of Procedures / T&RM which were used in the performance and verificati on of the analysis:

PROCEDURE/

T&RM REVISION TITLE NF-AA-100 Reload Control Procedure NF-AB-110 Bundle and Core Design (BWR)

Note: Procedures and T&RM not used should be lined out.

Ill. PROCEDURE / T&RM DEVIATIONS The following deviations and associated justifications from the Procedures / T&RM listed in Section I I were used for this analysis (mark N/A if no deviations were used):

NF-AB-110 Revision 10 Page 9 of 13 ATTACHMENT I Design Verification Guide - MULTI -CYCLE ANALYSIS Acceptance Review Page 2 of 6 IV. CHECKLIST Independent

  1. Item for Verification Reviewer Indicate Y, N, or N/A and Initial SECTION A - General Checks (N/A if not provided in report)

Al Does the report refer to the appropriate unit and fuel cycle(s)?

Are the identifying numbers and/or names for new fuel designs A2 correct?

A3 Is the correct fuel product line used for the reload and downstream fuel cycles?

Does the EOC N-1 projection cycle energy correspond to the value A4 specified in the CDIR?

Does the reload fuel cycle achieve the desired EOC energy specified A5 in the CDIR?

Do the batch size and splits for the reload fuel cycle match the batch A6 size and splits used in Exelon's core model?

Are the batch sizes for downstream fuel cycles reasonable (e.g.

A7 consider planned power uprates)?

SECTION B - Design Basis (N/A if not provided in report)

If provided in the report, do the criteria in the design basis of the report correspond to the CDIR criteria for:

131 Thermal limit margins?

B2 Hot excess reactivity?

B3 Cold and SLCS shutdown margin?

134 Fuel exposure limits and/or NEXRAT?

B5 Cell Friction Metric thresholds?

NF-AB-1 10 Revision 10 Page 10 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 3 of 6 IV. CHECKLIST Independent

  1. Item for Verification Reviewer Indicate Y, N, or N/A and Initial SECTION B - Design Basis (NIA if not provided in report)

B6 Rated or maximum reactor thermal power?

B7 Rated or maximum core flow and capability?

B8 Feedwater temperatures?

Bg Cycle extension techniques?

Hot and cold critical eigenvalue basis, including local cold k-critical B10 adjustment?

1311 Estimated OLMCPRs?

B12 Thermal Mechanical limits?

B13 SLCS shutdown margin parameters?

B14 Core average channel bow (e.g., BOWAVE)?

B15 Core loading pattern strategy?

B16 Target control rod pattern strategy?

NF-AB-1 10 Revision 10 Page 11 of 13 ATTACHMENT 1 Design Verification Guide - MULTI -CYCLE ANALYSIS Acceptance Review Page 4 of 6 IV. CHECKLIST Independent

  1. Item for Verification Reviewer Indicate Y, N, or N/A and Initial SECTION C -Calculations and Results (N/A if not provided in report)

Does the reload fuel cycle show conformance with all specified C1 design goals / limits?

Do the downstream fuel cycles show reasonable conformance with C2 all specified design goals / limits?

Do the results correspond reasonably to Exelon's core model of the C3 reload fuel cycle?

Does the core loading match Exelon's core model of the reload fuel C4 cycle?

C5 Are the core loadings for downstream fuel cycles reasonable?

Do the target rod patterns match Exelon's core model for the reload C6 fuel cycle?

Have all calculations been performed satisfactorily? (e.g. case C7 convergence to k-effective, modeling of FOR point)

Based on acceptability of reload fuel cycle results and general C8 reasonability of downstream fuel cycle results, are the fresh fuel bundle designs acceptable for use for their entire projected lifetime?

SECTION D - Other If any issues were identified with the Multi-Cycle Analysis, was an IR D1 Document IR #

written?

D2 Is the Multi-Cycle Analysis used by Fuel Supply for fuel amortization provided for the appropriate cycles (refer to NF-AA-100)?

NF-AB-110 Revision 10 Page 12 of 13 ATTACHMENT 1 Design Verification Guide -- MULTI -CYCLE ANALYSIS Acceptance Review Page 5 of 6 V. DESIGN VERIFICATION COMMENT SHEET PAGE OF VERIFIER COMMENTS I FUEL VENDOR RESOLUTION I VERIFIER RESPONSE (Date and sign each comment I resolution)

NF-AB-110 Revision 10 Page 13 of 13 ATTACHMENT 1 Design Verification Guide - MULTI-CYCLE ANALYSIS Acceptance Review Page 6 of 6 VI. SIGNATURES THE ACTIVITIES REVIEWED BY THIS DVG ARE COMPLETE, ACCUR ATE, AND ADEQUATE FOR APPLICATION TO:

(Plant / Unit and Cycle)

PREPARER:

FUEL VENDOR NAME DATE OF REPORT REVIEW:

CORE DESIGNER SIGNATURE DATE APPROVAL:

MANAGER SIGNATURE DATE NOTE: This DVG should be attached to the Multi-Cycle Analysis report when it is placed into Records.

The completed DVG is sufficient documentation of the Owner's Acceptance Review.

ATTACHMENT 7 EGC Transmittal of Design Information document NF1000236 Rev. 1

NUCLEAR FUELS TRANSMITTAL OF DESIGN INFORMATION q SAFETY RELATED Originating Organization NF ID# NF100(

x NON-SAFETY RELATED Nuclear Fuels Revision* 1 q REGULATORY RELATED 0 Other (specify) N/A SRRS # 3A130 Page 1 of 29 Station: Quad Cities Unit: 1 Cycle: 22 Generic: N/A

Subject:

Quad Cities Unit 1 Cycle 22 Cycle Design and Input Requirements (CDIR)

To: Lyfliam Sekkat (BWR Design, Westinghouse) EC/ECR#: 377652 Brian Henning Prepared by Signature John Wheeler Reviewed by Signature Date Jill Fisher Approved by Signature Date Status of Information: Z Verified q Unverified Action Tracking # for Method and Schedule of Verification for Unverified DESIGN N/A INFORMATION.

Description of Information: Revision 0: Provides design limits and targets for the Q1 C22 bundle design, core design, and cycle operation.

Revision 1 - Update information for cycle energy requirements, feedwater temperature operating domain, Boraflex degradation, and Zirio Lead Use Channels to close Unverified Assumptions of Revision 0.

Items A.25.b, A.27,b, A.32, and A.33 were adjusted to match the Revised Exelon CDIR procedure format.

Purpose of Information: Provides the bundle and core constraints for the cycle design.

Source of Information: As documented in the attachment.

Supplemental Distribution: E - Mail: Hard Gooy:

Craig Schneider (Quad Cities) Cantera Records Management Scott Vrtiska (Westinghouse)

Quad Cities 1 Cycle 22 CDIR TODI NF1000236 Rev. 1 Page 2 of 29 Section A -*

Item Core Design Data Related Value Referen ces/Comments Items Criteria Al. Cycle N-1 shutdown date 519/11 Exelon Nuclear Group's Planned Outage Schedule Rev. 28, 11111 /10 A.2 Cycle N startup date 6/8/11 Exelon Nuclear Group's Planned

-- ------ - ------ Outage Schedule, Rev. 28, 11/11/10 A.3 Cycle N shutdown date 4/1/13' Exelon Nuclear Group's Planned Outage Schedule, Rev. 28, 11/11/10 A.4 EOC N-1 nominal exposure 2049 GWd / 16,382 MWD/MT TODI NF1000203 Rev. 0 EUP Engineering input. Same band A.5 EOC N-1 minimum exposure 16,000 MWD/MT as 01021 and Q2C21 (-400 Cycle N EUP required FOR A.6.a energy and design power 1,842 GWD / 14,765 MWd/MTU;;. 2957 MWt TODI NF1000203 Rev. 0 (EUP)'

level FOR exposure acceptance A'6'b band, if applicable N/A N/A A.7 Cycle N EUP total cycle ever 1,921 GWD 115,391 MWd/MTU TODI NF1000203 Rev. 0 (EUP)'

I Exelon Reload Schedule Rev. 28 issued 11/11/10 revised the Q1 R22 outage date from 4/1/13 used in the EUP to 3/11/13. Because the bundle design and core design were already completed at the time of this revision, no changes to the EUP or CDIR energy requirements were made due to the Rev. 28 Outage Schedule. All of the Cycle N energies specified here are based on a 4/1/13 shutdown date.

II4Z°'t

Quad Cities 1 Cycle 22 CDIR TODI NF1 000236 Rev. 1 Page 3 of 29 Engineering input - no reference needed.

Design Limit must be met in both MICROBURN-132 and POLCA7.

A.8 I Design MFLPD limit 0,86 Design must meet these limits WITH applicable power/flow dependent limits (off-rated penalties) and WITH any applicable penalties from control blade history,

r- Jz,Ir0

Quad Cities 1 Cycle 22 COIR TODI NFl 000236 Rev. 1 Page 4 of 29 Engineering input -- no reference needed.

Design Limit must be met in both MI CROBU RN-82 and POLCA7.

Design must meet these limits WITH applicable power/flow dependent limits (off -rated penalties) and WITH any applicable penalties from control blade history.

A.9 Design MFLCPR limit 0.90 Design Limit is 0.90. Prior to incorporating CBH R- Factors, Design Limits should be 0.89 BOC - MOC and 0.88 MOC - EOC to provide margin to CBH R -Factors. If SCORE loading is used to eliminate or mitigate CBH R-Factors in limiting locations, 0.90 can be used for the whole cycle.

Additionally, the limit of 0.90 applies once the CBH R-Factors are O lamented.

2, 2 1 !o 1

Quad Cities 1 Cycle 22 CDIR T 1 NF1 0 236 Rev. 1 Page 5 of 29 Engineering input - no reference needed.

A.10 Design MAPRAT limit 0.86 Design Limit must be met in both MICROBURN-B2 and POLCA7.

1.38% margin value using best estimate target curve per NF-AB-110-Maintain 1,38% Ak/k based on short window using 3080, Rev. 10 and 1% margin to A.1 1 Cold shutdown margin limit best estimate cold target curve Tech Spec limit of 0.38%.

(apply this in both POLCA and MICROBURN-82) 1.38% margin is to include determination of SDM at most limiting temperature.

W methodology incurs no penalty for distributed vs. local critical k-effectives. A 0.04% Ak/k penalty A.12 SD M adjustments 0.04% Ak/k from inverted boron tubes is from ACTS 0920-01. Functionally, this 0.04% Ak/k penalty must be included in the SDM as part of the cycle R value.

<51- k 2'L 4%0

Quad Cities 1 Cycle 22 CDIR TOM NF1 0236 Rev. 1 Page 6 of 29 NF-AB-110-2060 Rev. 7 nominal values.

Peak set to minimize control rod BOG ? 0 . 80% dk density and maintain operational Peak <_ 1 . 70% . k flexibility; flatness set to minimize rod motion.

A.13 Hot excess reactivity limit (Peak - Minimum) < 0.50% Ak Design goals are as shown to Based on Nominal EOC N-1 maintain rod pattern flexibility and to Cycle Exposure reduce control rod inventory for shadow corrosion channel distortion mitigation. Hot excess is to be driven as low as possible under the limit of 1.70%, subject to complying with other design criteria.

MICROBURN-B2 POLCA GWD/MT Keff MB2 initial estimate target the same GWD/MT Kell 0 1 .0055 04 1 1.00227 as Q2C21 Eigenvalue Design Cycle N hot operating k-A.14 0.0 1.0015 .00 7 55 Analysis QDC-0000-N-1756 Rev. 0, effective target POLCA target eigenvalue from NF-13.0 1.0015 12.1 0.99100 17.0 1.0060 18.0 BEX-10-74 Rev. 1.

0.99100

Quad Cities 1 Cycle 22 CIDIR T of NF10 236 Rev. 1 Page 7 of 29 MICROBURN-B2 POLCA MB2 initial estimate target the same GWD/MT Keff GWD/MT Keff as Q2C21 Eigenvalue Design Cycle N cold k-effective 0 1.0020 0 1.0045 Analysis QDC-0000-N-1 756 Rev. 0.

A. 15 target 2.0 0.9995 4.0 1.0020 POLCA target elgenvalue from NF-18.0 0.9995 18.0 1.0020 BEX-10-74 Rev. 1.

Cold k-effective target SDM goal was increased from 1.0%

A.16 adjustments in addition to None to 1.38% since best estimate cold those in A.15 target was used.

NF-AB-110-3050, Rev. 10.

1.38% margin value using best estimate target curve with 1%

uncertainty and 0.38 Tech Spec limit.

Credit is taken for 30 alo enriched B-SLC shutdown margin limit, Maintain z 1.38% Ak SLCS Cold SDM based on short 10 (918 ppm natural boron A.17  % Ak and boron EOC N-1 © 918 ppm natural boron equivalent equivalent) per EC# 349585.

concentration (ppm)

SLCS SDM will be determined using POLCA7.

SLC SDM is to include determination of SLC SDM at most limiting temperature.

Quad Cities 1 Cycle 22 CDIR TODI NF1 236 Rev. 1 Page 8 of 29 WCAP 15836-P-A for Optimal exposure limits WCAP-15942-P- A Supplement 1 for Optima2 peak rod exposure limit of 62 GWd/MTU Peak Rod - 62 GWD/MTU Rod Nodal Exposure Limit based on Maximum Rod Nodal Exposure in A.18 Fuel exposure limits Peak Rod Nodal - 72 GWdD/MTU LHGR limit curve.

Peak Bundle - 58 . 7 GWD/MTU AmerGen Calculation No. C-1101-202-E620-4,43, Revision 0, "PWR &

BWR Isotopic Inventories for Spent Fuel Pool Gamma Heating Study."

The calculation states a pin exposure limit of 58,700 MWdIMTU but EC 378207 documents that the limit is a bundle limit, not a pin limit as stated in the calculation.

Engineering input - no reference needed Design allowance for Design to maintain A.19 exposure limits 3% margin to all exposure limits 3% margin to all exposure limits based on nominal EOC for all cycles in MICROBURN-132 and POLCA7.

WCAP 15836-P-A for O ptimal A.20 Residence time limits No limits for Optima2 fuel exposure limits; no residence time s e-cified.

110

Quad Cities 1 Cycle 22 CDIR TOO! NF1 236 Rev. 1 Page 9 of 29 Limits on design due to need Engineering input -No explicit A.21 to maintain full-core offload None reference needed to document there ca pability are no restrictions.

NF-AB-110-2210 Rev. 12 Limits on design due to need A.22 to maintain symmetry in core Load and operate to as near to 1/8 core symmetry as prat#icai . Maintain octant symmetry in exposure power and bundle type as much as possible; i maintain symmetric partners during bundle shuffles as much as possible.

Design Engineering input - no reference Sequence Sequence Sequ ence needed.

Exposures Start End MWD/MTU Rod Group Date Date Sequence Exchange dates and 0 - 2700 A2 - 10 6/8/11 10/1/11 exposures are approximate.

Sequences listed can consist of 27 00 - 5475 A2 - 9 10/2/11 1/28/12 different combinations (i.e. A2 can be Tar g et se q uence exchan ge A.23 groups 9 and/or 10).

plan 5475 - 8250 A2 - 10 1 / 29/12 5/26/12 If necessary to support thermal 8250 - 11000 A2 - 9 5/27/12 9/22/12 margins or other design criteria , the Rod Groups can be changed as long 11000 - 12475 A2 10 9/23/12 11/24/12 as no c o n t ro l rod rema i ns i nsert e d f or 12475 - 14800 A2 - 9 11/25/12 3/4/13 two consecutive sequence intervals.

