RS-13-288, Request for License Amendment to Technical Specifications Section 5.6.5. Core Operating Limits Report

Request for License Amendment to Technical Specifications Section 5.6.5. Core Operating Limits Report

ML13354C045
Person / Time
Site:	Dresden, Quad Cities
Issue date:	12/19/2013
From:	Gullott D Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML13354C065	List: RS-13-288, Request for License Amendment to Technical Specifications Section 5.6.5. Core Operating Limits Report
References
RS-13-288
Download: ML13354C045 (113)

Text

Proprietary Information - Withhold From Public Disclosure Under 10 CFR 2.390 RS-13-288 December 19. 2013 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 NRC Docket Nos. 50-237 and 50-249 Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 10 CFR 50.90

Subject:

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

In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC), requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS). Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed change revises Technical Specifications (TS) Section 5.6.5, "Core Operating Limits Report (COLR)," to add an NRC approved topical report reference to the list of analytical methods that are used to determine the core operating limits. Specifically, the proposed change adds a reference to Westinghouse topical report WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description. Qualification and Application."

This request is subdivided as follows.

• provides a description and evaluation of the proposed change.

Attachments 2 and 3 provide markups of the affected TS page for DNPS and QCNPS, respectively.

• provides Westinghouse Report NF-BEX-13-143-P, "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," Revision 2. contains information proprietary to Westinghouse Electric Company LLC; it is supported by an affidavit signed by Westinghouse, the owner of the information. The affidavit, Contains Proprietary Information. Withhold From Public Disclosure Under 10 CFR 2.390. When separated from Attachment 4, this document is decontrolled.

December 19, 2013 U.S. Nuclear Regulatory Commission Page 2 provided in Attachment 5, 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 respectfully requested that the information which is proprietary to Westinghouse be withheld from public disclosure in accordance with 10 CFR 2.390. A non-proprietary version of Attachment 4 is also provided in Attachment 5.

The proposed change has been reviewed by the DNPS and QCNPS Plant Operations Review Committees and approved by the Nuclear Safety Review Board in accordance with the requirements of the EGC Quality Assurance Program.

EGC requests approval of the proposed change by December 19, 2014. Once approved, the amendment will be implemented within 60 days. This implementation period will provide adequate time for the affected station documents to be revised using the appropriate change control mechanisms.

In accordance with 10 CFR 50.91, "Notice for public comment; State consultation,"

paragraph (b), EGC is notifying the State of Illinois of this application for license amendment by transmitting a copy of this letter and its attachments to the designated State Official.

There are no regulatory commitments contained in this letter. Should you have any questions concerning this letter, please contact Mr. Kenneth M. Nicely at (630) 657-2803.

I declare under penalty of perjury that the foregoing is true and correct. Executed on the 19th day of December 2013.

David M. Gullott Manager - Licensing Attachments:

1. Evaluation of Proposed Change

2. Markup of Proposed Technical Specifications Page for DNPS

3. Markup of Proposed Technical Specifications Page for QCNPS

4. Westinghouse Report NF-BEX-13-143-P, "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," Revision 2 (PROPRIETARY INFORMATION)

5. Westinghouse Application for Withholding, Affidavit, and Non-Proprietary Version of cc:

NRC Regional Administrator, Region III NRC Senior Resident Inspector - Dresden Nuclear Power Station NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Illinois Emergency Management Agency - Division of Nuclear Safety

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION 2.0 DETAILED DESCRIPTION

4.0 REGULATORY EVALUATION

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

6.0 REFERENCES

Page 1

ATTACHMENT 1 Evaluation of Proposed Change 1.0

SUMMARY

DESCRIPTION In accordance with 10 CFR 50.90, "Application for amendment of license, construction permit, or early site permit," Exelon Generation Company, LLC (EGC), requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS), Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2. The proposed change revises Technical Specifications (TS) Section 5.6.5, "Core Operating Limits Report (COLR}," to add an NRC approved topical report reference to the list of analytical methods that are used to determine the core operating limits. Specifically, the proposed change adds a reference to Westinghouse topical report WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application" (Le., Reference 1).

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

A license amendment request to add a reference to WCAP-16865-P-A to TS 5.6.5 for DNPS and QCNPS was previously submitted to the NRC in Reference 2; however, that license amendment request was withdrawn in Reference 3 as discussed below. In Reference 4, the NRC provided the acceptance review results of the Reference 2 license amendment request. In summary, the NRC determined that the following additional information was needed to complete the detailed technical review and render an assessment of the proposed action:

1. A revision to the proposed TS change to include revision numbers and dates for the new topical report reference being added to TS 5.6.5; and

2. The Emergency Core Cooling Systems (ECCS) analysis completed using the new methodology.

Relative to Item 2, the subject ECCS analysis was not scheduled to be completed in a timeframe that would have supported a prompt response to the NRC's request. Therefore, EGC withdrew the license amendment request in Reference 3. In discussions with the NRC prior to the withdrawal of the license amendment request, the NRC clarified that review of a representative ECCS analysis using the new methodology was necessary for the NRC to validate that the WCAP-16865-P-A methodology was implemented correctly.

Following withdrawal of the Reference 2 license amendment request, efforts to complete a representative ECCS analysis using the new methodology continued. A representative ECCS analysis was recently completed for QCNPS (Le., Attachment 4), and as a result, the proposed change to add a reference to WCAP-16865-P-A to TS 5.6.5 for DNPS and QCNPS is being re-submitted to the NRC for review. In addition, the proposed TS change has been modified to address Item 1 above.

Page 2

ATTACHMENT 1 Evaluation of Proposed Change 2.0 DETAILED DESCRIPTION The proposed change adds the following Westinghouse topical report reference to TS 5.6.5:

WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates:

Supplement 4 to Code Description, Qualification and Application," Revision 1, October 2011.

This topical report will be used for its method of determining the end of lower plenum flashing for analysis of a loss-of-coolant accident (LOCA). A markup of the proposed change is provided in Attachments 2 and 3 for DNPS and QCNPS, respectively.

3.0 TECHNICAL EVALUATION

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

Application of a new NRC approved Westinghouse methodology to support operation of the SVEA-96 Optima2 fuel contained in the reactor core, including core operating limits for future core configurations, requires a revision to TS 5.6.5.b. The proposed change will also allow DNPS and QCNPS to use the most current Westinghouse methodology for determination of the APLHGR limits associated with TS 3.2.1, "Average Planar Linear Heat Generation Rate (APLHGR}."

The NRC review of topical report WCAP-16865 is documented in Reference 5. This topical report is to be used for its proposed method of determining the end of lower plenum flashing for analysis of a BWR LOCA. The change credits steam cooling for a longer time than the previous evaluation model resulting in a corresponding reduction in the peak clad temperature (PCT).

The NRC concluded that the topical report was acceptable for referencing in licensing applications to the BWR LOCA ECCS Evaluation Model to the extent specified, and under the limitations and conditions described in the final safety evaluation.

There are two limitations and conditions specified in Reference 5, Section 5.0, "Limitations and Conditions." Note that the two specific limitations and conditions contain proprietary information and are not repeated in this section.

EGC confirms that the method to determine the end of lower plenum flashing proposed for use in the DNPS and QCNPS LOCA analyses includes the two specific requirements described in Reference 5, Section 5.0, "Limitations and Conditions." These two requirements are implemented by the use of the appropriate Westinghouse codes and procedures.

Page 3

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

As discussed above, a license amendment request to add a reference to WCAP-16865-P-A to TS 5.6.5 for DNPS and QCNPS was previously submitted to the NRC in Reference 2.

However, that license amendment request was withdrawn due to the inability to provide a completed ECCS analysis using the new methodology in a timeframe that would have supported a prompt response to an NRC request. In discussions with the NRC prior to the withdrawal of the license amendment request, the NRC clarified that review of a representative ECCS analysis using the new methodology was necessary for the NRC to validate that the WCAP-16865-P-A methodology was implemented correctly.

A representative ECCS analysis using the new methodology has been completed. The ECCS analysis completed using the WCAP-16865-P-A methodology for Quad Cities Units 1 and 2 is provided in Attachment 4.

4.0 REGULATORY EVALUATION

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

paragraph (c)(5).

The proposed change to TS Section 5.6.5 does not alter the design or function of any plant structure, system, or component; does not result in any change in the qualifications of any component; and does not result in the reclassification of any component's status in the areas of shared, safety-related, independent, redundant, or physical or electrical separation. The proposed addition of the Westinghouse topical report to TS 5.6.5 is necessary in order to use the most recent approved Westinghouse methodology to determine core operating limits for future core configurations. Therefore, compliance with 10 CFR 50.36 will be maintained.

Page 4

ATTACHMENT 1 Evaluation of Proposed Change 4.2 No Significant Hazards Consideration In accordance with 10 CFR 50.90. "Application for amendment of license. construction permit, or early site permit," Exelon Generation Company. LLC (EGC), requests an amendment to Renewed Facility Operating License Nos. DPR-19 and DPR-25 for Dresden Nuclear Power Station (DNPS). Units 2 and 3, and Renewed Facility Operating License Nos. DPR-29 and DPR-30 for Quad Cities Nuclear Power Station (QCNPS),

Units 1 and 2. The proposed change revises Technical Specifications (TS) Section 5.6.5, "Core Operating Limits Report (COLR)." to add an NRC approved topical report reference to the list of analytical methods that are used to determine the core operating limits. Specifically, the proposed change adds a reference to Westinghouse topical reportWCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates:

Supplement 4 to Code Description, Qualification and Application," Revision 1, October 2011. The proposed addition of the Westinghouse topical report to TS 5.6.5 is necessary in order to use the most recent approved Westinghouse methodology to determine core operating limits for future core configurations. This topical report will be used for its method of determining the end of lower plenum flashing for analysis of a BWR loss-of-coolant accident (LOCA).

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

(1)

Involve a significant increase in the probability or consequences of any accident previously evaluated; or (2)

Create the possibility of a new or different kind of accident from any accident previously evaluated; or (3)

Involve a significant reduction in a margin of safety.

EGC has evaluated the proposed change, using the criteria in 10 CFR 50.92, and has determined that the proposed change does not involve a Significant hazards consideration. The following information is provided to support a finding of no significant hazards consideration.

1.

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

Response: No TS Section 5.6.5 lists NRC approved analytical methods used at DNPS and QCNPS to determine core operating limits. The proposed change adds an NRC approved topical report reference to the list of analytical methods in TS Section 5.6.5.