14800 - 15425 ARC 3/5/13 4/1/13

Q uad Cities 1 Cycle 22 CDIR TODI NF1 236 Rev. 1 Page 10 of 29 NF-AB-440-1002 Rev. 2 (pending WCMS cutover)

NF-AB-440-1003 Rev. 1.

Specific control rod pattern This avoids reliance on complex rod guidance and limitations Avoid sequence exchanges in June-August; avoid significant rod pattern changes during sequence, if maneuvers to maintain adequate A.24a possible; avoid insertion of control rods from 48 in the thermal margin, etc; requires Periods to maintain constant relatively flat hot excess reactivity; to rod pattern middle of a sequence maintain MAPRAT within the design limits, rods may need to be insertediwithdrawn in the middle of a sequence; last sequence will allow for rods to be withdrawn to ARO Specific control rod pattern guidance and limitations A.24b +/-0.5 mk (+/-0.0005 k) NF-AB-110-3040 Rev. 11 Maximum deviation from target k-effective N F-A B-110-3040 Rev. 11 Specific control rod pattern NF-AS-440-1002 Rev. 2 (pending guidance and limitations Avoid use of B sequence rods; any rod that was WCMS cutover)

A.24c inserted during the previous rod sequence must be NF-AB-440-1003 Rev. 1 Control rod locations which fully withdrawn during the next sequence Fuel reliability concerns if any portion cannot be used of a bundle is continuously controlled for much longer than one control rod sequence

Quad Cities I Cycle 22 C iR TODi NF10 0 0 236 Rev. 1 Page 11 of 29 Engineering input -- no reference needed.

Preferred notch positions for control rods are 00, 06 20, and 48; control rods may be placed at Attempt to eliminate use of shallow Specific control rod pattern rod positions in design except as guidance and limitations intermediate notch positions 22-28, if necessary; control rods not allowed at notch positions 02, 04, required to provide MAPLHGR A.24d margin; goal is to maintain Rules for use of shallow rods and 30-46, but some shallow rod positions may be required for portions of the cycle; minimize banking of operational flexibility; avoid fuel rod groups at one position conditioning limit problems associated with intermediate rod movement Engineering input - no reference Specific control rod pattern needed guidance and limitations A.24e Assume no QOS rods It is assumed that any rods that are OOS rods COS at the end of the current cycle will be repaired and returned to service for the design cycle Specific control rod pattern Engineering input - no reference guidance and limitations Ensure adequate thermal margins exist for rod patterns at +/- 3 mk around the hot target eigenvalue needed A,24f Flexibility curve Maintain adequate flexibility -

Quad Cities 1 Cycle 22 CDIR TODI N F1 000236 Rev. 1 Page 12 of 29 Engineering input - no reference Specific control rod pattern needed guidance and limitations Preferred order of withdrawal to FOR (last sequence)

A.24g is group 9B, 9C, then 9A last. Operational preference. This order is Misc not required for compliance with core design requirements.

Engineering input - no reference needed Core flow window to use in 98 - 100% Design with nominal 98% core flow A . 25 . a (96.04 MIbfhr) for all state-points, design increasing to 100% (98 Mlb/hr) at EOR.

No ICF assumed.

Low Limit - 100% (100% CTP @ 98 Mlb/hr - No ICF)

High Limit - 101.7% (100% CTP @ 95.5 Mlb/hr)

Unit Rod-line Restrictions Engineering input - no reference (Low and High) for operation needed for High & Low FCL Limit A . 25 . b FCL = CTF7i x 100 to be used in the reload 22 .191 + .$9714 x i^t ,^ + - ^.{^J1 19S x Wr% ? TODI QDC-02-028.01 and QCOP sign.

design.

Where CTP% = percent of rated thermal power 0202-13 Rev. 15 for FCP equation Wt°/o = percent of rated core flow Quad Cities 1 Operating License DLO I SLO Values shown are expected to A.26 Cycle N-1 SLMCPR values 1.11 / 1.13 (Optima2) continue to be used in Q1C22 based on Westinghouse calculated values for previous Optima2 cycles. These are the values currently in Quad Cities Tech Specs.

c I+ 7.

t i lio

Quad Cities 1 Cycle 22 CDR TODI NF1 0 236 Rev. 1 Page 13 of 29 Engineering input - no reference needed Estimates shown are for Base Case NSS and no EGOS. Based on Current Estimates based on previous cycles.' Q2C21 actual limits of 1.43 BOC -

14,000 MWd/MTU / 1.45 from 14,000 A.27.a Estimated OLMCPRs to use Optima:?.: and Q1C21 actual limits of 1.43 from in design 1.44 for <11,000 MWD/MTU BOC to 13,275 MWdIMTU and 1.49 1.48 for X11,000 MWD/MTU from 13,275 MWd/MTU.

MCPR OL must be >1.41 for LOCA analysis per Westinghouse document NF-BEX-06-44-P (April 2006) and Quad Cities Design Analysis OPTIMA2-TR021 QC-LOCA Rev. 5.

N/A - OPRM amplitude setpoint values will be Range of acceptable OPRM adjusted on the OPRM system and will not set the A.27.b cycle OLMCPR limit Engineering input - no reference amplitude setpoint values needed Assume no setpoint change for design

'The values include margin to the best-estimate expected OLMCPR. The SLMCPR values may change due to the issue identified in IR 1129289, which would impact the Q1 C22 OLMCPR. The impact of the SLMCPR issue for Q1 C22 is unknown, and these values remain a reasonable estimate of the Q1C22 OLMCPR.

.nom I, (l t)l4Ii

Quad Cities 1 Cycle 22 IR TODI N F1000236 Rev. 1 Page 14 of 29 Lattice Burnup Bottom Plenum Middle Plenum GWd/MTU Bottom and Middle and To NF-BEX-09-188 Rev. 0 (Q2C21 RLR) 0 8.50 8.50 8.60 Based on limiting Q2C21 UE21 bundle design with --0.1 reduction in 12 9.00 9.00 9.20 MAPLHGR limits .

A . 28 APLHGR Limits to use in 24 9 . 50 9 . 50 9.80 d es i gn Better MAPLHGR estim ate s may be used based on preliminary 30 9 . 50 9 . 50 9. 80 MAPLHGR li m i t calculations 50 9.50 9.50 performed by Westinghouse during 9.90 the design as they become available.

60 9,50 9.50 9.80 72 9.50 9.50 9.80 GWD/MT LHGR Li mit 0 13.11 NF-BEX-09-188 Rev. 0 (Q2C21 RLR)

LHGR Limits to use in A.29 14.0 13.11 design 72.0 6.48 Standard OPTJMA2 LHGR limits A . 30 Transient LHGR Limits to use in design N/A Transient LHGR limits not used Assume off-rated thermal limits will be A.31 Off-rated thermal limits to similar to those calculated for Q2C21.

use in desi g n See NF-BEX-09-1 B8 (Q2C21 RLR)

These multipliers can be found in the NF-

-. BEX-09-188 Rev. 0.

Quad Cities 1 Cycle 22 CDIR TODI NF1 236 Rev. 1 Page 15 of 29 OLMCPR 1.41.

GE SIL 320 Supplement 3 For fluence related bow:

NF-AB-105 Revision 13 GE SIL 320 Supplement 3 summation criterion <2 (C lattice criterion). Bundles with X25 , 000 effective inch-days and less than 40 GWD/MT Fuel channel distortion For shadow corrosion (Inch-days), at FOG Q1 C22 , channel exposure may be monitored A.32 limitations and mitigation no cell shall contain: rather than being re-channeled strategy MCPR OL must be X1.41 for Channel Any bundle with 225,000 effective inch-days and 240 GWDIMT channel exposure Distortion analysis per Westinghouse document NF-BEX-10-70 Rev. 0 OR (May 21, 2010), which transmitted the 3 or more bundles with x20,000 effective inch- report BTF 0!9-0275, rev 2, days and X30 GWD/MT channel exposure. "Westinghouse methodology for enhanced channel bow indications."

I^rl 1`t2"

Quad Cities 1 Cycle 22 CDIR TODI NF1000236 Rev. 1 Page 16 of 29 NF-AB-130-2 220 Rev. 8 Minimum allowed calculated reactor period and maximum Individual notch worths between 04-36 and total Notch worths must consider A.33.a allowed single notch worth worth between 36-48 must be less than the Ak that sequence steps beyond Groups 1-4 for in-sequence notch worth would result in a 50 second period (Groups 7,8,9, and/or 10) as determinations necessary to satisfy requirements up to +3.5% keff around critical.

Range around the expected cold critical eigenvalue when NF-AB-130-2620 Rev. 8 A.33.b the notch worth and step -1.2°1 to +3.5% keff around the expected cold critical worth limits are to be eigenvalue evaluated at 120 OF and 320 T. Bounding range based on enforced uncertainties for non BOG criticals Step worth (sum of all notch worths between movement limits) must be less than the minimum of:

0.005 Ak NF-AB-130-2620 Rev. 8 or Maximum rod step worth (7/64) * [ ko- Wl-0.0112 ] Ak) Notch worths must consider allowed in the defined range where A.33.c sequence steps beyond Groups 1-4 around cold criticality for the startup sequence ko= the cold critical target keff at each exposure (Groups 7,8,9, and/or 10) as point of interest necessary to satisfy requirements up kARI = the cold all rods inserted keff at each to +3.5% keff around critical.

exposure point of interest Code packages which are to be utilized in the POLCA7 for official notch worth determination as A. determination of cold in- reported in the CMR. Criteria must also be satisfied Engineering input No reference sequence notch worth in MICROBURN-B2. needed values Core loading restrictions to ATI 271647 -02, "Determine the be used to minimize risk of Minimize clumping of once-burned bundles in high apparent cause of the high notch A.35 abnormally high notch importance areas, such as between the edge of the worths for D3C19 and address the worths during in-sequence fresh fuel zone and the periphery other recommendations in the lR."

critical 1/14/2005.

Quad Cities 1 Cycle 22 CIR TOQI NF1 000236 Rev. 1 Page 17 of 29 As documented in OPTIMA2-TR038QC-EOP, Rev.1, "SVEA-96 OPTIMA2 Fuel Input to the Emergency Operating Procedures."

Westinghouse letter NF-BEX-06-282 Parameters XB-cld-nat, XB-hot-nat, and MSWBP (dated 12/14/06) contains the must be verified for 01 C22. Optimal EPG parameters.

Hot SLC SDM requirement is ^ 0.38% Ak/k and QGA 101 Rev. 13.

A.36 EOP/EPG/SAG Parameters 700 ppm natural boron consistent with OPTIMA2-TR038QC-EOP, Rev.1.

MSBWP is to include determination Reactor is shutdown by 0.38% Ak/k xenon free at the of shutdown conditions at most most reactive temperature with the strongest rod at limiting temperature.

position 48. QGA 101 requires all other rods inserted to at least position 04. The site desired alternative to 04 is to evaluate all rods at 02 and change QGA 101 to require that all rods in to at least position 02 for MSBWP.

NF-AB-110-3040 Rev. 11.

Proposed core designs and rod MCO < 0.1 % per GENE MCO analysis model patterns will be checked via GENE MCO analysis model. This will A.37 MCO limit require performance of calculations to Maximum bundle RPF < 1.65 and no more than 3 bundles with RPF > 1.60 within 3X3 array. determine moisture carryover compliance based on results from MICROBURN-B2. RPF criteria are from NF T&RM NF-AB-110-3040

__m._ Rev. 11.

Quad Cities 1 Cycle 22 CDI R TODI NF1 236 Rev. 1 Page 18 of 29 Final Task Report for TSD DQW04-024 (Attachment 1 to NF-BEX-05-155

1. EOC CAVEX < 37,000 MWD/MT Rev. 0)
2. Initial core weight < 126.0 MTU A.38 Decay heat analysis 3. Initial core average enrichment between 3.90 and These criteria are the limits of 5.00 w/o U-235 (treat all fuel as fresh). applicability of the Westinghouse decay heat analysis. These criteria were also assumed in the Westinghouse evaluation of alternate source term.

GNF- GE-NE-A22-401033-02 Rev. 0, Axial and radial power profiles consistent (similar) August 2000.

A.39 Fluence assumptions with past analyses -- 1 natural U node at bottom, 2 at top; similar radial per profiles loading least reactive Westinghouse - NE BEX 45-114 fuel on edge Rev. 0 , 8f23f45.

EC 349583 Rev. 1 Section 4.1.16.

WCAP 16081-P-A Addendums I and CPR correlation limitations 2 to CPR correlation have been A.40 None. approved by NRC and incorporated on core design into POLCA, WCMS, and MICROBURN-B2.

-d 0

Quad Cities 1 Cycle 22 CDIR TODI NF1 2 6 Rev. 1 Page 19 of 29 These criteria are bounding values, and they were the basis for spent fuel

1. Bundle enrichment range from 3.38 to 4.50 w/o pool heating calculations as U235 documented in AmerGen Calculation
2. Bundle uranium loading (nominal) < 197.0 kgU No, C-1101-202-E620-443, Revision Spent fuel pool gamma 3. Bundle exposure < 58,700 MWD/MTU 0, "PWR & BWR Isotopic Inventories A . 41 heating constraints 4. Average bundle power (reactor rated thermal for Spent Fuel Pool Gamma Heating power / # bundles) < 5.586 MWt Study." The calculation states a pin
5. Radial Peaking Factor < 2.00 exposure limit of 58,700 MWd/MTU
6. Axial Peaking Factor < 1.80 but EC 378207 documents that the limit is a bundle limit, not a pin limit as stated in the calculation.

Fuel which resided on the periphery (on the first or A.42 Fuel promotion limitations second row) during the current or previous cycles NF-AB- 110-2210 Rev. 12 should not be promoted inward ast the fourth row.

Feedwater Temperature Operating Domain based on nominal FW temperature of:

TODI QDC-10-026 Rev. 0 (T)

A . 43 Reactor dome pressure and -98`(Q/Qs)2 + 242.78'(Q/QR) + 210.82 feedwater temperature NF-BEX-05-24, Revision 2.

Where: Q/QR is CTP/ rated CTP (Pressure)

Q1 C21 Reactor Dome Pressure O ratin Domain

Quad Cities 1 Cycle 22 COIR TODII NF1 236 Rev. 1 Page 20 of 29 Engineering input - No reference needed

1. Face adjacent shuffles within a cell increases shuffle time dramatically
1. Eliminate face adjacent within cell shuffles if 2. Cross quadrant shuffles will affect possible ability to move to another quadrant

. Miscellaneous Shuffling should an SRM fail during shuffling A 44 Restrictions 2. Avoid cross quadrant shuffles (i.e. compromise quadrant independence).

3. Control cell re-use
3. Insure that any bundle that was controlled during the last sequence in Q1C21 will have adequate periods of uncontrolled operation before being controlled in QiC22.

Axial Peaking for small Westinghouse memo NF-BEX A.45 Maximum peak in top half of core (node 20) < 1.6 break LOCA Anal sis 246

,1110

Quad Cities 1 Cycle 22 CD1A TODI NF1000236 Rev. 1 Page 21 of 29 Item Fuel Bundle Design Data Related Items Value References/Comments Criteria a8 Gd rods per lattice CN-EXELONBWR-05-24, Revision 0 Gd w/o based on Boraflex degradation and Gd w/o as OPTIMA2-TRO25DQ-SFP, "Optimal Spent Fuel defined in CN-EXELONBWR- Pool & New Fuel Vault Evaluation --

05-24 Rev. 0 Dresden/Quad Cities" Rev. 0 Average enrichment of the Boraflex Degradation assumptions to be used B.1 Spent fuel pool criticality criteria peripheral rods in each lattice based on preliminary interpretation of NETCO is less than the lattice-average Boraflex degradation results, enrichment a) 7 inch uniform gap b) 50% loss in thickness No Gadolinia rods on bundle c) 5% loss in width periphery Face adjacent Gd rods shall be counted as a 30 cm top and 15 cm bottom single rod.