The proposed change to TS Section 5.6.5 will add a Westinghouse methodology to determine the end of lower plenum flashing for analysis of a BWR LOCA. The PageS

ATTACHMENT 1 Evaluation of Proposed Change proposed change will allow DNPS and QCNPS to use the most current Westinghouse methodology for determination of the APLHGR limits associated with TS 3.2.1, "Average Planar Linear Heat Generation Rate (APLHGR)."

The addition of an approved analytical method in TS Section 5.6.5 has no effect on any accident initiator or precursor previously evaluated and does not change the manner in which the core is operated. The NRC approved method ensures that the analysis output accurately models the predicted core behavior, has no effect on the type or amount of radiation released, and has no effect on predicted offsite doses in the event of an accident. Additionally, the NRC approved method does not change any key core parameters that influence any accident consequences. Thus, the proposed change does not affect the probability of an accident previously evaluated.

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

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

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

2.

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

Response: No The proposed change to TS Section 5.6.5 does not affect the performance of any DNPS and QCNPS structure, system or component credited for mitigating any accident previously evaluated. The NRC approved analytical methodology for evaluating the APLHGR limits will not affect the control parameters governing unit operation or the response of the plant equipment to transient conditions.

The proposed change does not introduce any new accident precursors, modes of system operation, or failure mechanisms.

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

3.

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

Response: No The proposed change will add a reference to the list of analytical methods in TS Section 5.6.5 that can be used to determine core operating limits. The new Page 6

ATIACHMENT1 Evaluation of Proposed Change methodology has been previously approved by the NRC and accurately establishes the appropriate APLHGR limits. The proposed change does not modify the safety limits or setpoints at which protective actions are initiated and does not change the requirements governing operation or availability of safety equipment assumed to operate to preserve the margin of safety. Therefore, the proposed change does not impact the level of protection currently provided.

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

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

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

5.0 ENVIRONMENTAL CONSIDERATION

EGC has determined that the proposed amendment would change a requirement with respect to installation or use of a facility component located within the restricted area, as defined in 10 CFR 20, "Standards for Protection Against Radiation." However, the proposed amendment does not involve: (i) a Significant hazards consideration, (ii) a significant change in the types or significant increase in the amounts of any effluent that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, the proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22, "Criterion for categorical exclusion; identification of licensing and regulatory actions eligible for categorical exclusion or otherwise not requiring environmental review,"

paragraph (c)(9). Therefore, pursuant to 10 CFR 51.22, paragraph (b), no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

6.0 REFERENCES

1.

Westinghouse Electric Company Topical Report WCAP-16865-P-A, "Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application," Revision 1, dated October 2011 Page 7

ATTACHMENT 1 Evaluation of Proposed Change

2.

Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, "License Amendment Request to Technical Specifications Section 5.6.5, 'Core Operating Limits Report (COLR),'" dated April 3,2012

3.

Letter from D. M. Gullott (Exelon Generation Company, LLC) to U.S. NRC, "Withdrawal of License Amendment Request to Technical Specifications Section 5.6.5, 'Core Operating Limits Report (COLR),'" dated May 29,2012

4.

Email from B. Mozafari (U.S. NRC) to J. A. Bauer (Exelon Generation Company, LLC),

"Need to supplement application dated April 3,2012 (RS-12-036)," dated May 8,2012 (ADAMS Accession No. ML12130A339)

5.

Letter from J. F. Quichocho (U.S. NRC) to J. A. Gresham (Westinghouse Electric Company), "Final Safety Evaluation for Westinghouse Electric Company Topical Report WCAP-16865-PIWCAP-16865-NP, Revision 1, 'Westinghouse BWR ECCS Evaluation Model Updates: Supplement 4 to Code Description, Qualification and Application' (T AC No. ME2901)," dated September 29,2011 Page 8

ATTACHMENT 2 Markup of Proposed Technical Specifications Page for DNPS Dresden Nuclear Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-19 and DPR-25 REVISED TECHNICAL SPECIFICATIONS PAGE 5.64

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 5.6.6 CORE OPERATING lIMITS REPORT (COlR)

(continued)

9.

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

10.

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

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

c.

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

d.

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

Post Accident Monitoring (PAM) Instrumentation Report When a report is required by Condition B or F of lCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

11.

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

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

ATTACHMENT 3 Markup of Proposed Technical Specifications Page for QCNPS Quad Cities Nuclear Power Station, Units 1 and 2 Renewed Facility Operating License Nos. DPR-29 and DPR-30 REVISED TECHNICAL SPECIFICATIONS PAGE 5.6-4

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 5.6.6 CORE OPERATING LIMITS REPORT (COLR)

(continued)

8.

WCAP-15836-P-A, "Fuel Rod Design Methods for Boiling Water Reactors Supplement 1."

9.

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

10.

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

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

c.

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

d.

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

Post Accjdent Monitorjng (PAM) Instrumentatjon Report When a report is required by Condition B or F of LCO 3.3.3.1, "Post Accident Monitoring (PAM) Instrumentation," a report shall be submitted within the following 14 days.

The report shall outline the preplanned alternate method of monitoring, the cause of the inoperability, and the plans and schedule for restoring the instrumentation channels of the Function to OPERABLE status.

11.

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

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

ATIACHMENT5 Westinghouse Application for Withholding, Affidavit, and Non-Proprietary Version of Attachment 4

WestInghouse U.S. Nuclear Regulatory Commission Document Control Desk 11555 Rockville Pike Rockville, MD 20852 Westinghouse Electric Company Engineering, Equipment and Major Projects 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA Direct tel: (412) 374-4643 Direct fax: (724) 720-0754 e-mail: greshaja@westinghouse.com Proj leiter:

CAW-I 3-385 I November II, 2013 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

NF-BEX-13-143 P-Attachment, Revision 2, "Quad Cities I & 2 LOCA Analysis for SVEA-96 Optima2 Fuel" (Proprietary)

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CA W-13-38S1 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 ofthe Commission's regulations.

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

Correspondence with respect to the proprietary aspects of the application for withholding or the Westinghouse Affidavit should reference CA W-13-3851, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, Suite 310, 1000 Westinghouse Drive, Cranberry Township, Pennsylvania 16066.

Enclosures James A. Gresham, Manager Regulatory Compliance

CAW-I3-38S1 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

ss COUNTY OF BUTLER:

Before me, the undersigned authority, personally appeared James 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 off act set forth in this Affidavit are true and correct to the best of his knowledge, information, and belief:

Sworn to and subscribed before me COMMONWEAlTH OF PENNSYLVANIA Notarial Seal Anne M. Stegman, Notary Public UnIty Twp., Westmoreland County My CommIssion ExpIres Aug. 7, 2016 M~Wlr<! PENNSYLVANIA ASSOOAnON OF NOTARIES d~

1J)ames A. Gresham, Manager Regulatory Compliance

2 CAW~13~3851 (1)

I am Manager, Regulatory Compliance, in Engineering, Equipment and Major Projects, 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 infonnation 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 infonnation reveals the distinguishing aspects of a process (or component, structure, tool, method, etc.) whel'e prevention of its use by any of Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies,

3 CA W-13-3851 (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 infonnation, 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.

(iii)

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 ~ompetitive position.

(b)

It is information that is marketable in many ways. The extent to which such infonnation 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-13-3851 (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, anyone 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.

(iv)

The information is being transmitted to the Commission ill confidence and, under the provisions of 10 CFR Section 2.390, it is to be received in confidence by the Commission.

(v)

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.

(vi)

The proprietary information sought to be withheld in this submittal is that which is appropriately marked in NF-BEX-13-143 P-Attachment, Revision 2, "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel" (Proprietary), for submittal to the Commission, being transmitted by Exelon Generation Company letter and Application for Withholding Proprietary Infonnation from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse is that associated with the Quad Cities LOCA break spectrum analysis based on the implementation of WCAP-16865-P-A, Revision I, and may be used only for that purpose.

5 CAW-I 3-385 I (a)

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

(i)

Assist Exelon Generation Company in obtaining approval of the Quad Cities Units I & 2 Licensing Amendment Request for implementing the revised LOCA methodology of WCAP-1686S-P-A, Revision I.

(ii)

Provide licensing support for customer submittals of a related nature.

(b)

Further this information has substantial commercial value as follows:

(i)

Westinghouse plans to sell the use of the information to its customers for the purpose of improvement in MAPLHGR operating margins.

(ii)

Westinghouse can sell support and defense of analyses involving methods used for similar BWR LOCA reload applications.

(iii)

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 BWR LOCA methods 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 10caJ 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.

Westinghouse Non-Proprietary Class 3 to NF-BEX-13-1S0, Rev. 2 NF-BEX-13-143-NP, Revision 2 "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel" (90 pages not including this page)

NF-BEX-13-143-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 NF-BEX-13-143-NP Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel Revision 2 Page 1 of90 Revision 2 I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 2 of90 "This page intentionally blank".

NF-BEX-13-143-NP Revision 2 I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Westinghouse Review Documentation Author:

Verifier:

Approved:

November 2013 Patrick R. Kottas

• LOCA Analysis and Methods John A. Blaisdell

• Ajal J. Parikh

• LOCA Analysis and Methods D. Atkins for John Ghergurovich

• Manager, LOCA Analysis and Methods

• Electronically approved records are authenticated in the electronic document management system.

NF-BEX-13-143-NP Westinghouse Electric Company LLC 1000 Westinghouse Drive Cranberry, P A 16066

© 2013 Westinghouse Electric Company LLC All Rights Reserved Page 3 of90 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 4 0[90 Rev. #

Date 0

October 2013 1

October 2013 2

November 2013 NF-BEX-13-143-NP REVISION RECORD Revision Description Original issue

• Add a Record of Revisions page

• Section 3: Modify Reference 3-10 to "Not Used."

• Section 6: Remove Section 6 (UFSAR markups)

• Section B.2.1: Remove use of Reference 3 -1 0

• Section BA: Update to state that the ERV is the lowest pressure setpoint relief valve

• Section BA2: Update from "30 seconds" to "23 seconds"

• Section BA.3: Update to state that for breaks less than 0.15 ft2, the break is in the pump discharge leg

• Section B.5: Update to state that Table B-8 is the sequence of events for the limiting small break

• Update page numbering for the TOC, List of Tables/Figures

• Section 1: Delete reference to Section 6

• Section 2: Delete Reference 2-10

• Section B.1.l: Delete reference to Reference 2-10

• Section 2: Delete Reference 2-11

• Section 1: Delete reference to Reference 2-11

• Record of Revisions: Update "RV" to "ERV"

• Section B.l.1: Change "Automatic Depressurization System (ADS)" to "Adjustable Speed Drive (ASD)"

• Page 3: Update Manager signature to "D. Atkins for John Ghergurovich"

• Section BA.1: Delete"... of approximately 17 seconds... "

• Section B.1: Change"...lower flashing... " to "...lower plenum flashing... "

• Section AA.1: Change"... 1. 79... " to "... 1. 729... "

• Change dates from "October 2013" to "November 2013"

• Update "Revision 1" to "Revision 2"

• Section A.1: Change "...Section II... " to... Section 8.2... "

Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 5 of90 Table of Contents Section 1 Scope................................................................................................................................. 12 Section 2 Exelon Design Inputs and Associated References............................................................ 14 Section 3 List of Westinghouse Calculations and Documents.......................................................... 16 Section 4 General References........................................................................................................... 17 Section 5 Results............................................................................................................................... 18 APPENDIX A LOCA Input Report.......................................................................................................... 19 APPENDIX B LOCA Analysis Report..................................................................................................... 48 NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 60[90 List of Tables Table A-I Reactor Geometry - Volumes.................................................................................................... 23 TableA-2 Reactor Geometry - Heat Conductor Volumes.......................................................................... 23 TableA-3 Reactor Geometry - Reference Elevations (above vessel zero)................................................ 24 Table A -4 Reactor Geometry - Reference Dimensions.............................................................................. 25 Table A-5 Water Level Instrumentation Parameters.................................................................................. 35 Table A-6 Technical Specification Scram Speed....................................................................................... 45 TableA-7 Break Location, Single Failure and ECCSAvailability for Dresden and Quad Cities.............. 46 Table B-1 Break Location, Single Failure and ECCS Availability............................................................ 57 Table B-2 Break Spectrum Confirmation Results (Recirculation Line Breaks)........................................ 61 Table B-3 Case 1 (LPCI IV Failure): PCT Results for Recirculation Line Breaks................................... 61 Table B-4 Case 2 (EDG Failure): PCT Results for Recirculation Line Breaks........................................ 68 Table B-5 Case 3 (HPCI Failure): PCT Results for Recirculation Line Breaks....................................... 74 Table B-6 Case 4 (LSL Failure): PCT Results for Recirculation Line Breaks.......................................... 79 Table B-7 Sequence of Events for Limiting Large Break: 1.0DEGPS...................................................... 85 Table B-8 Sequence of Events for Limiting Small Break: O.lOFT2PD..................................................... 86 NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 7 of90 List of Figures Figure A-I Power - Flow Map for QC Analysis........................................................................................ 21 Figure A-2 QC FW Temperature Versus Fraction of Rated Power............................................................ 22 FigureA-3 Recirculation Pump Homologous Curves............................................................................... 36 Figure A-4 ECCS Schematic (short-term injection mode)........................................................................ 37 Figure A-5 HPCI Actuation Logic............................................................................................................. 37 Figure A-6 Performance of 2 LPCI Pumps................................................................................................ 38 Figure A-7 Performance of 4 LPCI Pumps................................................................................................ 39 Figure A-8 LPCI Actuation Logic.............................................................................................................. 40 Figure A-9 Performance of 1 LPCS Pump................................................................................................. 41 Figure A-I 0 LPCS Actuation Logic........................................................................................................... 42 FigureA-ll ADS Actuation Logic............................................................................................................. 42 Figure A-12 MSIV Closure Logic............................................................................................................. 43 Figure A -13 MSIV Closure Profile............................................................................................................ 43 FigureA-14 Simplified EDG Loading Diagram........................................................................................ 44 Figure A -15 Location of Leakages............................................................................................................. 44 Figure B-1 Flow of Information Between Computer Codes for a Two-Channel GOBLIN Model........... 55 Figure B-2 GOBLIN Two Channel Model Noding Diagram.................................................................... 58 Figure B-3 Summary of Revised Break Spectrum Confirmation............................................................... 60 Figure B-4 Case 1: Dome Pressure for LPCI IV Failure 1.0DEGPS Break.............................................. 62 Figure B-5 Case 1: Break Flow for LPCI IV Failure 1.0DEGPS Break................................................... 63 Figure B-6 Case 1: LPCS Flow Rate for LPCIIV Failure 1.0DEGPS Break.......................................... 63 Figure B-7 Case 1: System Mass for LPCI IV Failure 1.0DEGPS Break................................................ 64 NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 8of90 Figure B-8 Case 1: System Pressure for LPCI IV Failure 2.5 Fr Pump Suction Break........................... 64 Figure B-9 Case 1: Break Flow for LPCI IV Failure 2.5 Fr Pump Suction Break................................... 65 Figure B-10 Case I: LPCS Flow Rate for LPCI IV Failure 2.5 Fr Pump Suction Break........................ 65 Figure B-Il Case 1: HPCI Flow Rate for LPCI IV Failure 2.5 Fr Pump Suction Break......................... 66 Figure B-12 Case 1: Relief Valve, Safety Valve/ADS Flow Rate for LPCI IV Failure 2.5 Ft2 Pump Suction Break.................................................................................................................... 66 Figure B-13 Case 1: System Mass for LPCI IV Failure 2.5 Fr Pump Suction Break.............................. 67 Figure B-14 Case 1: Suppression Pool Temperature for LPCI IV Failure 2.5 Fr Pump Suction Break.. 67 Figure B-15 Case 2: Dome Pressure for EDG Failure 1.0DEGPS Break................................................. 69 Figure B-16 Case 2: Break Flow for EDG Failure 1.0DEGPS Break...................................................... 69 Figure B-17 Case 2: LPCS Flow Rate forEDG Failure 1.0DEGPS Break.............................................. 70 Figure B-18 Case 2: LPCI Flow Rate for EDG Failure 1.0DEGPS Break............................................... 70 Figure B-19 Case 2: System Mass for EDG Failure 1.ODEGPS Break.................................................... 71 Figure B-20 Case 2: Dome Pressure for EDG Failure 0.8DEGPS Break................................................. 71 Figure B-21 Case 2: Break Flow for EDG Failure 0.8DEGPS Break...................................................... 72 Figure B-22 Case 2: LPCS Flow Rate for EDG Failure 0.8DEGPS Break.............................................. 72 Figure B-23 Case 2: LPCI Flow Rate for EDG Failure 0.8DEGPS Break............................................... 73 Figure B-24 Case 2: System Mass for EDG Failure 0.8DEGPS Break.................................................... 73 Figure B-25 Case 3: Dome Pressure for HPCI Failure 0.10 Fr Pump Discharge Break.......................... 75 Figure B-26 Case 3: Break Flow for HPCI Failure 0.10 Fr Pump Discharge Break............................... 75 Figure B-27 Case 3: Relief Valve, Safety Valve/ADS Flow for HPCI Failure 0.10 Fr Pump Discharge Break................................................................................................................................. 76 Figure B-28 Case 3: LPCS Flow Rate for HPCI Failure 0.10 Fe Pump Discharge Break....................... 76 Figure B-29 Case 3: LPCI Flow Rate for HPCI Failure 0.10 Ft2 Pump Discharge Break........................ 77 NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 9 of90 Figure B-30 Case 3: System Mass for HPCI Failure 0.10 Ft2 Pump Discharge Break............................. 77 Figure B-31 Case 3: Suppression Pool Temperature for HPCI Failure 0.10 Ft2 Pump Discharge Break. 78 Figure B-32 Case 4: Dome Pressure for LSL Failure 1.0 Ft2 Pump Discharge Break.............................. 80 Figure B-33 Case 4: Break Flow for LSL Failure 1.0 Ft2 Pump Discharge Break................................... 80 Figure B-34 Case 4: Relief Valve, Safety Valve/ADS Flow for LSL Failure 1.0 Ft2 Pump Discharge Break................................................................................................................................. 81 Figure B-35 Case 4: HPCI Flow for LSL Failure 1.0 Fr Pump Discharge Break.................................... 81 Figure B-36 Case 4: Suppression Pool Temperature for LSL Failure 1.0 Fr Pump Discharge Break.... 82 Figure B-37 Case 4: LPCS Flow for LSL Failure 1.0 Ft2 Pump Discharge Break................................... 82 Figure B-38 Case 4: LPCI Flow for LSL Failure 1.0 Fr Pump Discharge Break.................................... 83 Figure B-39 Case 4: System Mass for LSL Failure 1.0 Ft2 Pump Discharge Break................................. 83 Figure B-40 Heat Transfer Coefficient at Peak Plane for Limiting Large Break....................................... 87 Figure B-41 Peak Cladding Temperature for Limiting Large Break......................................................... 88 Figure B-42 MAPLHGR Comparison for QC 1 C23: Previous vs. Revised LOCA Methodology............. 90 NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 LIST OF ACRONYMS Item Definition ADS Automatic Depressurization System AOR Analysis of Record ASB Acoustic Side Branch ASD Adjustable Speed Drives BWR Boiling Water Reactor CCST Contaminated Condensate Storage Tank CFR Code of Federal Regulations DBA Design Basis Accident DEG Double-Ended Guillotine DEGPS Double-Ended Guillotine Pump Suction DLO Dual Loop Operation DWP Drywell Pressure ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EM Evaluation Model EOOS Equipment Out Of Service ERV ctromatic Relief Valve FW Feedwater FWTR F eedwater Temperature Reduction HPCI High Pressure Coolant Injection HTC Heat Transfer Coefficient ICF Increased Core Flow IV Injection Valve L2 Reactor Vessel low-low water level L3 Reactor Vessel low water level LCO Limiting Condition for Operation LOCA Loss of Coolant Accident i

LOOP Loss of Off site Power LPCI Low Pressure Coolant Injection LPCS Low Pressure Core Spray LSL Loop Select Logic MAPLHGR Maximum Average Planar Linear Heat Generation Rate MCPR Minimum Critical Power Ratio NF-BEX-13-143-NP Page 10 0[90 Revision 2 November 2013 I

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I

WESTINGHOUSE NON-PROPRIETARY CLASS 3 MELLLA Maximum Extended Load Line Limit Analysis MOV Motor Operated Valve MR Medium Range MSIV Main Steam Isolation Valve NPSH Net Positive Suction Head PCT Peak Cladding Temperature PD Pump Discharge PS Pump Suction QC Quad Cities QCI Quad Cities Unit 1 QC2 Quad Cities Unit 2 RDV Recirculation Discharge Valve RV Reactor Vessel RVP Reactor Vessel Pressure SLO Single Loop Operation SP Suppression Pool SRV Safety Relief Valve I

SSV Steam Safety Valve TS Technical Specification TSD Task Scoping Document TSV Turbine Stop Valve UFSAR Updated Final Safety Analysis Report WLOP Westinghouse LOCA Operating Parameters NF-BEX-13-143-NP Page 11 of90 Revision 2 November 2013 I

i I

I i

I i

i

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Section 1 Scope Page 12 of 90 This report documents the current Emergency Core Cooling System (ECCS) perfonnance evaluation for a postulated Loss of Coolant Accident (LOCA) for Quad Cities (QC) Units I and 2. The scope of the original LOCAAnalysis was described in Task Scoping Document (TSD) DQW04-021, Reference 3-11, and the analysis was documented in the previous Analysis of Record (AOR), Reference 3-12. The latter provides the basis for this updated AOR. The updated analysis was perfonned in Reference 3-14.