Natural U blankets.

OPTIMA2-TRO25QQ-SFP, "Optima2 Spent Fuel Pool & New Fuel Vault Evaluation -

Max 235U Enrichment: 4.95% Dresden/Quad Cities" Rev. 0 B.2 New fuel vault criticality criteria Min. # Gd rods a Min Gd 203 Enrichment 5.5% Face adjacent Gd rods shall be counted as a single rod.

1a

Quad Cities 1 Cycle 22 CDIR TOQt NF1 236 Rev. 1 Page 22 of 29 Section B -

Item Fuel Bundle Design Data Related Items Value Referee Comments Criteria Max 236U Enrichment: 5 . 0% NRC Certificate of Compliance No. 9292 for the Min Gd203 Enrichment 4.0% Model No. PATRIOT Package, Revision 6, No Gd rods on edge or corner Docket Number 71-9292.

Fuel shipping containers criticality # Gd rods (96 fuel B.3 Gd rod diagonally rods) Letter, Brian Beebe "to Daniel Redden, NF-@EX-criteria 12 symmetric

  1. Gd rods (92 fuel rods) 10 05-166 Revision 0, Additional Information on Gd rod diagonally symmetric Patriot Container Certificate of Compliance " ,
  1. Gd rods (84 fuel rods) 8 October 21, 2005 and NF-BEX-06-67, "Additional Min 2 Lid rods per quadrant Information on Patriot Container Certificate of Compliance " , March 9 2006.

10 CFR 70.24 criticality monitoring B°4 exem tionS N/A No criticality monitoring exemptions.

B.5 Fuel type to be manufactured for this reload Optimal Engineering input - No reference needed BTK04-164, "Mechanical Data Input to Nuclear and Thermal- BTKO4-164, "Mechanical Data Input to Nuclear B.6 Source for fuel product line dimensions Hydraulic Design, Quad Cities and Thermal-Hydraulic Design, Quad Cities I & 2 1 & 2 and Dresden 2 & 3 and Dresden 2 & 3 SVEA-96 Optimal" , Rev. 1, SVEA-96 Optimal", Rev. 1, 3/24/05 3/24/05 B.7 Channel bow assumptions None No channel bow assumptions for bundle desi n.

CPR correlation limitations on bundle C} ptima2 bet sub- bundle factors 6.8 must be between 0 . 89 R and SEA WCAP-1 P. Optimal CPR correlation design is valid within this range.

mu No explicit lattice or power/exposure peaking B.9 Limitations on lattice power peaking None limits for bundle design. This includes local peaking factors.

Core loading restrictions due to "other" Load bundles with high 8.10 control/inch days in non- Engineering input - no reference needed factors nodded locations on periphery.

I i I =(,0

Quid Cities 1 Cycle 22 CDIR T'OD NF1 0236 Rev.

Page 23 of 29 Section B -

Item Fuel Bundle Design Data Related items Value Refere Comments Criteria

1. No gad rods face adjacent to central portion of water cross or face adjacent to other gad rods (gad rods 1. Basis is modeling uncertainty. No explicit Limitations on the are allowed to be next to reference.

B.11 piacement/enrichment of gadolinia connecting wings of water bearing rods cross); no gad rods in part 2. Appendix B of Westinghouse DR/QC length rods; no gad rods in contract. Manufacturing instrumentation outer row of bundle sensitivity

2. Concentrations in a single rod must differ b z 2 w/o Standard concentrations Maximum 8 w/o Limitations on the amount of gadolinia B.12 Appendix B of Westinghouse DR/QC contract allowed in the fuel pellets 3 concentrations per reload 2 concentrations per rod type (including zero concentration Engineering input - no reference needed.

Target bundle enrichments for this 13,13 3.84-4.20% Bundle enrichments in the lower part of this reload design range may be used to help maintain the peak hot excess reactivity, if necessary.

8.14 Target minimum length for bottom fuel Westinghouse fuel design does not have this zone None limitation on the bundle desi n.

Quad Cities 1 Cycle 22 CDI R TOD F1 236 Rev. 1 Page 24 of 29 Section B -

Item Fuel Bundle Design Data Related Items Value References/Comments Criteria EC 349583, `Implement Westinghouse Optimal Nuclear Fuel" 30 cm at top of bundle (2 nodes) Consistent with previous Optimal core designs.

Length of natural U zones at top and 8.15 bottom of bundle 15 cm at bottom of bundle (1 A more detailed reactor fluence evaluation is node) required before the natural U top zone can be decreased to less than 30 cm per information resented in EC 349883.

Engineering input - no reference needed W CBH evaluation tool has been developed for use during bundle design. Design strategy is to keep the final maximum MFLCPR with CBH penalty added to below, or as close as possible, Maximum allowable control blade Select pin enrichments to to the design goal of 0.90.

8.16 history delta MFLCPR penalty minimize CBH impact on CPR W will also evaluate proposed rod patterns for GBH impacts.

CBH will impact the bundle design and the extent to which fresh bundles are placed in cells that are tanned to be controlled during operation.

Maximum rotated bundle delta CPR None Westinghouse methods do not have this limit 6 . 17 allowed Fuel manufacturing constraints which No special restraints.

8 . 18 None may be challen ed ____

No explicit lattice or power{exposure peaking Exposure peaking limits on the lattice limits for bundle design. This includes local B , 19 None designs akin factors.

2

Quad Cities 1 Cycle 22 C DIR TODI NF1 000236 Rev. 1 Page 25 of 29 Section B It Fuel Bundle Design Data Related Items Value Reefers Corn tits Criteria Current limitation of Westinghouse automated fuel rod loading machines is 15 rod types. If >15 Manufacturing preference is rod types, then special hand loading is required.

maximum of 15 fuel rod types Note that unique rod types are characterized by B . 20 Maximum number of unique rod types in a given Optima2 bundle the enrichment, gadolinia concentration and axial type distribution, rod length (113 PLR, 2/3 PLR, or full length rods), and whether the rod is a tie rod, spacer-capture rod, or a regular fuel rod.

Per Westinghouse process, gadolinia power 8.21 Target gadolinium suppression penalty Not Allowed -suppression is not allowed for Optimal fuel.

SER-WCAP-16081-P Optirna2 CPR correlation is valid within this range of R-factor.

Optimal sub-bundle R-factor 8.22 Target R-Factors must be Z! 0.89 and 5 1.11 This will be checked and documented during bundle design work to ensure that the bundle designs comply with the R-factor requirements, rather than relying on the check to be done within the CPR correlation subroutine Westinghouse design methods do not require or use a target local peaking factor. Peaking 8.23 Target local peaking factors N/A factors must meet the design criteria identified in other sections of the CDIR.

NF-AB-440 Rev. 13 Verify lattices are bounded by NF-AB-0-1002 Rev. 2 (pending WCMS Optimal REMACCX PO curve cutover) in NF-AB-440-1002 8.24 Lattice Design for Fuel Reliability (LRG vs.nodalnexposure) if If lattice does not satisfy limits, notify site that (LH R v s. o is used for core generic Optimal PO curve does not apply for this monitoring

_- ^- c qLe i.e. will be altered slightly iil12ir

Quad Cities 1 Cycle 22 CDIR TOM NF1000236 Rev. 1 Page 26 of 29 Section C - Value Retere Comments Ite m M iscella neous Data Related Items Criteria NF-AB-440-1002 Rev. 2 (pending WCMS In accordance with Exelon C.1 Fuel conditioning guidelines cutover) procedure NF-AB-400-1003 Rev. 1 Engineering decision - no explicit reference.

Computer code(s) to be used for lattice PHOENIX-4 / POLCA-7 and Goal is to provide adequate design margins in C.2 and core design, and any special CASMC)-4 / MICRC}BURN-B2 both code packages; differences will be version requirements managed b en ineerin judgment.

Standard Quad Cities TOOl NF0400171 Rev. 0.

C.3 Power-flow map Power/Flow None planned.

Technical Specification changes for the SLMCPR is the only identified potential Technical C.4 SLMCPR may be required cycle Specification change.

based on SLMCPR analysis results C.5 Planned core component changes None N/A Studies or projects in progress which C.6 could impact the reload design or None None licensing i I (m /to

Quad Cities 1 Cycle 22 CDIR TODI N F 10. 0236 Rev. 1 Page 27 of 29 Section C -

item Value References/Comments Miscellaneous Data Related Items Criteria Bundle exposure <62 NF-AB-110-2210 Rev. 12.

GWD/MT Reg. Guide 1.183 "Alternative Radiological Peak rod average exposure <

Source Terms for Evaluating Design Basis 54 GI/MT Accidents at Nuclear Power Reactors';

Or, if Peak rod average exposure C.7 AST analysis limits QC design analyses QDC-0000-N-1267, Rev.

54 GWDlMT, then peak rod 1A, " Re-analysis of Fuel Handling Accident average LHGR must not (FHA) Using Alternative Source Terms " , and exceed 6 . 3 kW/ft .

QDC-0000-N-1268, Rev. 2A, "Re-analysis of Control Rod Drop Accident (CRDA) Using Maximum RPF must be <

Alternate Source Terms."

1.70.

C.8 Items which are critical to quality See CDIR Attachment 1 Engineering Input Fuel Type(s) to be used in downstream Assume Optimal will be loaded in all future C.9 Optima2 multi-cycle analyses cycles for multi-cycle analyses C.9.1 Discharge Average Burnup Target 48 GWd/MTU Westinghouse DRJQC contract.

Design Cycle + 4 future C.9.2 Multi-Cycle Analysis Length Engineering Input.

cycles,

Quad Cities 1 Cycle 22 C D IR TODI NF1000236 Rev. 1 Page 28 of 29 Section D -

Item Planned Change Related Items Value References/Comments Criteria ECs 374977, 365821, 373667, 366920.

The effects of the turbine modifications are D.1 Planned station modifications that may Turbine Overhaul included in Exelon TODl QOC 10-27 Rev. 0, impact reload analysis Modifications " OPL-W Parameters for Quad Cities Unit 1 Cycle 22 Transient Analysis", 9/15/14.

TODI QDC -1 0-026 Rev. 0, "Quad Cities Units I Planned setpoint changes that may Feedwater Temperature D.2 & 2 Nominal Feed Water Temperature Curve " ,

impact reload analysis Operating Domain revision 8/24110.

Planned station minor modifications that D3 m impact reload analysis None . N/A D'4 Planned component changes that may impact reload analysis None N/A A

102,1110

Quad Cities 1 Cycle 22 CDIR TOO1 NF1 0 239 Rev. 1 Page 29 of 29 CDIR Atta c hm ent 1 (ite m C.8)

Q1C22 Exelon Fuel Design CTQ's t . Attempt to reduce redundancies during the reload design effort such that both Exelon and W share in the preparation and review of output documents.

2. Design in compliance with CDIR specifications and meet Major Milestones in the Design & Licensing Schedule.
3. W performs multi-cycle analysis consistent with Exelon EUP following completion of Reference Loading Pattern. Exelon will utilize this information for economic/amortization purposes.
4. Open access to W core and bundle design procedures by Exelon. Mutual agreement on how to resolve conflicts, documenting exceptions when taken.
5. Improve design quality through collaboration.
6. Attempt to minimize the number of shadow corrosion induced channel bow susceptible bundles within a bladed cell. The objective is to minimize Exelon`s risk of channel-control blade interference utilizing Exelon and industry operating experience,
7. Adhere to SIL 320 (latest supplement) loading pattern restrictions to minimize fluence gradient induced bow impacts on channel to rod interference. All fuel types are to be treated as GNF fuel from a SIL 320 perspective.
8. Resulting design is acceptable when satisfactory results are obtained with both POLCA-7 and MICROBURN-82.
9. Expanded notch worth criteria and step worth criteria are met and margin to the limits is communicated.

lt(72 /tO

ATTACHMENT 8 EGG Calculation QDC-0000-N-1804 Rev. 0, "Quad Cities Unit 1 Cycle 22 Core Loading Plan"

CC-AA-309-1 001 Revision 6 Page 1 of 9 ATTACHMENT 1 Design Analysis Major Revision Cover Sheet Page I of 5 Design Analysis (Major Revision) Last Page No. 9 B14 Analysis No.: ' QDC-0000-N-1 804 Revision: 0

Title:

3 Quad Cities Unit 1 Cycle 22 Core Loading Plan EC/ECR No.: 4 377652 Revision: 0 Station(s): Quad Cities Component(s): "

Unit No.: 1 N/A Discipline: a NUDC Descrip. Code/Keyword: 'a N01 Safety/OA Class: SR System Code: N/A Structure: 'a N/A CONTROLLED DOCUM ENT REFERENCES Document No.: From/To Document No.: From/To OPTIMA2-TR038QC-EOP From QDC-0000-N-1806 To QDC-0000-N-1805 From QDC-0000-N-1653 From Is this Design Analysis Safeguards Information? 1e Yes[] Noo If yes, see SY-AA-101-106 Does this Design Analysis contain Unverified Assumptions?" Yes[] No Z If yes, ATI/AR#: VIA This Design Analysis SUPERCEDES: 1e N/A in its entirety.

Description of Revision (list changed pages when all pages of original analysis were not changed): 19 Original Issue Seth Spooner -^J / 1 e l t.^ r !

Preparers 20 Brandon de Graaf j y f fS hf Print Name SI Name ^ Date Method of Review: _' Detailed Review Alternate Calculatl (attached) q Testing q Reviewer: 22 Eric Bruch 412 91 11 Print Name Sign Name Date Review Notes: " Independent review 0 Peer review Sep 11 5.

ITPR : J oh n R e i mer ^ ^ -^ ^ `^/ fzv r (Fm External Analyses Only)

External Approver: u N/A N/A N/A Print Name Sign Name Date Exelon Reviewer: 25 N/A N/A N/A Print Name SI Name Date Independent 3`d Party Review Reqd? m Yes No Exelon Approver. 17 Jill Fisher 5 f

,1oLIfri.__-_

Print Name Sin Name Date

-f- Rtv;W o' cd blade- m aps b" A l e,

Exek,n. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page 2 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN DESIGN ANALYSIS TABLE OF CONTENTS SECTION: SUB-PAGE PAGE NO.

NO.

CC-AA- 309-1001 Major Revision Cover Page 1 N/A Design Analysis Table of Contents 2 N/A 1.0 Purpose 3 N/A 2.0 Inputs 3 N/A 3.0 Assumptions 4 N/A 4.0 References 4 N/A 5.0 Identification of Computer Programs 5 N/A 6.0 Method of Analysis 5 N/A 7.0 Numeric Analysis 6 N/A 8.0 Results 7 N/A 9.0 Conclusion 7 N/A Attachment A - Core Loading Plan Al - All N/A Attachment B - Independent Review Information B1 - B14 N/A

xel !,n. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page 3 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN List of Tables Table 1 - Computer Codes Used In Analysis .............................................................................5 Table 2 - CAKWFT 1.4 Output .....................................................................................................6 1.0 Purpose The purpose of this design analysis is to document the Core Loading Plan Revision 0 for Q1 C22 per T&RM NF-AB-130-2200, Revision 1 (Reference 4.3).

The current revision of this Design Analysis (Revision 0) is part of the work covered under the Fuel Change Package FCP 377652, Quad Cities Unit 1 Cycle 22 Reload Design (Reference 4.5).