This report (I) describes and implements the modification approved per WCAP-16865-P-A of Reference 3-13 which updates the LOCA methodology for defining the end oflower plenum flashing, (2) refers to a separate report, Reference 2-9, for most of the operating parameters required as input to the LOCA analysis, (3) is specific to QC, and (4) removes references to the Plant 5 designation (other than in Section A.3 to show the genesis of the bounding Plant 5 values) which previously referred to a set oflimiting operating conditions among all of the QC and Dresden Units.

The updated methodology of Reference 3-13 redefines the end oflower plenum flashing where credit is taken for the convective cooling due to flashing in the lower plenum, as long as the plenum continues to flash or the core spray flow reaches rated conditions. [

]a,c End oflower plenum flashing can be defined as the time after initiation oflower plenum flashing when either:

By this revised definition, lower plenum inventory may still be flashing when the core spray pumps reach rated flow. In order to comply with Appendix K requirements, CHACHA uses the Appendix K core spray Heat Transfer Coefficients (HTCs) when the rated core spray flow is reached, regardless of the state of lower plenum flashing.

A limitation and condition of the approval of Reference 3-13 is that [

With this modification, it is necessary either to confirm the conclusion of the previous break spectrum, or if it does not remain applicable, to determine the limiting break under the revised conditions. Based on the results of the break spectrum analysis of Reference 3-12 that analyzed a subset of the original full NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 13 of90 break spectrum, the following Double-Ended Guillotine Pump Suction (DEGPS), Pump Suction (PS) and Pump Discharge (PD) breaks are re-analyzed:

Low Pressure Coolant Injection ( LPCI) Injection Valve (IV) Failure - Recirculation Line Break:

l.ODEGPS (7.23 ft2), O.8DEGPS (5.78 W), O.6DEGPS (4.34 ft2), 2.5FT2PS, 1.0FT2PS Emergency Diesel Generator (EDG) Failure Recirculation Line Break: l.ODEGPS (7.23 ft2),

O.8DEGPS (5.78 W)

Loop Select Logic (LSL) Failure - Recirculation Line Break: 1.5FT2PD, I.OFT2PD High Pressure Coolant Injection (HPCI) Failure Recirculation Line Break: O.l5FT2PD, O.lOFT2PD This subset of the full break spectrum covers the limiting cases observed in large break, small break, and intermediate break sizes. As long as the previous trends are confirmed, there is no need to expand the list to other breaks and single failure scenarios. Based on the results of the above cases, three additional cases were added to the break spectrum analysis: LSL Failure for 2.0FT2PD, and O.5FT2PD and HPCI Failure for O.05FT2PD.

The break spectrum and peak cladding temperature (PCT) results are provided in Appendix BA. The GOBLIN system response code and the CHACHA fuel rod heat-up code are described in Appendix B.2.

The MAPLHGRs and associated PCT, maximum local oxidation, and core wide oxidation are reported in the cycle-specific reload licensing report.

Sections 2, 3, and 4 are updated to provide the references used in this updated LOCA report and to introduce Reference 2-9 which now provides the majority of the operating parameter inputs required for the QC LOCA analysis.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 140f90 Section 2 Exelon Design Inputs and Associated References 2-1 Not Used.

2-2 TODI NF0400184, Rev. 0 (NF-BEX-04-123), "Data Transfer, N.4.u - LOCA Analysis Parameters - Dresden and Quad Cities, August 17, 2004, which transmitted information from the following sources:

a.

Siemens Report, EMF-93-176, Rev. 8.

b.

Siemens Report EMF-95-165, Rev. 2.

c.

TOm NFM9800175, SEQ 02.

2-3 TODI NF0400246, Rev. 1 (NF-BEX-05-79), "Information on Dresden and Quad Cities Location of ECCS Leakages," June 20, 2005, which transmitted information obtained from the following sources:

a.

GE-NE-000-0021-3568-01, Rev. O.

b.

TOm NF0400095, Rev. O.

c.

TOm NF0400096, Rev. I.

d.

PDLB OPL-4, Rev. 3.

e.

NFM 0000009, SEQ 05.

f.

NFM 0000003 SEQ 04.

g.

GE Letter MDP-9518.

h.

GE Letter MDP-9521.

i.

GE Memo GE-NE-A0003981-1-14, Rev. 1.

j.

RDE 59-0792, Rev. 2.

k.

GE Drawing 104R921.

1.

GE Letter GENE 0000-00 1 0-0362-LTl, Rev. O.

m.

GE Letter GENE-0000-0021-4342-04, Rev. 2.

n.

NF:MW:02-0461.

2-4 TOm NF0400053, Rev. 0 (NF-BEX-04-108), "Data Transfer, N.4.fRecirculation and Jet Pumps," July 29, 2004.

2-5 TOm QDC-12-050, Rev. 0 (NF-BEX-12-111), "OPL-W Parameters for Quad Cities Unit 1 Cycle 23 Transient Analysis," August 20,2012.

2-6 TODI NF0500062, Rev. 0 (NF-BEX-05-27), "Reactor Water Level Instrument Information Quad Cities," March 17,2005.

2-7 GE Drawing 104R921, Sheet 1 (Rev. 7) & Sheet 2 (Rev. 7), "Reactor Assembly - Quad Cities 1 & 2," (TOm NF0400050 Rev. 0 and TODI NF0400099 Rev. 0, NF-BEX-04-60 and NF-BEX-04-30).

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRlETARY CLASS 3 Page 15 of90 2-8 TODI NF0600330, Rev. 0 (NF-BEX-06-286, Rev.1) "Information Related to Quad Cities ERV, SRV and MSSVs and ASB Modifications," Quad Cities 1 & 2, 14-Dec-2006, which transmitted information from the following sources:

a.

Acoustic Side Branch (ASB) Pressure Drop, S&L Calc. 11334-118-4, R1, 3/3/06,

b.

Electrical Drawings 4E-1461 Sheet 2 Typical of all ERV's on Both Units (Redundant 125 VDC Safety Related Power Supply).

2-9 TODI ES1200014, Rev. 1 (NF-BEX-12-161, Rev. 1) "Inputs for Westinghouse LOCA Analysis (WLOP-QC), July 2013.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 16 of90 Section 3 List of Westinghouse Calculations and Documents 3-1 RPB 90-93-P-A, "Boiling Water Reactor Emergency Core Cooling System Evaluation Model: Code Description and Qualification," October 1991.

3-2 RPB 90-94-P-A, "Boiling Water Reactor Emergency Core Cooling System Evaluation Model: Code Sensitivity," October 1991.

3-3 CENPD-283-P-A, "Boiling Water Reactor Emergency Core Cooling System Evaluation Model: Code Sensitivity for SVEA-96 Fuel," July 1996.

3-4 CENPD-293-P-A, "BWR ECCS Evaluation Model: Supplement 1 to Code Description and Qualification," July 1996.

3-5 WCAP-15682-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 2 to Code Description, Qualification and Application," April 2003.

3-6 WCAP-16078-P-A, "Westinghouse BWR ECCS Evaluation Model: Supplement 3 to Code Description, Qualification and Application to SVEA-96 Optima2 Fuel," November 2004.

3-7 WCAP-15836-P, "Fuel Rod Design Methods for Boiling Water Reactors Supplement 1,"

June 2002.

3-8 WCAP-15942-P, "Fuel Assembly Mechanical Design Methodology for Boiling Water Reactors Supplement 1 to CENPD-287," October 2004.

3-9 CENPD-300-P-A, "Reference Safety Report for Boiling Water Reactor Reload Fuel," July 1996.

3-10 Not Used.

3-11 NF-BEX-06-44-P, Task Report for TSD DQW04-2l, "Final LOCA Analysis for Quad Cities 1 and 2 and Dresden 2 & 3," Apri12006.

3-12 OPTIMA2-TR021QC-LOCA, Revision 5, "Quad Cities 1 & 2 LOCA Analysis for SVEA-96 Optima2 Fuel," September 2009.

3-13 WCAP-16865-P-A, Revision 1, "Westinghouse BWR ECCS Evaluation Model Updates:

Supplement 4 to Code Description, Qualification and Application," October 2011.

3-14 CN-LAM-13-6, Revision 0, "Quad Cities 1 and 2 Break Spectrum Analysis for WCAP-16865 Implementation Project," September 2013.

3-15 NF-BEX-05-2, Revision 2, "LOCA Input Parameters for Quad Cities 1 & 2 and Dresden 2 &

3," February 2006.

NF-BEX-13-143-NP Revision 2 November 2013

None used.

NF-BEX-13-143-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 Section 4 General References Page 170[90 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Section 5 Results Page 18 of90 This report documents the LOCAAnalysis for QC using SVEA-96 Optima2 fuel using the updated LOCA methodology of Reference 3-13. The MAPLHGRs and associated PCT, maximum local oxidation, and core wide oxidation are reported in the cycle-specific reload licensing report.

The LOCA input parameters description is in Appendix A.

The LOCA analysis, including the break spectrum and limiting breaks, is discussed in Appendix B.

Section B.4 provides the results of this revised break spectrum analysis.

Appendix C of the previous break spectrum AOR, Reference 3-12, contains the debris blockage evaluation. That Appendix is not repeated here.

Appendix D of the previous break spectrum AOR, Reference 3-12, contains the correspondence with the NRC regarding the use of containment backpressure in the analysis. That Appendix is not repeated here.

The break spectrum runs presented in this section use QC specific values as discussed in Appendix A and in Reference 2-9.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 APPENDIX A LOCA Input Report This attachment contains the LOCA input parameter discussion.

NF-BEX-13-143-NP Page 19 of 90 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 20 0[90 A.1 INTRODUCTION Reference 2-9 contains system-related input parameters that are used for the ECCS performance analyses of QC I & 2. The system-related parameters of Reference 2-9 are reviewed prior to performing the reload analysis for each cycle. Any required revisions to these input parameters are confirmed by Exelon. The LOCA analysis is performed using the guidance provided in Section 8.2 of Reference 3-9.

Sets of parameters are provided for a single bounding model that represents a conservative composite of the QC units. This model is used for the break spectrum and single failure confirmation analysis.

The bounding QC model is used for the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) analyses.

Basic plant and ECCS parameters are discussed in the following sections but since the LOCA input parameters are defined in Reference 2-9, they will not be repeated here. Note that Revision I of Reference 2-9 had several updates from that in Revision O. These changes were for the:

LPCS minimum injection flow to close bypass flow isolation valve increase to 874 gpm Relief valve re-opening maximum delay time increase to 18.0 sec Safety Relief Valve (SRV) open stroke time decrease to 0.1 secs Corrected sum for QC2 total LPCI leakage The results of this analysis have taken into account all of the input changes in Revision L It also accounts for responses to comments received on previous drafts of this report.

A.2 INITIAL CONDITIONS The following basic plant operating parameters are provided in,Reference 2-9, Section 1, Basic Parameters:

Core power for DLO, Single Loop Operation (SLO) at rated and analyzed conditions Core flow rate for DLO, SLO at rated and analyzed conditions Feedwater (FW) temperature at nominal and analyzed band conditions FW temperature reduction range Steam flow rate Recirculation flow Steam dome pressure for DLO, SLO Initial water level DWP for large and small breaks Figure A-I shows a power-flow map for the QC Units. Since the plants may be operated for DLO at 102% power (100% + 2% rated thermal power measurement uncertainty) over a range of core flow rates, both ends of the range (points A and B in Figure A-I) have been evaluated to determine which is limiting. Section B.5.2 of Reference 3 -12 presents a comparison of an analysis initialized at low NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 21 of90 flow Maximum Extended Load Line Limit Analysis (MELLLA) point, 95.3% of rated core flow, to the limiting case initialized at 108% flow. The conclusion of this study is that the analysis at the Increased Core Flow (ICF) point (108%) produces results that are more limiting than at the low flow MELLLA point (95.3%). The initial conditions considered for two-loop operation are taken from Reference 2-9, Section 1.

FW temperature and steam flow vary with reactor power. Since FW temperature is an input to the LOCA code, the input at the analyzed power level must be determined from a correlation. Figure A-2 shows the variation of FW temperature with the fraction of rated thermal power. Values are taken from Section 1 of Reference 2-9.

In Single Loop Operation (SLO), the jet pumps in the 'idle' loop can have either forward or reverse flow through them depending on the speed of the operating recirculation pump. SLO is detrimental to the LOCA, particularly if the break is in the operating loop. In this case, the beneficial effect of the operating recirculation pump coastdown on core flow is absent. The upper power limit during SLO is along the MELLLA boundary up to the core flow rate associated with point C in Figure A-I (55.1 %

of rated core flow). The power level assumed in the SLO LOCA analysis is 102% ofthe maximum power level permitted for SLO (70.2% + 2% rated thermal power), or 72.2% of rated thermal power.

The initial conditions considered for SLO LOCA analysis are taken from Section I of Reference 2-9.

110 100 90 ao

'6" 11/ '; 70 e:.

60 11/

~

Q.

ii 50 E

11/

40

.s:.

I-30 20 10 0

0 100",(, Power = 2957 MWth 100% Flow" 98 Mlb/hr A: 102% P /95,3% F B: 102% P /108% F C; 72.2% P /55.1% F 10 20 30 Quad Cities Power - Flow Map 40 50 60 70 Core Flow (%-rated)

,A_

Operating Boundary 80 90 Figure A-I Power - Flow Map for QC Analysis NF-BEX-13-143-NP 100 B _.

110 120 Revision 2 November 2013

390.0 370.0 s: 350.0 111

I ftI 330.0 111 Q.

E 111 I-....

310.0

~

-c 111 290.0

<1/

LL.

270.0 250.0 WESTINGHOUSE NON-PROPRIETARY CLASS 3 FW Temperature vs.. Normalized Power 0.3 Tfw =-98*QjQratedIl2+242. 78*QjQrated+21O.82 0.4 0.5 0.6 0.7 Q/Qrated 0.8 0.9 1

1.1 Figure A-2 QC FW Temperature Versus Fraction of Rated Power A.3 REACTOR SYSTEM GEOMETRY Page 22 of 90 Summary reactor system geometry data is provided in this section in Tables A-I throughA-4 and is taken from Reference 3-15. Westinghouse calculation notes provide the details of how the information obtained from plant drawings was converted into the detailed information required by the GOBLIN code.

Values used in the final model are based on the limiting of QC or Dresden data and has historically been referred to as Plant 5 data. Dresden data is provided to show the genesis of the bounding Plant 5 values.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 23 of90 A.3.t Volumes Table A-t Reactor Geometry - Volumes Description Units Dresden*

Quad Cities* I Plant 5 Notes Lower plenum ttl 2220.0 2220.0 2220.0 Control rod guide tubes ft3 1227.8 1227.8 1227.8 Active core ttl 930.1 930.1 930.1 Water cross ttl 112.4 112.4 112.4 Core bypass ttl lO23.4 lO23.4 lO23.4 Jet pumps ttl 279.5 279.5 279.5 Downcomer ttl 5347.9 5347.9 5347.9 Upper plenum ttl 2112.7 2112.7 2112.7 Steam dome ttl 5619.5 5569.1 5569.1 Recirculation loops ttl 1099.4,1036.3 lO99.4 lO36.3 1

Total ft3 19972.7, 19922.3 19859.2 19909.6

• If a single value is provided, it is identical for both units; otherwise two values are provided (unit n, unit n+ 1).

1.

In most instances, a lower system volume is conservative as it results in earlier dryout. One exception to this rule is the lower plenum volume, which must be filled by ECC water.

A.3.2 Heat Conductors Table A-2 Reactor Geometry - Heat Conductor Volumes Description Units Dresden*

Quad Cities*

Vessel bottom head ttl 407.1 407.1 Vessel cylinder ttl 1795.6 1795.6 Vessel top head ft3 391.8

[

391.8 Control rod guide tubes I

ft3 158.0 158.0 Control rod drive housing ttl 32.7 32.7 Control rod drive shafts ttl 89.5 89.5 Control blades ttl 104.5 lO4.5 Lower core plate assembly ttl 41.6 41.6 Bottom tie plate ttl 44.6 44.6 Fuel channels ttl 77.2 77.2 Water cross ttl 37.4 37.4 NF-BEX-13-143-NP Plant 5 Notes 407.1 1795.6 391.8 158.0 32.7 89.5 104.5 41.6 44.6 77.2 37.4 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 24 of90 TableA-2 Reactor Geometry - Heat Conductor Volumes Description Units Dresden*

Quad Cities*

Plant 5 Notes Top tie plate ft!

4.5 4.5 4.5 Upper core grid fe 30.4 30.4 30.4 Fuel assembly handles fe 8.2 8.2 8.2 Core shroud I head I separator ft!

482.2 482.2 482.2 Steam dryer assembly ftl 170.0 220.44 220.44 1

Jet pumps ft!

20.6 20.6 20.6 Recirculation line piping ft!

227.3 227.3 227.3 Total ft!

4123.2 4173.6 4173.6

• If a single value is provided, it is identical for both units; otherwise two values are provided (unit n, unit n+ I).

1.

Dryer heat conductor volume is based on a dryer weight of 85,000 Ibs for the Dresden units and 110,220 1bs for the QC units (includes +10% uncertainty).

A.3.3 Reference Elevations Table A-3 Reactor Geometry - Reference Elevations (above vessel zero)

Description Units Dresden*

Quad Cities*

Plant 5 Notes Bottom of active fuel (BAF) for inch 216.312 216.312 216.312 SVEA-960ptima2 Top of active fuel (T AF) for S VEA-inch 361.588 361.588 361.588 960ptima2 Bottom jet pump diffuser inch 136.97 136.97 136.97 Top jet pump suction inch 312.3 312.3 312.3 Feedwater sparger inch 486.0 486.0 486.0 LPCS sparger (lower) inch 378.0 378.0 378.0 LPCS sparger (upper) inch 388.75 388.75 388.75 Recirculation pump suction inch

-284.0

-284.0

-284.0 Recirculation pump outlet inch

-260.0

-260.0

-260.0

• If a single value is provided, it is identical for both units; otherwise two values are provided (unit n, unit n+ I).

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 25 of90 A.3.4 Reference Dimensions Table A*4 Reactor Geometry* Reference Dimensions Description Units Dresden*

Quad Cities*

Plant 5 Notes Reactor vessel ID inch 251.0 251.0 251.0 Shroud ID at BAF inch 203.125 203.125 203.125 Shroud 00 at BAF inch 207.125 207.125 207.125 Vessel height inch 823.625 823.625 823.625 Jet pwnp dimensions Nozzle drive ID inch 3.31 3.31 3.31 ThroatID inch 8.170 8.170 8.170 Throat length inch 89.0 89.0 89.0 Diffuser & tail length inch 99.58 99.58 99.58 Diffuser exit ID inch 20.12 20.12 20.12 Core inlet orifice IDs Side entry 1 nwnber inch 1#

1.424/68 1.425/84 1.424/68 1

Bottom entry 1 nwnber inch /#

0.77/16 NA 0.77 /l6 1

Central/number inch 1 #

2.262/640 2.262/640 2.262/640 Recirculation piping Suction (ID) inch 25.774,25.594 25.625 25.774 2

Discharge (10) inch 25.466,25.548 25.250 25.548 2

Ring manifold (ID) inch 20.042, 19.750 20.000 19.750 3

Risers-external (ID) inch 11.376 11.750 11.376 3

Risers-internal (ID) inch 10.020 10.020 10.020

• If a single value is provided, it is identical for both units; otherwise two values are provided (unit n, unit n+ I).

L The QC reference asswned that the bottom entry orifices would be sized so that the loss across them would be the same as the peripheral side entry orifices. Thus, the 84 includes the 16 bottom entry orifices. The Dresden data were selected for Plant 5 because the two types of peripheral orifices were kept separate.

2.

Largest ID is used at potential break location to increase potential break flow rate.

3.

Smallest 10 is used at locations that are not limiting potential break locations to minimize system volwne as it leads to the earliest predicted core dryout.

NF-BEX-13-143-NP Revision 2 November 2013 i

I i

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 26 of90 A.4 SYSTEM PERFORMANCE DATA A.4.1 Recirculation System The recirculation system, which is an extension of the Reactor Vessel (RV) pressure boundary, provides a variable volumetric flow rate of forced circulation water flow through the reactor core.

The system takes water from the lower downcomer region of the RV and returns it to the RV under-core region via the jet pumps. The volumetric flow rate of the water circulated through the core is varied by adjusting the speed of the recirculation pumps.