2.0 Inputs 2.1 Core Loading by Assembly ID and Fuel Type The fuel assembly ID loading map is from the Westinghouse Reference Loading Pattern (Reference 4.6, pg Al -15, Al -16), modified with two changes: 1) replace bundle QAA21 1 in location 05-40 with bundle QAA1 39; and 2) replace bundle QAA213 in location 05 -22 with bundle QAA1 41. Both QAA1 39 and QAA1 41 were planned to be discharged in the Reference 4.6 loading pattern. These changes were made to address a suspected fuel failure detected in 01 C21 (see Reference 4.9 for loading changes).

The fuel type loading map is from the Q1C22 Design Analysis for the MICROBURN-132 basedeck (Reference 4.9). This design analysis incorporates the replaced bundles, therefore no modifications were made to the map.

2.2 Core Weight The as-built core weight is determined using the Q1C22 Design Analysis for the MICROBURN-B2 basedeck (Reference 4.9). Fresh fuel as-built bundle mass values were used as input to the deck, and the MICROBURN-B2 output file shows overall core weight as 124.83 MTU.

2.3 Control Blade Information This information is from the 01 C21 CLP (Reference 4.10) together with the changes described in Quad Cities blade replacement TODI (Reference 4.11), which shows details regarding planned blade replacements and shuffles during Q1 R21. Blade IQs are compared to Reference 4.4 Attachment 11 to determine which of the Marathon control blades are part of etch indication population.

Exe kn. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page 4 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 2.4 EPG/SAG Information Parts of this information are directly from the Reference 4.6 report (# of Optima2 bundles on page `A1-14') and Reference 4.7 Westinghouse report (Optimal active fuel lengths, total mass of U02, total mass of clad and channels on pages 9-12). There is also calculated information (Optima2 bundles' mass, total mass of Uranium), which is presented in Section 7.0 Numeric Analysis.

3.0 Assumptions None.

4.0 References 4.1. Exelon Procedure, CC-AA-309, Revision 10, "Control of Design Analyses".

4.2. Exelon T&RM, CC-AA-309-1001, Revision 6, "Guidelines For Preparation And Processing of Design Analyses".

4.3. Exelon T&RM, NF-AB-130-2200, Revision 1, "Core Loading Plan Generation".

4.4. Exelon T&RM, NF-AB-135-1410, Revision 8, "BWR Control Blade Lifetime Management".

4.5. Exelon Fuel Change Package 377652-000, "Quad Cities Unit 1 Cycle 22 Core Reload Design".

4.6. Westinghouse Report NF-BEX-10-184, Revision 0, "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22", dated 12/3/10. (Attachment 14 to FCP 377652-000) 4.7. Exelon Design Analysis OPTiMA2-TRO38QC-EOP, Revision 1, "SVEA-96 OPTIMA2 Fuel Input to the Emergency Operating Procedures," dated 12/18/06. (QC record available via EDMS) 4.8. Westinghouse Report NF-BEX-10-162, Revision 1, "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22", dated 11/18/10. (Attachment 18 to FCP 377652-000) 4.9. Exelon Design Analysis QDC-0000-N-1805, Revision 1, "Quad Cities 1 Cycle 22 MICROBURN- B2 Basedeck," dated April 2011.

4.10. Exelon Design Analysis QDC-0000-N-1 653, Revision 1, "Quad Cities 1 Cycle 21 Core Loading Plan," dated 5/11/09.

4.11. TODI QDC-11-025, Revision 1, "Q1 R21 Control Rod Blade and LPRM Replacements,"

dated 4/21/11.

Exekrn. Nuclear Fuels - BWR Design QDC-0000-N-1 804 Rev. 0 Nuclear Quad Cities Design Analysis Page 5 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 5.0 Identification of Computer Programs Table 1 - Computer Codes Used In Analysis Cksum Name Ref DTSQA Info IiP/UX 11.1 CAKWFT (1.4) N/A for PC See DTSQA Database EX0006886 Level CC 6.0 Method of Analysis This Design Analysis was prepared and reviewed in accordance with the governing procedure CC-AA-309 (Reference 4.1) and the associated T&RM CC-AA-309- 1001 (Reference 4.2). The Core Loading Plan Generation T&RM was utilized to create the CLP. This procedure can be found in Reference 4.3.

ExeIn. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page 6 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 7.0 Numeric Analysis Core Average kW/ft This value is the rated core power (2957 MWt) divided by the summation of all the linear ft of fuel pellet stack in the core. The calculation includes part length correction as shown in the table below. The table also shows the linear ft calculation for each of the different bundle types loaded in the core by cycle.

The Gad rod correction is not required for Optimal fuel. This correction is for fuel types in which the Gad rods have a shorter length than full length rods, which must be accounted for in the linear ft of fuel calculation (GE 14 fuel has this characteristic). This cycle will contain all Optimal fuel, which contains Gad rods that -hayd do not have a

^sts^alu unique length.

sqj SAM Information for input into CAKWFT 1.4 was obtained from Reference 4.6.

Table 2 - CAKWFT 1.4 Output Unit/Cycle Quad Cities nit J. Cycle 22 Fuel Type ODt mat Oates O-Number of Bundles 236 268 220

  1. Total rods per bundle 96 96 96
  1. Full length rods 84 84 84
  1. Gad rods # N/A #N/A #N/A
  1. Short Part length rods 4 4 4
  1. Long Part length rods 8 8 8 Full length rode, In. 145.28 145.28 145.28 Gad rod length, In. 145.28 145.28 145.28 Short Part length rods, in. 50.39 50.39 50.39 Long Part length rods, in. 99.61 99.61 99.61 Total fuel length, In., per bundle 13201.96 13201.96 13201.96 Total fuel length, In. (by fuel type) 3115662.56 3538125.28 2904431.2 Total fuel length, ft. (by fuel type) 259638.5467 294843 .7733 242035.9333 Total fuel length, ft. (for the core) 796518.2533 Rated thermal power (MWth) 2957 Core Average KW/ft Core Average kW/ft 3.712 kW/ft

Exe1n. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page 7 of 7 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN EOP Information Bundles mass = (Mass of U02 + Mass of Clad/channel)/# of assemblies For Optimal, using Reference 4.7, Table 1 values:

Optimal bundles mass = (316877 Ibm + 134824 Ibm)/724 = 623.9 Ibm Total mass of uranium is calculated per Reference 4.3. Total mass of U02 value is based on full core of Optimal fuel per Reference 4.7, Table 1:

Total mass of uranium = total mass of U02 / ratio of the molecular weight of U02 to U 316877 Ibm / (270/238) 279321 Ibm Core weight is calculated per Reference 4.3. The core weight value is based on full core of Optimal fuel per Reference 4.7, Table 1:

Core weight = total mass of uranium in Ibm converted to MTU

= 279321 Ibm x (1 ST/2000 Ibm) x (1 MT/1.10231 ST)

= 126.70 MTU 8.0 Results Attachment A contains the Core Loading Plan for 01 C22.

9.0 Conclusion Using the inputs from Section 2 and the Numeric Calculations from Section 7, the Core Loading Plan document was created as Attachment A of this Design Analysis. The data presented in Attachment A meets the requirements of T&RM NF-AB-1 30-2200, Revision 1.

ExeItn_ Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page Al of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Attachment A Core Loading Plan

ExeI^n- Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A2 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN REVISION 0 This Design Analysis section contains the Core Loading Plan as specified by NF-AB-130-2200, Revision 1, "Core Loading Plan Generation".

Revision Status: 0 Revision Notes: N/A CLP Distribution:

  • Dave Schumacher (QC Reactor Engineering Manager)
  • Mike Priaulx (QC Unit 1 Lead QNE)
  • Jill Fisher (NF Manager BWR Design)
  • Chuck Alguire (QC Manager Mechanical Design)
  • Benone Lohan (Westinghouse)
  • Charles Kuebel (QC Operations Staff)

Exek,n.. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A3 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Table of Contents Item of Interest: Page Fuel Bundle Inventory A4 As-Built Core Weight A5 Core Average kW/ft A5 Fuel Assembly ID Loading Map A6 Fuel Type Loading Map A8 Control Blade Inventory A9 Control Blade Identification Array Al0 Emergency Operating Procedure (EOP) Information Al 1

Exe1tn. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A4 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Fuel Bundle Inventory Bundle No. of Fuel Bundle Description Cycle Initial Batch lD Range Type Bundles Loaded 25-29 56 Opt2-3.99-15GZ8.00-3G6.00 (QA20) 20 QAAOO1 -QAA056 Cycle 20, Non-control - Bundle Type # 25 Cycle 20,713 Locations - Bundle Type # 26 Cycle 20, 7C locations - Bundle Type # 27 Cycle 20, 9C locations - Bundle Type # 28 Cycle 20, 10A locations - Bundle Type # 29 30-36 114 Opt2-4.00-13GZ8.00-3G6.00 (QB20) 20 0AA057-0,AA192 Cycle 20, Non-control - Bundle Type # 30 Cycle 20, 7C locations - Bundle Type # 31 Cycle 20, 8A locations - Bundle Type # 32 Cycle 20, 8B locations - Bundle Type # 33 Cycle 20, 9B locations - Bundle Type # 34 Cycle 20, 913 locations - Bundle Type # 35 Cycle 20, 108 locations - Bundle TWe # 36 40-44 66 Opt2-4.05-12GZ7.00-2G6.00 (QC20) 20 QAA193-QAA260 Cycle 20, Non-control - Bundle Type # 40 Cycle 20, 7A locations - Bundle Type # 41 Cycle 20, 8A locations - Bundle Type # 42 Cycle 20, 8B locations - Bundle Type # 43 Cycle 20, 10C locations - Bundle Type # 44 4-10, 148 Opt2.3.98.18GZ8.00 (QD21) 21 QABOO1-0AB148 12-14 Cycle 21, Non-control - Bundle Type # 4 Cycle 21, 98 locations - Bundle Type # 5 Cycle 21, 9C locations - Bundle Type # 6 Cycle 21, 9D locations - Bundle Type # 7 Cycle 21,10A locations - Bundle Type # 8 Cycle 21,108 locations - Bundle Type # 9 Cycle 21, 7A locations - Bundle Type # 10 Cycle 21, 78 locations - Bundle Type # 12 Cycle 21, 7C locations - Bundle Type # 13 Cycle 21, 8A locations - Bundle Type # 14

xe1c'n- Nuclear Fuels- BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A5 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN 45-47 64 Opt2-3.99-16GZ8.00 (QE21) 21 QAB149-QAB212 Cycle 21, Non-control - Bundle Type # 45 Cycle 21, 7C locations - Bundle Type # 46 Cycle 21, 8B locations - Bundle T ype # 47 21-24 56 Opt2-4.01-14GZ6.00 (0F21) 21 QAB213-GAB268 Cycle 21, Non-control - Bundle Type # 21 Cycle 21, 91) locations - Bundle Type # 22 Cycle 21, 10C locations - Bundle Type # 23 Cycle 21 813 locations - Bundle Type # 24 60 104 0p12-4.07-19GZ7.50/5.50 (0022) 22 QAC001 -QAC104 Cycle 22, Non-control - Bundle Type # 60 80 56 Opt2-4.07-17GZ7.50/5.50 (QH22) 22 QAC105-QAC160 Cycle 22, Non-control - Bundle Type # 80 90-92 60 Opt2-4.12-12G5.50-2GZ5.50 (0122) 22 QAC161 -QAC220 Cycle 22, Non-control - Bundle Type # 90 Cycle 22, 9D locations - Bundle Type # 91 Cycle 22, 10C locations - Bundle Type # 92 As-Built Care Weight Quad Cities Unit I Cycle 22 124.83 MT1J' Core Average kW/tt Rated Thermal Power = 2957 MWth Quad Cities Unit 1 Cycle 22 3.712' kW/ft For Use with INPO Fuel Reliability Index Calculation

Attachment 1 to Q1C22 CLP Nuclear Fuels - BWR Design Quad Cities Unit 1 Cycle 22 Care Loading Plan Revision Page QDC-0000-N-1604 0 A6 of A11 Fuel Assembly ID Loading Ma p 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 OAA033 QAA248 CM177 QAAOSS QAA161 QM049 QAAOBI OAA201 OAA225 QAA025 QAA017 QAA169 OAA041 QAA075 QM153 OAA209 0M217 QA8001 Q46009 QM009 QAA003 0M233 QAA057 QAA129 QM113 QAA123 QA8253 QAC181 0A0133 QAC169 QM257 OAAOB7 OAA241 QAA193 QA8125 QABO37 QACI77 QAC185 QAB281 QACIOS 0AB061 QAAOOI OAA085 QA1097 0M185 QABIO1 QAC193 QAC201 QAC113 QABO29 OAC121 Omni QAC129 QM171 QAA235 QM243 QAA187 QMO99 QAB181 OAC209 QAC001 OAB189 OAB205 OAC009 QAC137 QABIO9 0AA043 0AA059 QM195 Q3.B103 QA8183 0AC217 QAC017 QAB023 QABO77 QAB173 QAB053 QAC025 OAB237 QAA073 OM131 QAB12T QAC195 QAC211 QAC019 QABO17 QACA33 QAG145 OAC041 QAB213 QAC153 QA8045 QAA051 QAA155 QAA115 QABai9 QAC203 CA0003 QAB021 QACO35 QA8141 OAC049 QAB163 OAC057 QABO71 QACOBS QAA035 QAA083 QM139 0AA121 QAC119 QAC115 QA8191 OABO79 OAC147 QAC051 OA8119 QAa245 QA8085 OAC073 QA8197 OAA251 OAA203 QAA219 QA8255 QAC157 QA8031 QAB297 OA8175 0AC043 QAB167 OAB247 OAB117 QACO81 0A8229 QA8093 QM179 QAA227 QAB003 QAC163 QAB283 QAC123 OAC011 QABO55 0AB215 OACO59 QAB087 QAC083 QAB1S7 OACO89 QA8149 QM091 OAA027 QA8011 QA8135 QAC107 0A8223 OAC139 QAC027 QAC155 QABOB9 QAC075 OA5231 QAC091 QABI59 QAC097 QM163 QM019 QMOII QAC171 QA8083 QAC131 QAB111 064,0239 QA6047 QAC067 QAB199 QA809S 0A8151 QAC099 QA8143 am165 QMB21 0M013 QAC173 QA8085 QAC133 OAB113 QAB241 OABO49 QAC069 GAB201 QASO97 QAB153 QAC101 QA8145 OM093 QM029 OABO13 QAB13T QACIO9 QA8225 QAC141 QAC02B QAC157 QA5075 QAC977 QA8233 QAC093 QABI81 QAC1O3 OAA151 QAA229 QAB005 QAC165 QAB265 OAC125 QAC013 QABO57 GAe217 QACO61 OABO89 CAM 85 QAB163 QACOOS QAB155 OAA253 0M205 QAA221 QA8257 QAC189 QABO33 12AB209 OA8177 QACO45 QA8169 OA8249 QA8123 OAC087 OAB235 QA8099 OAA037 CAA095 QAA141 OM127 QAC181 QAC117 0A8193 QAB081 QAC149 QACO53 QA8121 0A8251 QABO91 QAC079 QA8203 OAA053 QAA157 OAA117 W0&1 OAC205 QACOOS QAB027 OAC037 QAS147 OACOS5 OA8171 QAC063 QABO73 OAG071 OAA079 OAA133 OA8129 QAC197 QAC213 QACO21 OABO19 OAC039 QACI51 OAC047 OAS219 QAC1S9 OAS051 OAAO45 QAAO81 QAA197 0*8105 0A5185 OAC219 QAC023 QABO25 0A8083 0A8179 0A8059 QAC031 0A9243 OM173 QAA237 OAA245 QM189 OM101 0A8187 OAC215 QA0007 0A5195 QA8211 OAC015 QAC143 0A8115 00007 0AA071 QM103 QM191 CAB107 QAC199 QAC207 OAC119 0A5035 0A0127 QA8227 OAC135 QAA259 OAA099 OAA247 OM199 CA8131 0A8043 0AC183 QAC191 0A8267 QAC11I 0A8087 QAA005 QAA238 QAA063 QAA135 QM119 OAA125 QA8259 QAC167 0A8139 QAC175 4M175 OAA047 00077 QAA159 0A9215 GAA223 0A8007 0A5015 0AA015 QAA055 QAA087 OAA207 QAA231 OAA031 OAA023 0A9039 0AA255 QAA183 OAA095 0AA187