Two separate recirculation loops are provided. Each loop has a variable speed pump, which discharges to a 28-inch discharge line that is equipped with a motor-operated isolation valve. The discharge line of each recirculation pump supplies water to a 'ring header'. The ring header is comprised of separate sections for each recirculation loop. There is no fluid communication between the two sections of the ring header. Five supply risers tap off each section of the ring header. Each riser penetrates the RV and provides driving flow to two jet pumps.

The units are able to operate with only one of the recirculation pumps powered provided the core power and flow are restricted as described in Section A.2.

Of importance to the LOCA analysis, the recirculation system provides:

Core flow during blowdown due to the inertia of the rotating components, and A closed return path to the jet pump nozzles for the LPCI system.

Section I of Reference 2-9 provides inputs for the recirculation system. The pump single-phase homologous curves are presented in Figure A-3, which was copied from Reference 2-2.b.

The jet pumps, which are located between the core shroud and the RV wall, are supplied driving flow from the recirculation pumps. Each jet pump consists of a nozzle section, a throat section, a diffuser and a tail section, which is attached at its lower end to the shroud support. The driving fluid emerging from the jet pump nozzle has a high velocity. Suction fluid is entrained from the downcomer by a process of momentum exchange. The combined flow enters the mixing section, or throat, where the velocity decreases and the resultant pressure increases.

Jet pump perfonnance is measured by the flow ratio (M) and the efficiency (11), which is the product of the flow ratio and the head ratio (N). These tenns are defined as follows:

NF-BEX-13-143-NP M=

WI N= Pd -P2 PI-Pd 1]=MxN p=!!..L+l.v 2 P

2 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 27 0[90 where WI and W2 are the driving and driven mass flows; PI. P2 and Pd are the specific energies of the driving flow, suction flow and discharge flow; ps is the static pressure, p is the fluid density; and V is the fluid velocity.

Jet pump performance for the Dresden units is provided in Attachment 5 of Reference 2-4. Although no information was provided for the QC units, Tahle 5.4-1 of the QC UFSAR indicates identical information for the QC units. This information is as follows:

M 1.729

'11= 0.308 The actual recirculation flows for each unit are different corresponding to M-ratios for the QC units.

Higher M-ratios than the 1.729 are used in the break spectrum analysis. Sensitivity studies show no sensitivity to changes in M-ratio on the timing of boiling transition. There is no effect of M-ratio after the downcomer water level drops below the top of the jet pumps and boiling transition actually occurs well after this time. Therefore, the M-ratio has a negligible impact on the result. Also, as shown in Reference 2-9, there can be a mismatch between the flow rates of the individual recirculation pumps. This mismatch can have an impact on the time of core dryout, especially for breaks in the recirculation line. For these breaks, the pump having the highest initial speed will be located in the recirculation loop containing the break as the pump coastdown in the broken loop will have no positive effect on core flow.

A.4.2 High Pressure Coolant Injection (HPCI) System The HPCI system consists of a steam turbine driving a multi-stage high-pressure pump and a gear-driven single stage booster pump, valves, high-pressure piping, water sources and instrumentation. A schematic of the entire ECCS is shown in Figure A-4. The HPCI system takes suction from the Contaminated Condensate Storage Tank (CCST) or the Suppression Pool (SP). When the water level in the CCST decreases below a predetermined level, the pump suction is transferred automatically to the SP. Water from either source is pumped into the RV through the FW sparger. Although the HPCI system can take suction on either the CCST or the SP, only the SP can be credited for a design basis accident since it is seismically designed.

The system performance and actuation parameters are described in Reference 2-9, Section 2. The logic for actuation of the HPCI system for the QC units is shown in Figure A-5.

The QC HPCI performance is simple in that it will deliver rated flow of 5000 gpm, provided the system pressure is within the operating band of HPCI, 55 seconds (Section 2 of Reference 2-9) after the actuation signal regardless of the position of the minimum flow valve.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 28 of90 A.4.3 Low Pressure Coolant Injection (LPCI)

As shown in Figure A-4, there are two LPCI divisions. Each division consists of a heat exchanger (not shown in the figure), two LPCI pumps in parallel and associated piping and valves. During LPCI operation, the pumps take suction from the SP through normally open Motor-Operated Valves (MOVs) and discharge to the RV through the discharge leg of the selected recirculation loop. The pumps discharge through normally open manual pump discharge isolation valves to the LPCI heat exchangers. Ifdesired, the LPCI water can be cooled by this heat exchanger. AMOV allows LPCI flow to bypass the heat exchanger. The bypass valves are normally open until the operator takes action. The heat exchangers are used for containment and SP cooling modes. During initial LPCI injection, the LPCI flow bypasses the heat exchangers.

The two LPCI subsystems are cross-connected downstream of each heat exchanger by piping containing two normally open MOVs. The cross-connection allows all of the LPCI pumps to discharge to one recirculation loop. After the pumps and crosstie header, LPCI flow enters a header where it can be used for multiple purposes. One flow path available is injection to the RV via the recirculation discharge piping. Water is injected through two MOVs. The outboard MOV is normally open and the inboard MOV is closed. When injection is required, the inboard MOV opens automatically after the RV pressure decreases to the permissive pressure setpoint. The outboard MOV can be used to throttle flow.

LPCI parameters for use in the LOCA analysis are provided in Section 4 of Reference 2-9.

Loop Selection Logic (LSL) - If the LOCA were in one of the recirculation loops (e.g., pump discharge piping) the open LPCI crosstie valves would result in almost all of the injected LPCI coolant being diverted to the break. The loop-select logic is designed to prevent this loss of water.

The loop selection logic is initiated by either high DWP or low-low water level as -follows:

A.

If one of the recirculation pumps is not running, the pump mode selector section of the logic trips the running recirculation pumps and waits for the reactor pressure to reach the pressure permissive (P2 in Section 4 of Reference 2-9). This permits the previously running recirculation pump to coastdown.

B.

A time delay imposes a wait for momentum effects to establish the maximum differential pressure for loop selection.

C.

Four differential pressure detectors compare the pressure between riser pipes in loop A and the corresponding riser pipes in loop B.

D.

If the loop A pressure is 2.2 psi greater than the loop B pressure (Reference 2-9), the logic selects loop A for injection. Otherwise, a timer runs out causing loop B to be selected.

E.

The logic seals in the loop selection and sends a close signal to the recirculation pump discharge valve CRDV) for the selected loop and to the LPCI IV in the opposite loop.

F.

Upon receipt of the reactor low-pressure permissive signal, the selected IV opens.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 29 of90 Since breaks smaller than 0.15 it? cannot be detected reliably, the LPCI loop select logic is assumed not to function for breaks equal to or smaller than this threshold. For breaks equal to or smaller than 0.15 ft2, LPCI is assumed to inject into the broken loop.

LPCI Minimum Flow - Each LPCI subsystem is equipped with a minimum flow line that routes water from the pump discharge to the torus. The minimum flow line prevents the pump from becoming deadheaded should the IV be closed or the reactor vessel pressure be too high. A MOV in the minimum flow line opens and closes automatically and is controlled by a flow transmitter located downstream of the heat exchanger.

Instrumentation is provided to select an undamaged path for the injection of LPCI flow. Power for the LPCI pumps normally comes from an auxiliary AC power bus, but if this source is not available, power is available from the emergency diesel generators supplying these buses.

Should a Design Basis Accident (DBA) LOCA occur on the 'A' recirculation loop, the LPCI loop-select logic would divert the flow from the 'A' pumps to the 'B' recirculation loop through the normally open crosstie valve. In this case, the minimum flow bypass valve will NOT receive a close signal because the 'A' flow element would be bypassed. To account for this, the LOCA analysis assumes the minimum flow valves do not close.

The LPCI flows for two and four pumps as functions of differential pressure between the reactor vessel and the wetwell are as listed in Section 4 of Reference 2-9. Performance curves for 2 LPCI pumps are shown in Figure A-6. Performance curves for 4 LPCI pumps are shown in Figure A-7.

Each figure shows curve-fits that are used for the pump performance in the QC bounding analysis:

2-Pump FQc* 2LPCI = -0.000428. Ll.P3 + 0.029. Ll.P2 - 15.4. Ll.P + 9300 4-Pump FQc* 4LPCI = -0.0005844. ~

+ 0.0096. Ll.P2 24.958* Ll.P + 15700 where F is the flow rate in gpm and ~P is the pressure difference in psid The various setpoints, permissives and delays are also presented in Section 4 of Reference 2-9. Note that the delays for reaching the RV pressure permissive and the low-low RV water level are determined from the analysis. The flow characteristics of the RDV and LPCI IV as a function of stem positions are provided in Section 9 of Reference 2-9. The logic for the actuation of the LPCI system is shown in Figure A-8.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 30 of90 A.4.4 Low Pressure Core Spray (LPCS)

As shown in Figure A-4, there are two independent LPCS systems, each with a pump, valves, piping and an independent spray sparger just above the core. Suction water is supplied by the SP through a common ECCS ring header, which has four suction lines with stainless steel strainers located in the suppression chamber. The pumps receive an automatic start signal if a low-low reactor level signal exists concurrently with a low reactor pressure signal OR a high DWP condition exists. The pumps will also start automatically if the low-low reactor level exists for a sustained time interval. Each pump is protected by a minimum flow recirculation line that prevents deadheading the pump prior to the opening of the IV. The minimum flow valve closes when the injection flow reaches a prescribed flow rate. The power source for each LPCS subsystem is located on a separate emergency bus.

Power to each bus can be supplied from the EDG if off site power is not available.

The core spray flow, for a single pump as a function of differential pressure between the RV and the wetwell, is as shown in Section 3 of Reference 2-9. Leakage is removed from core spray prior to the penetration of the core shroud as well as inside the shroud prior to reaching the spray sparger. As a result, the determination of a conservative core spray delivery curve considers these leakages. Figure A-9 provides the delivery curve, with and without leakage, for the QC units. The curve-fits for the QC units prior to subtracting leakage are shown below:

FQc,LPCS= -0.00012249' LlP3 + 0.0312' LlP2 -14.596' LlP + 5650 where F is the flow rate in gpm and iJP is the pressure drop in psid LPCS setpoints, permissives and delays are listed in Section 3 of Reference 2-9. The actuation logic of the LPCS system is shown in Figure A-I O. Delivery through the LPCS IVs may be assumed to begin as soon as the valves begin to open. The flow through the IV as a function of valve stem position is listed in Section 9 of Reference 2-9.

Note that the long-term cooling analysis credits the operation of the LPCS flow for cooling of the upper portion of the core. For this analysis, it is necessary to demonstrate that there is fully developed spray flow, which requires at least 4500 gpm delivered out of the spray nozzles.