Attachment 1 to 01 C22 CLIP Nuclear Fuels - BWR Design 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 Legends Center ASYID

Exel 6 n_ Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A8 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Fuel Type Loading Map 1 1 3 5 7 9 11 131161 17 19 21 23 25 27 29 31 33 35 37 39 41143 1 45 47 49 51 53 55 57 59 28 42 30 35 34 34 35 30 42 0 60 59 15 58 29 30 40 43 25 25 25 25 43 40 30 130126130 30 40 44 13 9 27 27 9 13 44 40 56 55 14 64 43 31 30 30 36 24 90 10 92 92 10 90 24 36 31 43 1 25 40 8 4 90 90 22 4 4 80 0 4 8 40 40 33 1 411 52 51 13 50 5 90 90 80 9 80 21 80 80 21 80 9 80 90 5 30 3033 25 19ii 30 S 90 60 45 45 60 80 80 45 90 45 0 30 40 43 47 12 46 26 91 60 4 4 45 12 601 24 24 60 12 45 4 4 60 91 45 5 40 31 44 60 7 60 80 60 21 80 6 80 21 80 7 60 90 90 8 30 30 4 42 29 30 30 14 601 1E 14 60 47 60 4 60 4 60 90 4 30 30 29 4 40 39 10 28 30 90 80 FIKTTJ 60 60 2 14 5 60 14 21 4 60 80 4 45 80 90 3fi 40 30 28 401 38 42 40 44 24 90 9 45 45 60 47 23 4 4 23 60 4 21 47 60 45 45 9 90 24 44 40 42 38 38 35 9 30 43 13 90 22 84 60 12 21 60 W46 60 46 46 60 6 6 -4 3 - 3 6 -

34 35 25 10 80 21 80 60 60 14 46 60 60 60 23 60 14 80 60 80 21 80 10 9 25 35 3 34 25 92 60 4 60 4fi 4 45 60 6 24 4 80 4 92 27 25 34 3 31 8 34 25 27 92 4 80 4 24 6 60 45 4 46 601 4 4 60 46 4 45 60 6 24 4 60 4 92 27 25 34 30 28 27 7 35 25 9 1 10 1 80 121 8F) 80 14 601 23 60 46 1 60 60 60 14 80 60 80 21 80 10 9 25 35 26 26 130143 13 90 22 80 60 12 21 60 14 60 46 60 46 0 46 6 0 14 60 21 12 60 80 22 90 13 43 30 26 24 140 44 2 90 9 45 60 47 21 60 0 23 60 2 47 60 45 45 9 90 24 44 40 42 2 23 6 28 30 30 36 90 801 45 4 80 60 4 21 14 60 45 45 60 14 21 4 60 80 4 45 80 90 3fi 40 30 28 22 29 30 30 4 90 60 60 4 60 47 60 14 60 60 14 60 47 60 4 60 4 60 90 30 29 20 18 19 5 30 8 90 90 60 7 60 80 60.21 80 6 6 80 21 60 80 60 7 60 90 90 8 30 30 18 26 31 40 5 45 60 4 4 45 12 60 4 24 60 2 45 4 4 60 91 45 5 4 40 31 26 16 14 15 4 43 40 30 30 45 90, 60 45 45 60 80 4 4 80 60 45 45 4 30 5 40 43 30 12 25 33 30 90 90 60 9 80 21 80 80 21 0 90 900 90 5 30 33 25 12 0- 11 3 r17 41 33 40 40 8 4 90 90 22 80 4 4 80 0 90 4 8 40 40 33 41 10 7 2 25 43 31 36 24 90 10 92 92 10 90 24 36 30 30 31 43 25 8 0 26 30 30 401 13 9 27 27 9 13 40 30 30 261 8 44 3 1 0 30 40 43 25 25 25 25 43 40 30 29 4 2 28 42 30 35 34 34 35 30 42 28 2 A Iq D E F 3 H iC L M N P R 2 6 18 22 26 34 38 42 50 54 58 4r2of t j

, 3

Exel6 n - Nuclear Fuels - SWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A9 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Control Blade Inventory INITIAL BLADE IDENTIFIER RANGE PPLX NO. OF CONTROL BLADE BLADE BLADES DESCRIPTION TYPE NO.

CX019R, CX062R, CY016R, D0075 1 4 GE D-100 (GE STD) 0077 to 0165, C0140 1 9 GE D-120 D104 to D691 2 15 GE D-230 M532 to M534, M545 to M551 3 10 GE Marathon with Etch Indication M1061 to M1432, M1537 to M1658, 3 42 GE Marathon M1869 to M1878 AA050 to AA078 1 4 Westinghouse (ABB) CR-82 02-431 to 02-776, 02-1200 to 02-1351 1 93 Westinghouse (ABB) CR82M-1

Exekn. Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Desi gn Analysis Page Al 0 of All QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Control Blade Identification Array TOOL No. ODC-11-025 Rev. I Page 3 of I

Q1C22 Blade Map 2 6 10 14 18 22 26 30 34 38 42 46 50 54 58 ABS02 #8882 02-63 Maft1 1RB882 59 59 14551 02-609 02-617 D2-606 1102.661 A8582 GESTD A8802 ABB62 Mann 2 8882 GE230 rii 18882 55 55 AA0 76 D0075 02-462 02-705 141330 07-696 02139 D579 02-776 Marie A8882 A8892 ABB82 ABB82 ABea2 A8882 882 Martn U18Ba2 AB882 51 51 02-768 02-1204 141328 02-1205 02-662 M533 02-704 02-455 02-695 02-607 02-463 Mann GESTD 47 47 Cx062R M532 Mann 888 2 IMartri (#8882 Mam GE230 Mann 18882 Mann ABM X48862 GE230 #8802 43 43 02-457 0609 02-1242 02-431 M1657 0E230 48882 Mam 882 Marts 06120 Marts 06120 Mann 82 ann 8882 0882 ast 39 39 48052 2- 772 2-452 348 02-630 0304 02-693 141873 02-1203 M1427 0092 147062 0387 I 141072 02- 1244 41876 48887 . #0882 36230 Marro Marro Marro A8862 Mann 06120 Mann A8882 Mann 48882 GE37D 35 35 GE230 02450 M1068 2449 YO16R 0499 02 -440 D2- 701 0690 141070 14549 141073 02-1202 M1061 C0140 M1323 2

A^BWB Marts 82 Marls 06120 Maim A19B82 Mann #8882 Mann 06120 Mann A8882 Mann 82 31 31 41658 2-702 2-1339 02-632 141329 02-703 M1538 0165 14595 02 459 M546 02- 775 M547 079 1428 2 #8852 2 882 14ann A8882 862 62 Maim 2 Marta 682 #8882 682 82 27 27 02-453 02-461 02-699 02-059 147326 02- 1207 02- 773 02-1201 M1655 112.1206 147430 02-451 02-700 02-664 3 2 06230 AW82 Marls 08982 Marts GE12D Mann GE120 Mann 2 MaM Afl882 GE230 2 23 23 02-616 02-633 D303 02- 760 141870 02.1200 141653 0D79 141064 0003 141320 02-660 141878 02-7239 0202 GE230 09882 [ABN2 Mann ABB82 Mann iAB Mann GE230 Mann AB382 Maim AB802 19 19 Dt04 02-1340 02-446 141652 02-1351 M1066 02-1241 M1654 D580 M1656 02.769 M1327 02-634

  1. 8682 Mann Mann GE120 Marla IA6582 Maim A8882 Mann AB882 AB882 Maim GE230 15 15 02-1243 02.615 14550 0583 02-657 M534 M1071 0017 141869 02-1345 141069 02-447 M1063 1.8882 ASS82 GE230 46882 A8 A8882 882 8882 GESTD A8892 11 02.606 78 0494 02-774 02-767 02-766 02-698 02-770 02-456 CX019R AAB50 j0BB82 #8882 GE230 802 Mann A8882 [AM 82 r 82 GE230 7 7

02-1240 02-443 0306 02.663 141429 02-697 02.448 02-1344 D581

  1. 5882 46882 06230 A8882 [ASUB2 82-610 02-1219 D6911 02-635 02.636 2 10 14 18 22 26  : 30 34 38 42 46 50 54 58 V/t 'Ile 2J6 s/^/ I w

i Exel Vf_ Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page A11 of Al 1 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN EMERGENCY OPERATING PROCEDURE (EOP) INFORMATION Bundle Type Mass Ibm Westinghouse Optima2 Bundles 623.9 Parameter Value Number of Westinghouse Optimal Bundles 724 Optimal Active Fuel Length (in.) 145.28 Core Weight (MT) 126.70 Total Mass of Uranium (Ibm) 279,321 Total Mass of U02 (Ibm) 316,877 Total Mass of Clad and Channels (Ibm) 134,824

Exekn- Nuclear Fuels - BWR Design QDC-0000-N-1804 Rev. 0 Nuclear Quad Cities Design Analysis Page B1 of B14 QUAD CITIES UNIT 1 CYCLE 22 CORE LOADING PLAN Attachment B Independent Review Information

aC.-0000-tJ -1$o P,ei 0 AHuc.ti ^,+ 8 C'

2. o 4 i e) k '

NF-AB-130-2200 Revision 1 Page 5 of 16 ATTACHMENT 2 Design Verification Guide - Core Loading Plan Page 1 of 5 NY-

1. DOCUMENT NUMBER QOC -°°°" N'ig°y REV NO. 0 (if applicable) a ih LIST OF PROCEDURES AND T&RMs USED The following Is a list of Procedures and/or T&RMs that were used In the performance and verification of the analysis:

PROCEDURE!

T&RM _ REVISION BTU NF-AB-130-2200_ I Loading Core Pattern Generation CC-AA-3a9 10 Gan^rdl 01 tk s cr A nv ly SeS CC-AA - 3W-ici to (yuiictincs ^rtj'cc ;on ar.A rbLxss. +^

OZ i's. c.,.5n A+.41 ys c^

t__--AIN-135-1106 a

  • 3u)R co..i,-di Qi%aie., t.:L4t1;^w. f4 L u i ..t-Note: Procedures and T&RM not used should be lined out.

III. PROCEDURE I T&RM DEVIATIONS The following deviations and associated justifications from the Procedures / T&RM listed in Section II were used for this analysis (mark N/A if no deviations were used):

N /A t)ew :t .t.Sior+S 40 i'ti+M p".. oeLs ,rVM Pst .n*o x^tecir o*+tfe prff'r ko.L L,.LA th grrca..'1 iht .^+w . i:s.on :4 bJF - A6- "W-21c* dOCS n.4 4-* "- 41.4 rti.if% &(- +k+{

.1 -- Cr- tte tallow:

h..ry¢ L 'mss (-%.p . It ^ Lr lty ^l:.;^^

^ibLitl^..

^uJsGLlritd a.4 c. tits q s +a+.s1+JSs'S

'4444 CA.? [.^ c. LItS**n k...Zl^ ,S* 1S 1L-h LaiaY.^k'0.17^t M1^2} ^dl eQf' a6G.,.,l..fm^:nn>>

t;NL.w 'j,)'1Me [.,^ .u^.gLi Usa.^^ F.rt1 f bl+. d^ ^n.^ra { :^Cv^ti4tCLr .2.L u+s^ . r l- t c$v:r^^d  :. tt^Q t%i.[1^ JL e L_P L ._.Sr c.

QJ^ 1-15 tti.+

. 1+ L t*n tis 44 rJ-.4* .. 4f n..ai .., :.. i ^.y C 4. ^

j +s ne..^ ra.r:s.o.n -M Nt=-A$ % i iH{0 d4[] w. tU. rtlr 4s rt}U{!y ++C i' :S ,c u+^-tr+#- ijCt.Laie

  • r 1

r' C rrClc

.( C1a `*'^ ^sr^.ati8**

+e Grt bL. .(}liq ZQ VMo.31310 0 513/i f

Qc - cxx O - is - I E by Rw a a .tk^.t.. S NF-AB- 130-2200 Revision 1 Page 6 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 2 of 5 IV. CHECKLIST Independent Core Designer Item for Verification Reviewer Indicate V. N, or N/A and Initial SECTION A - Verification Checklist is the full core loading pattern Identical to the latest map sent to site Reactor Engineering for generation of the Core Component Transfer n 0-Al Auth or izati on Sh eets ? f ¶y'i A2 Are the fresh fuel types Identified by actual handle Identifications? lbc.'t' Y

Do the fuel types and quantities match those described In the N e^.+cti  :, s }.?

Reference Loading Pattern document or subsequent revision? G.,.. g*+ D f

A4 Do the number of bundles add up to a full core size? B 6A D Wii an in-core shuffle be performed during the refueling outage? If no 1t g, D A5 (i.e., full core off-load), beginning-of-cycle shutdown margin data may 1 `-ef '.^ '

need to be incorporated Into the CLP.

A6 Do all pages contain the correct revision number of the report? J D 'l Al Is the control blade Inventory table correct? Y Is the control blade Identification array Identical to the latest map W lr^ 8n^t^ N//'t- c3 A8 sent to site Reactor Engineering for generation of the Core Component Transfer Authorization Sheets?

xx, A9 is the EPG/SAG Information correct? BMD Y Ili

.)

All} Is the core average kW/ft calculation correct? dMO Y U-3

- Rl l ^rovr ^i k . T49 t"d ie Sw"^S MPt^4 {^r^. '(p . W ' 44 1wre'.

(p(- -Dodo-V-a- ttc1 R,Lv d NF AB %30-2200 Revision I Page 7 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 3 of 5 IV. CHECKLIST (Cont'd.)

Core Designer Item for Verification Indicate Y, N, or N/A and Initial SECTION B - Documentation Are all of the following included in the Core Loading Plan:

"Full core" core loading pattern with actual handle identifications for B1 flesh fuel; I Quantity, description, NFT number, handle ID range, and cycle loaded B2 for each unique bundle type in the core, as modeled by the reactor simulator code; 83 I Core weight In appropriate units?

SECTION C - Additional Questions

QDC-0000-N-1804 Revision 0 Attachment B B5 of B14 N F-AB+130.2200 Revision 2 Page 8 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 4 of 5 V. DESIGN VERIFICATION COMMENT SHEET Review ofQDC-0000-N-1804 Page 1 of 1 Document In Review: ODC-0000-N-1804 Quad Cities Unit 1 Cycle 22 Core Loading Plan Responsible Engineer Brandon De Graaf/ Seth Spooner Independent Review Engineer. Eric Bruch

  1. Independent Review Comment Independent Responsible Engineer Response Review Closure Page 2, 9.0 Conclusions should read Page 2 changed to 'Conclusion'. Concur.

'Conclusion', or the title on page 7 SJS 04/22/11 EB 4/28/11 1 should be adjusted. This only needs to be fixed if either page must be redone anyway.

E B 4120/11 Section 2.1, no mention is made for a Text added to mention suspected fuel Concur, reference of where the 2 fuel shuffles failure in 01 C21 and Reference 4.9. EB 4/28/11 2 came from. Indicate the reference SJS 04/22/11 and that the rationale behind the moves is explained there.

ES 4/20/11 Section 2.2, the calculated core This is an input as it must be listed in No need to yell.

weight should not be included in this the results and is coming from an ES 4/28/11 3 section since it is for the 1nyuta only. external analysis. No change made EB 4/20/11 to address this comment.

SJS 04122/11 Reference 4.7 is actually an Exelon While this is both an Exelon andda Concur.

Design Analysis. Change the wording Westinghouse document (with the ES 4/28/11 4 to indicate that. same name), it will be updated to ES 4/20111 refer to the Exelon reference for consistency.

SJS 04/22/11 Reference 4.8, add a 0 to make it Comment incorporated. Concur.

5 377652-000 to remain consistent. SJS 04/22111 ES 4/28/11 EB 4/20/11 Reference 4.11 updated for Revision 6 1 of Quad blade TODI. Concur.

SJS 04/22/11 EH 4118/11 Page Al0 updated for Revision 1 of Concur .

7 Quad blade TOOL SJS 04/22/11 E8 42111 8

9 10

NF-AB-130-2200 Revision 1 Page 9 of 16 ATTACHMENT I Design Verification Guide - Core Loading Plan Page 5 of 5 VII. SIGNATURES THE ACTIVITIES REVIEWED BY THIS DVG ARE COMPLETE, ACCURATE, AND ADEQUATE FOR APPLICATION TO:

(Plant ! Unit andlor Cycle)

REVIEW:

Erg c ay-UC At VERIFIER" SIGNATURE DATE APPROVAL:

MANAGER" SIGNATURE DATE

'THE PREPARER AND REVIEWER MUST BE DIFFERENT INDIVIDUALS

Proiect Name Responsible Engineer - Rev, Page Quad Cities 1 Cycle 22 Brian Henning 2 - 4127/11 1 of 5 Purpose and Scope Ensure that all recommendations made by the Core Manager, Core Designer, Independent Reviewer, etc. for the identified core reload design evaluated and resolved prior to the appropriate milestone (e.g. start up date). are 1Measurement Standard All recommendations resolved & documented.

Generic Analysis Checks Item Actio Responsible Independent Nu

_mber Engineer Reviewer 1 Is the responsible engineer qualified to perform the analysis? 8MC Y 615 2 Is the independent reviewer qualified to perform the analysis? y*v 63 T S

Project Name Responsible Engineer Rev. Page Quad Cities 1 Cycle 22 Brian Henning 2 - 4127/11 2 of 5 Generic Reload Items Resolution Completed Milestone Action Reference Responsible Independent Reviewer byiate Cod e Engineer Not to be All Work Design Analysis/Product page numbering and IR 521107 6c^,G closed. formatting are in compliance with the $i3 Applicable requirements of CC-AA-309 and CC-AA-309-to all new 1001 or other CC process. See IR 521107 for rI items.

further information (internal OPEX).

Rod Ensure that the axial power peaking for hot Q1 C18 Transient WA SO -V Patterns, cases does not exceed 1.8 for Quad Cities and Selection Exp Acct Dresden due to a constraint on the LOCA /A analysis.

CMR, Ensure that the core average void fraction for GE-NE-A2200103- N1A BSA Exp Acct Quad Cities units is less than 48.3%. If this is 56-01-Q Rev. 0 y /'\

not satisfied, Tech Support should be notified. it (Task Report 0611)

PPLX POWERPLEX Input Deck IR 662389 t3/A 9. b A typo or incorrect print of the LHGRFAC table heading in the "SVL" file generated by the POWERPLEX-III USEP05 version was identified. When the "SHARE LIMITS" card is used in the POWERPLEX-III input, the heading "LHGRFAC FOR FUEL TYPE" in the "SVL" file should be "LHGRFAC FOR LIMIT TYPE" instead. This has no impact on calculations.

MYH

Proeect Name Responsible Engineer Rev. Pane Quad Cities 1 Cycle 22 Brian Henning 2-4/27/11 3of5 Resolution Completed Milestone Action Reference Responsible Independent Reviewer by/Date Engineer Code Keep until MB2/ An error in AREVA document EMF-2480. IR 679312 A g,r k, the error Nl PPLX "Maneuvering and Conditioning Criteria for is ATRIUM-10 Fuel (REMACCX -10) to Use with corrected POWERPLEX-III," Revision 3 was identified.

in the The unit for the parameter "beta" in Equations EMF- (3.5) and (3.12) is incorrect. It should be in units 2480 of kW/ft, instead of MWd/kgU. (MYH) document Deck, In EMF-2147, "MICROBURN -B2 User's IR 637562 t l^fl^ B. D PPLX- Manual," Revisions 10 and 11 (Page 5-270), the IIl/M-B2 default value for EDGE_ DECQNDITIONING keyword is I. This is correct in the Default field, but is incorrect in the Description field. The f c3 t'

current description states: if this keyword is not ` 6 31 input, the deconditioning exposure increment will default to the nodal average." The correct description should read, if this keyword is not input, the deconditioning exposure increment will default to the rod nodal ex posure." MYH)

Powerplex AREVA Condition Report CR 2010 -6767 IR 01123765 Decks identified a code error in MICROBURN -B2 and Nf A 9- #-%Z>

POWERPLEX-i11. The error occurs when the OVERRIDE subgroup is used to force specific assemblies to use a different REMA ACCX type.

The use of OVERRIDE caused the incorrect calculation of the preconditioned state for those specific assemblies.

Do not use OVERRIDE feature of MICROBURN-82 and POWERPLEX - Ill until the error has been corrected.

Project Name Responsible Engineer Rev. Page Quad Cities 1 Cycle 22 Brian Henning 2 - 4/27/11 4 of 5 Resolution Completed Milestone Action Reference Responsible Independent Reviewer bMp_ate Code Engineer CBH, Ensure that the actual exposures at which each /A cp Exp. Acct, Optimal bundle type is controlled in Q1C22 are Y Target within the bounds of the assumed bundle Rod controlled exposure ranges specified in the Patterns, appropriate Westinghouse CBH evaluation. If EOC Optimal bundles are controlled for significant `jf Projection periods outside of these assumed exposure Iv ranges, then additional Westinghouse CBH evaluation may be required to support Q1C22 operation.

In exposure accounting, document this compliance by including a detailed table that summarizes the controlled periods of operation for each bundle type and the assumed ranges.

Consider updating this table after every se q uence exchan g e durin g the cycle, Target Ensure that Westinghouse performs ASI W/A at-l>

Rod compliance checks as soon as possible for any SKI Patterns, significant changes to the target rod patterns for 'N! t EOC the remainder of Q1C22. If the rod patterns Projection change significantly during Q1C22, then additional Westinghouse ASI compliance evaluation may be required to support 01 C22 operation.

ATI COLR Include WCAP-16081-P-A Addendum 1 in Q1 o 298639- COLR. Addendum supports low bundle mass NJ+^

S 07 flux correlation limits required for Optimal Nl bundles on the periphery.

MB2 Deck Include Reviewer comments on ASSEMBLY QDC-0000-N-1805 f l/H ef- D F_TYPE cards, Lattices 4 and 25, XMASS card, Rev. 0. cy /

FLOW_TEMP_F, and QAC bundle I NAME_RANGE card as described in the commens to QDC-0000-N-1805 Rev. 0.

Project Name Responsible Engineer Rev. Page Quad Cities 1 Cycle 22 Brian Henning 2-4/27/11 5 of 5 W .0 Q

n Resolution completed Milestone Action Reference Responsible Independent Reviewer by/Date Code Engineer ATI M132 Deck The Thermal Limit set for QDC-0000-N-1805 QDC-0000-N-1805 piA .-,

985968- POW ERP Rev. 1 is based on Rev. 0 RLR. ATI has been Rev, 1 -rL 4c# a.e> ^af^c4 49 LEX Deck created to verify TL set or implement changes. .;+.c c:.e .^rj+r*

TL data is preliminary until RLR Rev. 1 has been aa'- -OJ--ti v approved.

M82 Deck Update Steam Flow to FW temperature curve QDC-0000-N-1805 tt(A %" V POWERP and add comment card stating Lattice Types 4 Rev. 1. £-c' LEX Deck and 25 are not used In accordance with reviewer i^ /`

comments to QDC-0000-N-1805 Rev. 1.

G

OVC 0", ACV K:> ik 'to -u" Me.- 4 21 lL CC-AA-309-1001 Revision 6 Page 40 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 1 of 3 Purpose (Section 4.3.2.)

Are the statement of problem and purpose of Design Analysis sufficiently detailed to reveal how the Design Analysis will address, resolve, and approach the stated problem? yes inputs (Section 4.3.3.)

1. Are design inputs into the Design Analysis clearly identified and their sources listed?

yes

2. Was the input information obtained from the correct revision of the source document?

Yes

3. Are the design inputs relevant and directly applicable to the purpose of the Design Analysis? Yes
4. Are the inputs sufficient considering the purpose of the Design Analysis? Yes
5. Is the use of engineering judgment clearly documented and justified? Yee-,
6. Has all input data been used correctly? Yts
7. Do the sources of inputs used meet current technical requirements as committed under the site license and regulatory commitments? Vcs
8. Are the Critical parameters designated for ease of verification? r/A Assumptions (Section 4.3.4.)
1. Have the assumptions necessary to perform the analysis been adequately documented? yes tom)
2. Is justification provided for all assumptions (except those based upon recognized engineering practice, physical constants or elementary scientific principals)? $(A
3. Are they reasonable for the Design Analysis? N to
4. Where necessary, are the assumptions identified for verification when the detailed design activities are completed. If so, have Action Tracking Items been established?

,IA References (Section 4.3.5.)

1. Are applicable codes, standards and regulatory requirements, including issues and addenda, employed in the Design Analysis properly identified and were their requirements met? ye S
2. Do the stated references reflect the appropriate revision? Y e-S
3. Are references that are not easily retrievable included as an attachment? N/N
4. Have supporting technical documents and references been reviewed when necessary?

4/21/11 1 es

Rc^ l^ A 14+c-c^^^+,

'C,_ Oro - ^^ ^8d^1 CC-AA-309-1001 Revision 6 Page 41 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 2 of 3 Identification of Computer Programs (Section 4.3.6.)

1. Have the versions of the computer codes employed in the Design Analysis been certified for the application? Y<-5 Are 2. the codes used identified along with source, computer type, inputs and outputs?

yes

3. Is the code suitable for present analysis? 7c-,
4. Does the computer model (modeling time steps, etc.) adequately represent the physical SySLWWWTi:l td(LU$U WIlUIUUIIS) ( t es Is the output reasonable
5. when co iparedlto inputs? yes
6. Does the Computer Program conform with the NRC SER or similar document when applicable? z /A Method of Analysis (Section 4.3.7.)

Is the method 1. used appropriate considering the purpose and type of DesigAn alysis?

Is the method 2. in accordance with codes, standards, and regulatory requirements? if so are these codes, standards, and regulatory requirements referenced by number or title?

3. Has the method been employed elsewhere in industry or in license applications? Are these uses discussed? P 1A Are the methods4.used and recommendations given conservative relative to the design and safety limits? Yes Numeric Analysis (Section 4.3.8.)

Are 1. the equations used consistent with recognized engineering practice? YcS Is justification provided

2. for any equations not in common use? Is the j ustification reasonable? N t A
3. Have the adjustment factors, uncertainties, and empirical correlations used in the analysis been correctly applied? yeS Is the result presented
4. with proper units and tolerances? y S Are 5. any of the results overly sensitive to small changes in input? Do tolerance and conclusions consider this sensitivity? h}O, 0 fA
6. What checking method was used to review the Design Analysis?

^^- Detailed Design Review.

-- Alternate Calculation Review. 'Reviews 3 Qualification Testing Review.

Q tac - 0000 - - -

13 H 0C B1 1i CC-AA- 309-1001 Revision 6 Page 42 of 73 ATTACHMENT 4 Design Analysis Design Review Guidance Page 3 of 3 Results I Conclusion (Sections 4.3.9. and 4.3.10.)

1. Is the magnitude of the results reasonable? Ye-5
2. Is the direction of trends reasonable? YfS Are the conclusions
3. justifiable based on the results? Ye S
4. Have margin impacts been appropriately identified? a/A
5. Has an IR been initiated for any reduced margin that was identified during the preparation of the new/revised analysis? la /A Are the recommendations/results/conclusions
6. reasonable based on previous experience? y e.s
7. Has the effect on plant drawings, procedures, databases, and/or plant simulator been addressed? vi /A
8. Has the effect on other systems been addressed? i i / A Have any changes 9. in other controlled documents (e.g. UFSAR, Technical Specifications, M&TE selection sheets or procedures, COLR, etc.) been identified and tracked? g /A /Ei
10. When applicable, are the analysis results consistent with the proposed license amendment? k /11-
11. Have other documents that have used the calculation as input been reviewed and revised as appropriate? to lPr General
1. Are applicable pages properly numbered and marked with the valid Design Analysis number? `ye-<,

Is all information legible and reproducible? kc_S

3. Have all cross-outs or overstrikes in the documentation been initialed and dated In ink?

I" Have all affected4.design analyses been documented on the Affected Documents List (ADL) for the associated Configuration Change?

FGP 3-+ % 652-levie 412f/ti

ATTACHMENT 9 Westinghouse Affidavit for Bundle Design Report and Reference Loading Pattern Report

Westinghouse Westinghouse Electric Company Nuclear Services 1 000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@ westinghouse.com Rockville, MD 20852 Proj letter: NF-BEX-11-142 CAW-11-3248 September 14, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Attachment to NF-BEX-10-162, Rev. 1 "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) and Attachment 1 to NF-BEX 184 "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced reports is further identified in Affidavit CAW-1 1-3248 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

The subject documents were prepared and classified as Westinghouse Proprietary Class 2. Westinghouse requests that the documents be considered proprietary in their entirety. As such, non-proprietary versions will not be issued.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3248, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, J. A. Gresham, Manager Regulatory Compliance Enclosures

CAW-11..3248 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

J. A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this 141 day of September 2011 COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member. Pennsvivania Association of Notaries

2 CAW-11-3248 (1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-1 1-3248 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use 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 a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense.

4 CAW-11-3248 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390; it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is contained in Attachment to NF -BEX-10-162, Rev. 1 "Rev. 1 Bundle Design Report for Quad Cities 1 Cycle 22" (Proprietary) and Attachment 1 to NF-BEX 184 "Reference Loading Pattern for Quad Cities Unit 1 Cycle 22" (Proprietary), for submittal to the Commission, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.

This information is part of that which will enable Westinghouse to:

(a) Support Exelon's use of Westinghouse Fuel at Quad Cities.

5 CAW-11 3248 (b) Assist the customer to obtain license change.

Further this information has substantial commercial value as follows:

(a) Westinghouse can use this information to further enhance their licensing position with their competitors.

(b) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith is the proprietary version of a document furnished to the NRC in connection with requests for generic and/or plant-specific review and approval. The document is to be considered proprietary in its entirety.

COPYRIGHT NOTICE The report transmitted herewith bears a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in this report which is necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

ATTACHMENT 13 Westinghouse Affidavit for Quad Cities Unit 1 Cycle 22 SLMCPR Report Revision 2

Westinghouse Westinghouse Electric Company Nuclear Services 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 720-0754 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 Proj letter: NF-IBEX-11-3 R.ev. 2 CAW-1 1-3242 September 12, 2011 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

NF-BEX-11-3 Rev. 2 P-Attachment, "Quad Cities Unit 1 Cycle 22 SLMCPR" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW-1 1-3242 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying affidavit by Exelon Generation.