A.4.5 Safety Valves, Relief Valves, and the Automatic Depressurization System (ADS)

The safety and relief valves may open to control RV pressure for some small breaks prior to the actuation of the ADS. Each unit also has eight safety valves, in addition to the dual function SRV that provides the passive safety function.

The LOCA analysis models the actuation of both the relief and safety valves for controlling system pressure during small break LOCA. The relief valve data is presented in Section 5 of Reference 2-9.

Note that the two relief valves with the lowest setpoint perform the 'low-low set' function, which is to reduce the cycling of relief valves for over-pressure events. The' open delay time' for these valves changes to a larger delay time after the first opening of the valves. Each unit also has eight safety valves and one dual function SRV that provide this safety function. The steam safety valve (SSV) data are presented in Section 5 of Reference 2-9. The Electromatic Relief Valve (ERV), SSV, and NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 310[90 SRV capacities, are conservatively reduced 2% from the reference data to account for the additional pressure drop introduced by the Acoustic Side Branch (ASB) modification (Reference 2-8).

Along with the HPCI system, the ADS is used to mitigate small break LOCAs. Although HPCI or ADS can mitigate small break LOCAs, due to its SP temperature limitation, the HPCI system may not run very long after ADS has actuated. The ADS is used to depressurize the RV for small breaks to enable coolant makeup using LPCS and/or LPCI pumps. ADS depressurization is accomplished by opening the four relief valves and the one SRV described above. There are two timers in the ADS.

The first timer is actuated after the coincident indication oflow-Iow water level and high DWP. The second timer is actuated on indication oflow-Iow water level alone and is intended to cover very small breaks within the containment or breaks outside of the containment, which would not generate a high containment pressure signal. The setpoints for high DWP and low-low water level are the same as indicated previously (Section 3 of Reference 2-9 for example). The ADS actuates after either of the timers has timed out provided there is indication that at least one of the low-pressure ECCS pumps is running. Note that, in addition to the availability of emergency power, starting the low-pressure ECCS pumps requires coincident indication oflow-Iow RV water level AND low steam dome pressure OR indication of high DWP.

The ADS actuation logic is shown in Figure A-II. The system performance and actuation parameters are described in Section 5 of Reference 2-9.

The single active failure of the ADS is due to a mechanical failure of one relief valve to open. Since the initiation logic is redundant and passive electrical failures are not required to be evaluated, mechanical failures are the only failure modes that are required to be evaluated.

A.4.6 Steam Line Isolation The main steam piping consists of four lines that carry the reactor-generated steam to the main turbine. Each steam line is equipped with two main steam isolation valves (MSIVs), one on each side of the primary containment wall, and a combination flow restrictor and flow-measuring venturi located between the reactor and the first isolation valve. The MSIV s are part of the containment isolation system, which serve to limit the release of radioactive material to the environment in the event of a design basis accident. The Main Steam Isolation Valve (MSIV) closure time is controlled between 3 and 5 seconds. MSIVs receive an automatic closure signal on any of the following:

• low-low R V water level,

• low main steam line pressure,

• high main steam tunnel temperature, or

• high main steam line flow.

Closure of the MSIVs results in a rapid decrease in the flow of steam through the steam lines, which causes the RV pressure to increase and voids to collapse. For the LOCA event, this can delay the NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 32 0[90 shutdown of the reactor due to void feedback. MSIV parameters are presented in Section 6 of Reference 2-9. The MSIV closure logic is described in Figure A-l2. The MSIV closure profile (flow area vs. time), is provided in Section 9 of Reference 2-9 and is graphically displayed in Figure A -13. The MSIV closure only affects the behavior of the steam line break since for other breaks, the steam line is isolated by the rapid closure of the turbine stop valves (TSVs) as described in the following paragraph.

Since Loss of Off site Power (LOOP) is assumed coincident with the LOCA, the TSVs will close very rapidly (-0.1 sec). Also, due to the LOOP, the turbine bypass valves are assumed to remain closed.

Therefore, the steam line will isolate very rapidly whether or not the MSIVs close. For some small breaks, this will result in pressurization of the RV until the safety andior the relief valves (including the SRV) open. In these cases, the system pressure will remain high until the ADS actuates.

A.4.7 Feedwater Isolation The condensate and FW systems supply the RV with demineralized water to make up for the steam generation in the core. The condensate and feedwater systems include condensate pumps, condensate booster pumps, feedwater heaters, feed pumps, feedwater regulating valves, associated piping, controls and instrumentation.

There are four condensate-pumping units, each consisting of one condensate pump and one condensate booster pump driven by a common motor. If a LOCA were detected, one of the units would trip to limit the loading on the 4kV buses.

The FW heaters are divided into three parallel strings, with three low-pressure FW heaters (A, B, and C) and one high-pressure FW heater (D) in each string. The reactor feed pumps take suction from the C low-pressure FW heaters and discharge through the FW regulating valves to the D high-pressure FW heaters. Each of the three feed pumps is driven by an electric motor.

FW to the reactor is controlled by throttling the FW regulating valves. Two full-flow FW regulating valves are provided for power operation. One low-flow regulating valve is used for lower power operation. FW flows from the high-pressure FW heaters to the RV through two IS-inch diameter lines.

In the more likely LOCA event, offsite AC power would be available and the condensate-FW systems would provide significant makeup capacity. However, the DBA LOCA event is assumed concurrent with LOOP. Therefore, the condensate pumps, condensate-booster pumps and feed pumps are assumed to lose motive power at the onset of the event. FW will continue to be supplied for a short time due to inertia of spinning components and then coast to zero flow. Since the addition of FW during the blowdown is beneficial to the outcome of the analysis, it is conservative to isolate FW more rapidly than it would occur. The conservative assumptions for FW flow coastdown are provided in Section 6 of Reference 2-9.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 330[90 A.4.8 Emergency Electrical Power Three EDGs are provided for the QC site to provide emergency AC power to the ECCS or safe shutdown equipment in the event off site AC power supply is not available. The dedicated EDG can supply one division of safety buses. In addition, there is a swing EDG that can supply the other division of safety buses.

The EDGs are designed to start and accept full load without reliance on any other on-site system, except 125 VDC. Each EDG has an air start system consisting of equipment necessary to compress, store and deliver high-pressure air sufficient to start the diesel engine. The air start system is independent of all other plant systems except 125 VDC, which is required to operate the air start solenoid valve, energize the governor shutdown solenoid and provide the initial excitation of the generator. The swing EDG has two 125 VDC power supplies, one from each unit's 125 VDC system, controlled by an automatic transfer switch.

The EDGs auto-start on any of the following signals:

• Under-voltage on the 4-kV ECCS bus

• 4 kV ECCS bus main feed breaker open

• Under-voltage on the 4 kV bus that normally feeds the 4 kV ECCS bus

• Both the main and the reserve feed breakers for the 4 kV bus that normally supplies the 4 kV ECCS bus are open

• ECCS initiation signal The ECCS pumps' motor logic is interlocked with the EDG output breakers to ensure sequential starting of the ECCS motors to prevent overloading the EDG due to high motor starting current.

Timers are added to the 'B' and 'D' LPCI pumps, and the Core Spray pumps, so that they start sequentially if the EDG is powering the bus when an ECCS initiation signal is received. The sequence timers are part of the individual pump motor start logic and are not part of any EDG logic.

However, the timers are enabled by the EDG output breaker being closed. A simplified EDG loading diagram is shown in Figure A-14. If the bus is being powered from any source other than the EDG, all ECCS pumps will start immediately on an initiation signal.

The analysis setpoints and time delays are provided in Sections 3 and 4 of Reference 2-9.

A.4.9 Leakage Jet pump, core spray line and vessel leakage amounts, the locations of these leakages and the leakage bases are provided in Section 8 of Reference 2-9. Figure A-15, (Reference 2-3), shows the locations of the leakages. The leakage ill numbers presented in Reference 2-9 refer to the numbers shown in Figure A-15.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 34 0[90 The leakages at the jet pump slip joints, the low and high shroud leakages, and the bottom head drain are modeled as regular flow paths. [

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 35 of90 A.4.10 Reactor Vessel Water Level Instrumentation The reactor vessel water level instrumentation consists offive ranges oflevel indication instruments:

narrow range, medium range, wide range, fuel zone and shutdown. The reactor trip and ECCS actuation signals are derived from the Medium Range (MR) instruments. See Table A-5.

The reference leg of the MR instruments connects to the reactor vessel at the upper tap. A line connects the tap to a condensing pot, which is located in the drywell at an elevation slightly above the tap. The reference legs are connected to the condensing pots in the containment. The reference leg passes through the containment into the reactor building at an elevation slightly below the condensing pot.

The reference legs are continuously backftlled to prevent gases from coming out of solution during vessel depressurization.

The MR instrumentation is calibrated at normal operating conditions and is not compensated for changes in reactor vessel pressure or reactor building temperature.

Table A-5 Water Level Instrumentation Parameters Quad Cities Description Units Value Ref.

Elevation top MR instrument tap inch 567.5 2-7 Elevation bottom MR instrument tap inch 434.5 Nominal Temperature of Reference OF 104 2-6 Leg NF-BEX-13-143-NP Notes Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 36 0[90

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t To suppression pool WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 37 of90

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.F.. dwaler-----,..,.-'!'-----'!'++-.-.:J:-+!

IPCS IV

,/

\\ 1 1

'.... ~

I '\\. Minimum flow va:ve //

\\ Minimum flow ValvO.! i

"""""'" ". 'I'

'1,

,I FI-elemonl \\, I' ----l------,

.----+---i'l Flow element \\~,._J

~.4-I---~51 LPCIIV--

NF-BEX-13-143-NP ROV Recirculation Pump Recirculation Pump Figure A-4 ECCS Schematic (short-term injection mode)

I-___ --. __ n_

..... -§.Ive Opened

'--=-=='--

t l HPCllnjll<:lIon II

§'

Valve Slarls 10 f-. -------*T4cC --------+1' Valvo Open Opened !

I I\\:

FIOW Detected I

~----------+l). In In/action Una I

Valve Open Minimum Flow Valve Sterls 10 Closa Figure A-5 HPCI Actuation Logic T7 V.lvo Closed Revision 2 November 2013

10000 9000 8000 7000 E

6000

a.

~ 5000 3

0 i:&:

4000 3000 2000 1000 0

0 NF-BEX-13-143-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 38 of90 QC lPCI2-PUMP PERFORMANCE 50 100 III Ref. 2-9

--"-~"""""-"---""----"-:::::'::-CiirveFit:'Section A.4.3 150 6P(psi) 200 250 300 Figure A-6 Performauce of 2 LPCI Pumps Revision 2 November 2013

18000

~

16000 14000 12000 e-10000

a.