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse affidavit should reference this letter, CAW-11-3242, and should be addressed to J. A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company LLC, Suite 428, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Very truly yours, J J. A. Gresham, Manager Regulatory Compliance Enclosures

CAW-11-3242 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared J. A. Gresham, who, being by me duly sworn according to law, deposes and says that he is authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief.

J. A. Gresham, Manager Regulatory Compliance Sworn to and subscribed before me this day of September 2011 COMMONWEALTH OF PENNSYLVANIA Notarial Seal Cynthia Olesky, Notary Public Manor Boro, Westmoreland County My Commission Expires July 16, 2014 Member. Pennsvivania Assodation of Notaries

2 CAW-11-3242 (1) I am Manager, Regulatory Compliance, in Nuclear Services, Westinghouse Electric Company LLC (Westinghouse), and as such, I have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

(2) I am making this Affidavit in conformance with the provisions of 10 CFR Section 2,390 of the Commission's regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

(3) I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

(4) Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commission's regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitutes Westinghouse policy and provides the rational basis required.

Under that system, information is held in confidence if it falls in one or more of several types, the release of which might result in the loss of an existing or potential competitive advantage, as follows:

(a) The information reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) where prevention of its use by any of

3 CAW-11-3242 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

(b) It consists of supporting data, including test data, relative to a process (or component, structure, tool, method, etc.), the application of which data secures a competitive economic advantage, e.g., by optimization or improved marketability.

(c) Its use 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 a similar product.

(d) It reveals cost or price information, production capacities, budget levels, or commercial strategies of Westinghouse, its customers or suppliers.

(e) It reveals aspects of past, present, or future Westinghouse or customer funded development plans and programs of potential commercial value to Westinghouse.

(f) It contains patentable ideas, for which patent protection may be desirable.

There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

(b) It is information that is marketable in many ways. The extent to which such information is available to competitors diminishes the Westinghouse ability to sell products and services involving the use of the information.

(c) Use by our competitor would put Westinghouse at a competitive disadvantage by reducing his expenditure of resources at our expense,

4 CAW- 11-3242 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

(e) Unrestricted disclosure would jeopardize the position of prominence of Westinghouse in the world market, and thereby give a market advantage to the competition of those countries.

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iii) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(iv) The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(v) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in, NF-BEX-1 1-3 Rev. 2 P-Attachment, "Quad Cities Unit 1 Cycle 22 SLMCPR" (Proprietary) for review and approval, being transmitted by Exelon letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the review of Quad Cities Unit 1 Cycle 22 SLMCPR, and may be used only for that purpose.

This information is part of that which will enable Westinghouse to:

(a) Support Exelon's use of Westinghouse Fuel at Quad Cities.

5 CAW-11-3242 (b) Assist the customer to obtain license change.

Further this information has substantial commercial value as follows:

(a) Westinghouse can use this information to further enhance their licensing position with their competitors.

(b) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar analyses and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and/or non-proprietary versions of documents furnished to the NRC in connection with requests for generic and/or plant-specific review and approval.

In order to conform to the requirements of 10 CFR 2.390 of the Commission's regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

ATTACHMENT 14 Quad Cities Unit 1 Cycle 22 SLMCPR, Revision 2 (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 Quad Cities Unit 1 Cycle 22 SLMCPR Westinghouse Electric Company Nuclear Fuel 1000 Westinghouse Drive Cranberry Township, PA 16066 C0 2011 Westinghouse Electric Company LLC, All Rights Reserved NF-BEX-11-3 Rev. 2 NP-Attachment

1.0 Introduction This document contains a description of the Safety Limit Minimum Critical Power Ratio (SLMCPR) evaluation for Quad Cities Nuclear Power Station Unit 1 (QCNPS 1) Cycle 22.

Dual recirculation loop operation (DLO) and single recirculation loop operation (SLO)

SLMCPRs of 1.11 and 1.14, respectively, have been calculated for the Westinghouse SVEA-96 Optima2 assemblies in QCNPS 1 Cycle 22. Application of the Westinghouse methodology in Reference 1 requires modification of the QCNPSI Technical Specifications, Reference 3, to support DLO and SLO SLMCPRs of 1.11 and 1.14, respectively, for the SVEA-96 Optima2 fuel in Cycle 22. The SLMCPR values for QCNPSI Cycle 21 and Cycle 22 are shown in Table 1.

For QCNPS 1 Cycle 22, Exelon Generation Company, LLC, (EGC) will load Westinghouse SVEA-96 Optima2 reload fuel for the third consecutive cycle. SVEA-96 Optima2 reload fuel was also loaded in QCNPSI Cycle 20 and Cycle 21. In Cycle 22, all remaining legacy GE14 fuel shall be discharged, resulting in a full core loading of SVEA-96 Optimal assemblies.

Therefore, the Westinghouse NRC-approved methodology described in Reference 1, and further clarified in the response to Request for Additional Information (RAI) D13 of Reference 2, was used to determine the SLMCPRs for Cycle 22. Further clarification of the Westinghouse SLMCPR methodology was also provided to the NRC in support of the transition to SVEA-96 Optimal fuel in the Quad Cities and Dresden Units as follows:

The response to NRC Request 19 in Reference 5 which supported the Licensing Amendment Request for transition to SVEA-96 Optima2 fuel in the Dresden and Quad Cities plants provided in Reference 4, The technical information supporting the Quad Cities Nuclear Power Station Unit 2 (QCNPS2) Technical Specification SLMCPR changes transmitted by Reference 6 as supplemented by the clarifying information in Reference 7.

The same SLMCPR methodology described in these references was followed to establish appropriate SVEA-96 Optimal SLMCPRs for QCNPSI Cycle 22.

The EGC proposed license amendment to use the Westinghouse methodology for core reload evaluations at the Dresden and Quad Cities units was submitted to the NRC in Reference 4.

This submittal was approved by the NRC, and supported QCNPS2 Cycle 19, Dresden Nuclear Power Station Unit 3 (DNPS3) Cycle 20, QCNPS 1 Cycle 20, and Dresden Nuclear Power Station Unit 2 (DNPS2) Cycle 21, all of which are cores containing a reload of SVEA-96 Optima2 fuel, and QCNPS2 Cycle 20 and DNPS3 Cycle 21 which contain two reloads of SVEA-96 Optima2 fuel.

Condition 7 in the NRC safety evaluation for Reference I requires that the conservative factor be applied to the GE14 Operating Limit Minimum Critical Power Ratio (OLMCPR) and that this factor be identified in licensee applications. Since no legacy GE14 fuel assemblies are Page 2 of 17 NF-BEX-l 1-3 Rev. 2 NP-Attachment

loaded into the QCNPS 1 Cycle 22 core, this factor is not relevant for this cycle and will not be included in the Cycle 22 Core Operating Limits Report.

a,c I

2.0 SVEA-96 Optimal SLMCPR for Cycle 22 The SVEA-96 Optima2 SLMCPR for QCNPS1 Cycle 22 is based on a Reference Core design (SVEA-96 Optima2 bundle designs, core loading pattern and state point depletion strategy) that represents realistic current plans for the Cycle 22 loading and operation. The Reference Core loading pattern for QCNPS 1 Cycle 22 is shown in Figure 1. The Reference Core design was generated via collaboration between EGC and Westinghouse based on EGC's cycle assumptions and design goals. The Reference Core was designed to meet the cycle energy requirements, to satisfy all licensing requirements, to provide adequate thermal margins and operational flexibility, and to meet other design and manufacturing criteria established by EGC and Westinghouse.

In general, the calculated SLMCPR is dominated by the flatness of the assembly CPR distribution across the core, and the flatness of the relative pin CPR distribution based on the pin-by-pin power/R-factor distribution in each bundle. Greater flatness in either parameter yields more rods susceptible to boiling transition and thus a higher SLMCPR.

The calculation of the SLMCPR as a function of cycle exposure captures the interplay between the relative fuel assembly CPR and the bundle relative pin-by-pin CPR distributions established from the power/R-factor distributions and allows a determination of the maximum (limiting) SLMCPR for the entire cycle. This limiting SLMCPR is conservatively applied throughout the entire cycle, The SVEA-96 Optimal SLMCPR for QCNPS 1 Cycle 22 was determined as a function of cycle exposure based on radial assembly power distributions with about the same "flatness" as the Page 3 of 17 NF-BEX-11-3 Rev. 2 NP- Attachment

cycle exposure-dependent radial power distributions from [

1 a,c Accordingly, the SVEA-96 Optimal SLMCPR for DLO was calculated at 100% power and 100% flow at various exposures throughout the cycle to assure that the limiting SLMCPR was identified. In addition, the DLO SLMCPRs were calculated at 100% power at the minimum allowed core flow at rated power (95.3% flow), as well as the maximum licensed core flow at rated power (108% flow) to confirm that a limiting SLMCPR had been established. Figure 3 shows a current QCNPS1 power-flow map which is applicable to Cycle 22. Consistent with Figure 3, a flow window of 95.3% to 108 % of rated core flow was analyzed.

SLO SVEA-96 Optima2 SLMCPR calculations were also performed. These SLMCPR calculations were performed at [

a,c I The SLO calculations used the same procedure as the DLO cases, except that the SLO cases applied a larger uncertainty for the core flow.

The SLMCPR results for Cycle 22 are plotted in Figure 4. As shown in Figure 4, the DLO SLMCPR [

I a,c Since the uncertainties at each DLO point are the same, this behavior is due to the interplay between the assembly relative CPRs and the relative fuel rod CPRs. In general, as the number of assemblies or fuel rods with CPRs in the vicinity of the assembly or fuel rods with the minimum CPR increases, the number of fuel rods with a potential for experiencing dryout increases. Therefore, a larger SLMCPR is required to assure that less than 0.1 % of the rods are in dryout.

Experience has shown that the assembly CPR distributions tend to become [

Ia,c Consequently, the peak SLMCPR tends to occur when the assembly CPR and rod CPR distributions combine to place the maximum number of fuel rod CPRs close to the minimum CPR.

This behavior is shown for the QCNPS 1 Cycle 22 SLMCPR by the relative assembly CPR histograms shown in Figures 5 through 7, respectively.

Inspection of the DLO histograms in Figures 5 through 7 leads to the following observations, which explain the SLMCPR behavior in Figure 4:

a,c I

Page 4 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment

Ia,c Therefore, the DLO SLMCPR results at rated conditions in Figure 4 can be explained in terms of [

1 a,c The adequacy of a DLO SLMCPR of 1.11 for the range of core flows at rated power permitted by Reference 3 was demonstrated by [

I a,c Therefore, the results in Figure 4 confirm the adequacy of a DLO SLMCPR of 1.11 for a QCNPS1 Cycle 22 flow window at rated power conditions of 95.3% to 108 % flow.

The SLO results calculated at [

I a,c Therefore, the results in Figure 4 confirm the adequacy of a SLMCPR of 1.14 for SLO.

The relative fuel rod CPRs in the SLMCPR calculations are [

a,c I

Page 5 of 17 NF-BFX-11-3 Rev. 2 NP-Attachment

la,c In addition to the strong dependence on assembly CPR and relative fuel rod CPR distributions, the SLMCPR is strongly dependent on the distribution of assembly and relative fuel pin CPRs about their mean values leading to an overall distribution of fuel rod CPRs relative to their mean values. The wider these distributions, the higher the SLMCPR must be to prevent 0.1%

of the fuel rods from experiencing boiling transition. The distributions of fuel rod CPRs relative to their mean values are determined by the uncertainties relative to the mean CPRs.

Accordingly, the uncertainties used in establishing the SVEA-96 Optimal SLMCPR for Cycle 22 are shown in Table 2.

3.0 References

1. Licensing Topical Report, Reference Safety Report for Boiling Water Reactor Reload Fuel, CENPD-300-F-A, July 1996.
2. CENPD-3 89-P-A, I Ox] 0 SVEA Fuel Critical Power Experiments and CPR Correlations: SVEA-96+,

August 1999.

3. Quad Cities Technical Specifications, Section 2.1.1.2
4. Letter, Patrick R. Simpson (Exelon Generation Company, LLC) to NRC, Request for License Amendment Regarding Transition to Westinghouse Fuel, dated June 15, 2005.
5. RS-06-009, Additional Information Supporting Request for License Amendment Regarding Transition to Westinghouse Fuel, January 26, 2006.
6. Letter from Patrick R. Simpson, Exelon Nuclear, to U.S. NRC, "Request for Technical Specifications Change for Minimum Critical Power Ratio Safety Limit ", QCNPS, Unit 2, December 15, 2005.
7. RS-06-024, "Additional Information Supporting Request for Technical Specifications Change for Minimum Critical Power Ratio Safety Limit ", QCNPS, Unit 2, February 13, 2006.

Page 6 of 17 NF-BEX-11-3 Rev. 2 NP- Attachment

Table 1 Comparison of Cycle 21 and 22 Cores Description Cycle 21 Cycle 22 Number of Bundles in Core 724 724 Limiting Cycle Exposure Point Near EOC Near MOC Cycle Exposure at Limiting Point, EFPH 16809 EFPH 8357 EFPH Reload Fuel Type SVEA-96 O tima2 SVEA-96 Optima2 Reload Batch Average Weight % Enrichment 3.99 w/o 4.08 w/o Reload Batch Fraction (%) 37% 30.4%

Batch Fraction of SVEA-96 O tima2 Fuel 72.9% 100.0%

Batch Fraction of GNF GE14 Fuel 27.1% 0.0%

Core Average Weight % Enrichment 4.02 w/o 4.03 w/o 1.11 for GE14, 1.11 Calculated Safety Limit MCPR (DLO) and SVEA-96 Optima2 SVEA-96 Optima2

1. 13 for GE 14, 1.14 Calculated Safety Limit MCPR (SLO) and SVEA-96 Optima2 SVEA-96 Optima2 Page 7 of 17 NF-BEX-11-3 Rev. 2 NP- Attachment

Table 2 - Uncertainties used in Quad Cities 1 Cycle 22 SVEA-96 Optimal SLMCPR Determination a,c Page 8 of 17 NF-BEX-1 1-3 Rev. 2 NP-Attachment

Figure 1 - Quad Cities 1 Cycle 22 - Reference Loading Pattern 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 Legends Center QA20 0C20 QBZO 0B20 Q620 ASYTYP QA20 0620 Qcza 0020 0A20 0A20 QB20 QA20 0620 QB20 0C20 0C20 0021 0021 0A20 QA20 QC20 QB20 Q620 QB20 0820 QF21 0122 0021 Q122 QC20 QB20 QC20 QC20 QD21 QD21 0122 0122 0221 0H22 0021 QA20 QB20 QB20 QB20 Q021 Q122 Q122 QH22 0021 QH22 QF21 QH22 QB20 QC20 QC20 QB20 QB20 QE21 Q122 QG22 0821 QE21 QG22 QH22 0021 QA20 QB20 QC20 QD21 QE21 Q122 QG22 QD21 QD21 0821 0021 QG22 0F21 QB20 Q620 01121 0122 Q122 QG22 QD21 QG22 QH22 QG22 0221 QH22 0029 0820 QD21 Q122 QG22 0021 0022 QD21 QG22 QE21 QG22 0021 QG22 QC20 0820 0122 QH22 QE21 3021 QH22 QG22 0021 0221 0021 QG22 QE21 0C20 QC20 0F21 Q122 QD21 0E21 QE21 QG22 QE21 0221 0021 0022 QF2I 0021 QD21 0122 0F21 QH22 0022 QD21 QF21 QG22 0021 0022 0E21 QG22 0E21 QD21 QD21 QH22 QF21 QH22 QG22 QH22 QD21 QG22 0221 QG22 QE21 QG22 QA20 Q122 Q021 QH22 QD21 QF21 QD21 QG22 0E21 0021 0E21 QG22 QD21 QB20 QA20 0122 QD21 QH22 Q021 QF21 QD21 QG22 0821 0021 0E21 QG22 0D21 28 0620 QD21 QD21 QH22 QF21 QH22 QG22 QH22 QD21 QG22 0221 QG22 QE21 QG22 26 Q021 Q122 QF21 QH22 QG22 Q021 QF21 Q022 0021 0022 0821 QG22 0521 24 0C20 0F21 0122 Q021 QE21 QE21 QG22 QF-21 QF21 QD21 QG22 0221 QD21 22 0C20 QB20 0122 QH22 QE21 0021 QH22 QG22 0021 QF21 0021 0022 0E21 20 0820 0820 QD21 Q122 QG22 QD21 QG22 QD21 0322 0E21 0322 0021 QG22 18 QB20 QB20 QD21 Q122 0122 0022 QD21 QG22 QH22 QG22 QF21 QH22 0021 16 QA20 QB20 QC20 0021 QE21 0122 QG22 QD21 0021 0E21 0021 QG22 0221 14 QB20 QC20 QC20 QB20 QB20 QE21 Q122 QG22 0E21 0E21 QG22 QH22 QD21 12 QA20 Q620 QB20 QB20 Q021 Q122 0122 01122 0021 01122 0221 01-122 Q021 Q021 QI22 QI22 QF21 QH22 0021 QA20 QC20 QB20 QB20 QB20 0820 0221 Q122 0021 0122 QA20 QB20 QB20 0C20 0020 0021 QD21 0A20 QA20 0820 0020 0C20 0A20 QA20 QA20 0020 0B20 0820 QB20 Page 9 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment

Figure 1 - Quad Cities 1 Cycle 22 - Reference Loading Pattern 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 Legends Center 60 0820 01320 0820 QC20 QA20 ASYTYP 58 QA20 QA20 QC20 QC20 QB20 QA20 56 0A20 QD21 0D21 0C20 QC20 QB20 Q620 0A20 QB20 0122 0021 0122 QF21 QB20 QB20 QB20 QB20 QC20 0A20 0021 QH22 0F21 0122 0122 QD21 QD21 QC20 QC20 Q820 QC20 QH22 0F21 QH22 0021 01122 Q122 Q122 QD21 QB20 Q820 Q820 QA20 0D21 QH22 QG22 QE21 QE21 QG22 0122 QE21 QB20 QB20 QC20 QC20 Q920 GF21 QG22 QD21 0821 0021 QD21 QG22 Q122 QE21 QD21 QC20 QB20 QA20 0021 01122 0F21 QG22 QH22 QG22 0021 QG22 Q122 0122 QD21 QB20 QB20 QG22 0021 QG22 QE21 QG22 QD21 QG22 Q021 QG22 Q122 QD21 0B20 Q820 QA20 0E21 QG22 0021 0F21 0D21 QG22 QH22 Q D21 QE21 QH22 Q122 QB20 QC20 QB20 QA20 QD21 0F21 QG22 QD21 0F21 QE21 QG22 QE21 QE21 QD21 Q122 QF21 QC20 QC20 QC20 QE21 QG22 QE21 QG22 QD21 QG22 QF21 Q021 QG22 QH22 QF21 Q122 Q021 QC20 Q620 QG22 0E21 RG22 QF21 QG22 QD21 QH22 QG22 QH22 QF21 QH22 QD21 Q021 QA20 Q620 32 QD21 QG22 0821 0021 QE21 0022 QD21 0F21 QD21 QH22 QD21 0122 QA20 0A20 0820 30 0D21 0322 0E21 QD21 0821 QG22 0021 QF21 OD21 QH22 Q021 QI22 QA20 0A20 Q820 28 QG22 QE21 QG22 QF21 QG22 Q021 QH22 QG22 QH22 QF21 QH22 QD21 QD21 QA20 QB20 26 0E21 QG22 QE21 QG22 0021 QG22 0F21 QD21 QG22 QH22 QF21 Q122 QD21 QC20 QB20 24 QD21 0F21 QG22 QD21 0F21 QE21 QG22 QE21 0E21 QD21 Q122 QF21 QC20 QC20 QC20 22 QE21 QG22 0D21 0F21 QD21 QG22 QH22 QD21 QE21 QH22 Q122 QB20 QC20 QB20 QA20 0322 0021 0022 QE21 QG22 QD21 QG22 QD21 QG22 0122 QD21 QB20 Q820 QA20 0D21 QH22 QF21 QG22 QH22 QG22 QD21 QG22 Q122 Q122 0D21 QB20 QB20 0F21 0322 QD21 0821 0021 QD21 Q022 Q122 QE21 QD21 0C20 0820 GA20 0021 01122 0322 QE21 0821 QG22 0122 0E21 QB20 QB20 QC20 QC20 OB20 QH22 0F21 01-122 0021 QH22 Q122 Q122 QD21 QB20 QB20 QB20 QA20 0021 QH22 QF21 0122 0122 QD21 QD21 QC20 QC20 QB20 I QC20 0122 0021 0122 QF21 0820 0820 QB20 0820 QC20 0A20 RAID 0021 0021 QC20 QC20 QB20 QB20 0A20 0B20 QA20 0A20 0C20 QC20 0620 QA20 0820 0820 0820 0C20 QA20 Page 10 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment

Figure 2 - Quad Cities 1 Cycle 21 - Reference Loading Pattern 01 03 05 07 09 11 13 15 17 19 21 23 25 27 29 Legends Center 60 2826 2825 2826 2825 2825 ASYTYP 58 2825 2825 2825 2825 2826 2825 56 2825 2825 2826 2825 2826 QB20 0B20 0A20 QA20 54 2825 2825 2826 2825 QA20 0B20 0B20 0F21 QF21 QF21 52 2826 2825 2825 2825 QA20 QF21 QF21 QC20 0E21 QE21 QC20 50 2825 2825 2826 QB20 0A20 QF21 QE21 0E21 0E21 0021 4820 QE21 2825 2825 2825 QB20 2826 QF21 QC20 QE21 0020 0021 0020 0E21 Q320 2825 2826 2826 QA20 QF21 QD21 QD21 QG20 0021 0020 0021 0820 0021 2826 2825 QA20 0F21 QC20 Q021 QC20 QD21 0520 QD21 0B20 0021 QB20 2825 2825 QA20 QF21 QE21 QE21 QC20 QD21 QB20 0D21 0820 0D21 0820 QD21 2826 2825 2826 QB20 QF21 QE21 0C20 QD21 QA20 QD21 Q620 0D21 0C20 0D21 0020 2825 2826 QB20 QB20 0C20 QE21 QD21 QC20 0D21 QB20 0D21 Q820 0021 0820 Q021 2826 2825 4B20 QF21 QE21 QD21 QC20 QD21 0520 QD21 0C20 0021 0820 0021 0A20 2825 2826 QA20 QF21 QE21 QB20 QE21 QB20 Q021 QB20 0021 0620 0021 0820 QD21 2825 2825 QA20 QF21 0C20 QE21 QB20 0021 QB20 QD21 0620 Q021 0520 0021 0620 2825 2825 0A20 QF21 QC20 QE21 Q820 Q021 Q520 Q021 0820 0021 0A20 0021 0820 2825 2826 QA20 QF21 QE21 QB20 QE21 QB20 Q021 QB20 0021 0820 0021 0820 0021 2826 2825 QB20 QF21 QE21 QD21 QC20 QD21 QB20 QD21 0C20 0021 0620 0021 QA20 24 2825 2826 QB20 QB20 0C20 QE21 QD21 QC20 QD21 QB20 0021 0820 0021 0820 0021 22 2826 2825 2826 QB20 0F21 QE21 QC20 QD21 QA20 0021 0620 0021 0020 0021 0620 20 2825 2825 QA20 0F21 QE21 QE21 QC20 Q021 QB20 0021 0B20 0021 0820 QD21 18 2826 2825 QA20 QF21 QC20 QD21 QC20 QD21 QA20 0021 0620 0021 QB20 16 2825 2826 2826 0A20 0F21 QD21 0D21 QC20 0021 0020 0029 0820 Q021 14 2825 1 2825 2825 QB20 2826 QF21 QC20 QE21 0020 0021 0020 0621 QB20 12 2825 2825 2826 QB20 QA20 QF21 QF21 0621 0621 0021 0820 0621 10 2826 2825 2825 2825 QA20 QF21 0F21 0C20 0821 0021 QC20 08 2825 2825 2826 2825 QA20 0620 0820 QF21 0F21 0F21 06 2825 2825 2826 2825 2826 0820 0820 QA20 0A20 04 2825 2825 2825 2825 2826 2825 02 2826 2825 2826 2825 2825 Page 11 of 17 NF-BEX- 11-3 Rev. 2 NP-Attachment

Figure 2 - Quad Cities 1 Cycle 21 - Reference Loading Pattern 31 33 35 37 39 41 43 45 47 49 51 53 55 57 59 Legends Center 60 2825 1 2825 1 2826 2825 2826 ASYTYP 58 1 2825 1 2826 1 2825 2825 2826 2825 56 0A20 0A20 0B20 Q620 2826 2825 2826 2825 2825 54 0F21 QF21 0F21 Q820 0820 0A20 2825 2826 2825 2825 52 0C20 0E21 0E21 0C20 0F21 QF21 QA20 2825 2825 2825 2826 50 0821 0520 0021 0E21 0E21 0E21 QF21 0A20 0620 2826 2825 2825 48 QB20 0E21 QC20 0021 0C20 0E21 0C20 0F21 2826 0020 2825 2825 46 0021 0B20 QD21 QC20 0021 0020 0021 0021 0F21 QA20 2826 2826 44 0520 QD21 0620 0021 QA20 QD21 QC20 0021 0C20 QF21 QA20 42 0021 0620 0021 0620 0021 0820 QD21 QC20 0E21 0E21 QF21 QA20 2825 40 0820 0021 0C20 QD21 QB20 QD21 QA20 0021 0C20 QE21 0F21 QB20 2626 2826 38 QD21 QB20 QD21 0620 0021 0620 0021 0020 0021 QE21 QC20 QB20 QB20 36 0A20 0D21 0B20 0021 QC20 QD21 0520 0021 0C20 QD21 QE21 QF21 QB20 34 0021 0820 0021 0620 0021 0820 QD21 0820 0E21 QB20 QE21 QF21 QA20 32 0820 QD21 0A20 Q021 0820 QD21 0820 0021 Q620 0621 QC20 QF21 QA20 30 Q820 0021 0A20 0021 0B20 QD21 0620 0021 0820 QE21 QC20 QF21 QA20 28 0021 0820 0021 QB20 0021 QB20 QD21 0620 0E21 QB20 QE21 QF21 QA20 2825 26 QA20 0021 0620 QD21 QG20 QD21 0820 0021 0C20 QD21 QE21 0F21 QB20 2826 24 0021 0B20 QD21 Q620 QD21 0520 QD21 0C20 0021 QE21 OC20 QB20 QB20 22 0820 0021 0C20 QD21 0820 QD21 0A20 0021 0C20 QE21 QF21 0820 2826 20 0021 0820 0021 0B20 0021 0820 QD21 0C20 QE21 0E21 QF21 QA20 2825 18 0620 0021 0820 0021 0A20 0021 QC20 0021 0020 QF21 QA20 2825 2826 0021 0B20 0021 QC20 0021 0C20 0021 0021 QF21 QA20 2826 0620 QE21 QC20 QD21 0C20 QE21 0C20 0F21 2826 QB20 2825 2825 0521 0520 0021 QE21 0621 QE21 QF21 0A20 0820 2826 2825 2825 0C20 0621 QE21 QC20 QF21 0F21 0A20 2825 2825 2825 2826 0F21 0F21 QF21 0820 Q62D 0A20 2825 2826 2825 2825 0A20 0A20 0820 0820 2826 2$25 2826 2825 2825 2825 2826 2825 2825 2825 2825 2825 2825 2826 2825 2826 Page 12 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment

Bundle Cycle Name Number Enrichment Type Loaded QA20 Opt2-3.99-15GZ8.00-3G6 00 56 3.99 20 QB20 Opt2-4.00-13GZ8.00-3G6.00 112 4.00 20 QC20 Opt2-4.05-12GZ7.00-2G6.00 68 4,05 20 QD21 Opt2-3.98-18GZ8.00 148 3.98 21 QE21 Opt2-3.99-16GZ8.00 64 3.99 21 QF21 Opt2-4.01-14GZ6.00 56 4.01 21 QG22 Opt2-4.07-19GZ7.50/5.50 104 4.07 22 QH22 Opt2-4.07-17GZ7.50/5.50 56 4.07 22 Q122 Opt2-4.12-12G5.50-2GZ5.50 60 4.12 22 Bundle Cycle Name Number Enrichment Type Loaded 2825 GE14-P1 ODNAB409-17GZ-1OOT-145-T6-2825 128 4.09 19 2826 GE14-PIODNAB408-15GZ-1GOT- 145-T6-2826 68 4,08 19 QA20 Opt2-3.99-15GZ8.00-3G6.00 56 3.99 20 QB20 Opt2-4.00-13GZ8.00-3G6.00 136 4.00 20 QC20 Opt2-4.05-12GZ7,00-2G6.00 68 4.05 20 QD21 Opt2-3.98-18GZ8.00 148 3.98 21 QE21 Opt2-3,99-16GZ8.00 64 3.99 21 QF21 Opt2-4.01-14GZ6.00 56 4.01 21 120 0003 ."PU Power - 2957 MW.,.

1101 100`% Core F10w - 98.0 Mi6/3r t 3200 A: 43.6 3 / 23.0 1 F E 3: 54.2 8 2 / 35.5 i F C: 1D0.0 6 1 / 95.3 4 F 2800 0: 100.0 100.0 100.0 9 1 / 108.0 F

2 F 1 F: 27.0 z ? / 108.0 "_. F 18.8 +. ? / 36.6 12400 04.9 : P / 88.5 8 c (7331 . Jrzq 11.00 ?ower - 2511 tOW., I 2000 1600 T 400 10 Cavitation Interlock Line 0 0 0 10 20 30 40 50 60 70 80 90 100 110 120 Core Flow(%)

Page 13 of 17 NF-BEX-11-3 Rev. 2 NP-Attachment

Figure 4 Quad Cities 1 Cycle 22 SLMCPR Results for SVEA-96 Optima2 Fuel a,c Page 14 of 17 NF-BEX-l 1-3 Rev. 2 NP-Attachment

Figure 5 - Assembly Histograms a,c Page 15 of 17 NF-BEX-1 1-3 Rev. 2 NP-Attachment

Figure 6 - Assembly Histograms a,c Page 16 of 17 NF-BBX-11-3 Rev. 2 NP-Attachment

Figure 7 - Assembly Histograms a,c Page 17 of 17 NF-BEX- l 1-3 Rev. 2 NP-Attachment

Retrieved from "https://kanterella.com/w/index.php?title=RS-11-155,_Additional_Information_Supporting_the_Request_for_Technical_Specification_Change_for_Minimum_Critical_Power_Ratio_Safety_Limit&oldid=2465899"