S 8000 3:

0 6000 4000 2000 a

-2000 NF-BEX-13-143-NP WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 39 of90 QC lPCI 4-PUMP PERFORMANCE so 100 150 200 llP (psi)

---

CurveFit - Section A.4.3 II1II Ref. 2-9 300 Figure A-7 Performance of 4 LPCI Pumps Revision 2 November 2013

NF-BEX-13-143-NP WESTINGHOUSE NON-PROPRlETARY CLASS 3 Thlologlc diagram dooo not shaw the closure of the ** Ive In the minimum flow IIna. SInce the flow element that detects Injection flow Is assumed to be In the broken ICOIll the signal that Initiates valva closure would not be generated. Therafore, the minimum flow valve I. assumed to remain opan.

Figure A-8 LPCI Actuation Logic Page 40 of90 Revision 2 November 2013

6000 5000 4000 E

Q.

~

~

3000 i:i:

2000 1000 o

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 41 of90 QC lPCS PERFORMANCE

-"'--'0 ___ "'_----.. __..... _._-_._-------_...... _.. _-_._-

.~-----...... -

e1. 2-9 FQC w/leakage r-----------------------"~--_+=_=

--FQC Curve Fit o

so 100 150 200 250 300 350 6P(psi)

Figure A-9 Performance of 1 LPCS Pump NF-BEX-13-143-NP Revision 2 November 2013

LOCAILOOP NF-BEX-13-143-NP WESTINGHOUSE NON-PROPRlETARY CLASS 3 High Drywell Pressure Low-Low RV Water Level LowRV Pressure Figure A-tO LPCS Actuation Logic Start Timer 2 ECCS Pump(s) f---~

Running Figure A-U ADS Actuation Logic ADS Valve Actuation Page 42 of90 Revision 2 November 2013

WESTINGHOUSE NON-PROPRlETARY CLASS 3 430f90 Line Pressure Detected High Main Steam

.--------c;>; Tunnel Temperature LOCNLOOP Detected Low-Low Reactor Vessel Level Detected High Main Steam Flow Rate Detected MSIVs Start to Close Figure A-12 MSIV Closure Logic MSIV Closure Profile T1--1>-

MSIVs Closed 1.2.---------------------~------------------------------------------_,

1.0 +--------------+1L o

~

I!!

u..

0.8 e 0.6

~

0.4 0.2 0.0 NF-BEX-13-143-NP 0.5 1.0 1.5 Tlme(s) 2.0 Figure A-13 MSIV Closure Profile 2.5 3.0 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 44 of 90 Diesel Start LPCS Injection Valve Powered LPCI Pump A Loaded LPCI Pump B Loaded LPCS Pump Loaded 0-------...

0-----------------------------------+

o--------------------------------~

o~----------------------------~

o----------------------------~

Swing Bus Transfer Complete 0- - - - - - - - - - - - - - --.

LPCI Injection Valve and RDV Powered o~------------------------+_;

Time NF-BEX-13-143-NP o~--------------------------------------~

Figure A-14 Simplified EDG Loading Diagram Figure A-IS Location of Leakages Revision 2 November 2013

WESTINGHOUSE NON-PROPRlETARY CLASS 3 Page 45 of90 A.S REACTOR PROTECTION SYSTEM One of the design bases of the reactor trip system, in conjunction with the containment and containment isolation system, is to prevent the release of radioactive materials in excess of the limits of 10 CFR 100 as a consequence of any of the design basis accidents.

The reactor protection system is arranged as two separately powered trip systems. Each trip system has three trip logics, two of which are used to produce automatic trip signals. The third trip logic is used for a manual trip signal.

Each of the two trip logics used for automatic trip signals receives input signals from at least one trip channel for each monitored variable. The outputs of two trip channels are combined in a one-out-of-two logic such that an input signal on either or both of the independent trip channels produces a logic channel trip. The outputs of the remaining two trip channels are combined in another one-out-of-two logic, independent of the first logic channel. The outputs of the two logic channels are combined such that they must agree to initiate a scram.

The signals that will result in reactor trip include (not a complete list):

• High neutron flux

• High reactor pressure

• HighDWP

• Low reactor water level

• Scram discharge volume high level

• Partial closure of the main steam isolation valves (MSIVs)

• Manual depression of push-buttons for both logic channels

• Turbine control valve fast closure

• Turbine stop valve closure Of importance to the LOCA event are the reactor trips on high D WP, low reactor water level, and high dome pressure reactor trips. These setpoints are provided in Section 7 of Reference 2-9. Table A-6 provides the Technical Specification (TS) control rod scram speed.

CR Insertion 0/0 Time s

NF-BEX-13-143-NP Table A-6 Technical Specification Scram Speed 0

5 20 50 90 0.200 0.490 0.900 2.000 3.500 100 Ref.

3.875 2-5 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 46 of90 A.6 BREAK LOCATION AND SINGLE FAILURE INFORMATION The importance of the various single failures depends on break location as the location of the break can disable an ECCS component. Table A-7 shows the impact of various single failure I break location combinations. Because of the HPCI turbine lubrication oil cooler's dependence on the temperature of the water it pumps, there is the potential for the HPCI system to fail at high SP temperatures. To avoid operability issues, the HPCI pump is not credited in the LOCA analysis when the SP temperature exceeds l40°F, or for operating longer than the required mission time ifHPCI room cooling is lost (e.g., the Unit EDG fails to start).

The failures of the Loop Selection Logic (LSL) and the LPCI IV are considered as potentially limiting single failures. Since the former can result in diverting LPCI flow to the broken loop, the analysis model conservatively assumes that all LPCI flow feeds the area near the break node.

Therefore, the LSL and LPCI IV failures in the break spectrum analysis are assumed to result in the failure to deliver any LPCI to the reactor vessel. However, the LSL failure would also result in large flow rates in the SP suction line. Since the core spray pumps also take suction from the same line, the high flow rates have the potential to result in insufficient Net Positive Suction Head (NPSH) for the core spray pumps as the pool temperature increases. However, coincident failure of the core spray pump is not considered since analyses have shown there is adequate NPSH until after two-phase cooling has been established. This issue is considered in the long-term cooling analysis.

Table A-7 provides a cross-reference to the Limiting Condition for Operation (LCO) action statements in Technical Specification 3.5 regarding the time required to restore ECCS equipment to operable status while in mode 1. These action times ensure high availability I reliability of this equipment.

Table A-7 Break Location, Single Failure and ECCS Availability for Dresden and Quad Cities Assumed Single Break Location Failure Systems Remaining Remarks Notes Recirculation LPCI Injection Valve 2 LPCS + HPCI + 5 ADS Line or EDG or 125 VDC 1 LPCS + 2 LPCI + HPCI + 5 ADS Steam Line HPCI or 250 VDC 2 LPCS + 4 LPCI + 5 ADS Loop Selection Logic 2 LPCS + 4 LPCI + HPCI + 5 ADS ADS 2 LPCS + 4 LPCI + HPCI + 4 ADS LPCS Line LPCI Injection Valve -11 LPCS + HPCI + 5 ADS EDG or 125 VDC 2 LPCI + HPCI + 5 ADS HPCI or 250 VDC 1 LPCS + 4 LPCI + 5 ADS Loop Selection Logic I LPCS + 4 LPCI + HPCI + 5 ADS NF-BEX-13-143-NP Failure is to align to intact loop Break in powered LPCS line.

Failure has no effect due to break Revision 2 November 2013 I

2,3 4

5 6

1 2,3 NA

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 47 of 90 Table A-7 Break Location, Single Failure and ECCS Availability for Dresden and Quad Cities Assumed Single Break Location Failure Systems Remaining Remarks Notes location.

ADS I LPCS + 4 LPCI + HPCI + 4 ADS Feedwater Line LPCI Injection Valve 2 LPCS+5 ADS (break assumed to prevent EDG or 125 VDC 1 LPCS + 2 LPCI + 5 ADS delivery of HPCI) or 250 VDC 2 LPCS + 4 LPCI + 5 ADS Loop Selection Logic 2 LPCS + 4 LPCI + 5 ADS Failure has no effect due to break location.

ADS 2 LPCS + 4 LPCI + 4 ADS

1.

Single failure of LPCI IV supports the action statement in TS 3.5.1 D.

2.

Single failure of EDG or 125 VDC support the action statement in TS 3.5.1 A, B, C; TS 3.8.1 Band TS 3.8.4 B.

3.

The HPCI turbine oil cooler and gland seal condenser are cooled by water from the suppression pool.

Since these components are rated at 140°F, continued operation above a suppression pool temperature of 140°F is not permitted. Also, suppression pool temperatures above 140°F would exceed current NPSH calculations for rated HPCI pump flows. Another problem during a LOCA is the use of the HPCI room cooler to support the HPCI system's EQ qualifications. The HPCI room cooler is only powered from the Unit EDG. Therefore, any single failures of the Unit EDG would need to also consider the consequential loss of the HPCI system after 10 minutes of operation. The HPCI is not credited in the LOCA analysis when the suppression pool temperature is predicted to exceed 140°F, or for operating longer than the required mission time if HPCI room cooling is lost (e.g., the Unit EDG fails to start).

4.

Single failure of HPCI or 250 VDC supports the action statement in TS 3.5.1 F.

5.

Failure of the Loop Selection Logic supports the action statement in TS 3.5.1 D.

6.

Single failure of ADS supports the action statement in TS 3.5.1 G.

NF-BEX-13-143-NP Revision 2 November 2013 6

1 2,3 4

NA 6

WESTINGHOUSE NON-PROPRIETARY CLASS 3 APPENDIXB LOCA Analysis Report This attachment contains the LOCA Analysis report.

Page 48 of90 This document updates the previous report of Reference 3-12 to implement the methodology improvement of Reference 3-13.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRlETARY CLASS 3 Page 49 of90 B.t INTRODUCTION AND SCOPE The purpose of this report is to document the QC LOCA analysis with the updated methodology of Reference 3-13. The latter redefines the end oflower plenum flashing where credit is taken for the convective cooling due to flashing in the lower plenum as long as the plenum continues to flash or the core spray flow reaches rated conditions. [

End oflower plenum flashing can be defined as the time after initiation oflower plenum flashing when either:

By this revised definition, lower plenum inventory may still be flashing when the core spray pumps reach rated flow. In order to comply with regulatory requirements, CHACHA uses the Appendix K core spray HTCs when the rated core spray flow is reached, regardless of the state of lower plenum flashing.

A limitation and condition of the approval of Reference 3-13 is that [

]a,c Further, the present analysis uses the revised QC input contained in Reference 2-9. The revised analysis contains the following changes from the previous AOR of Reference 3-12:

DWP was updated from 22.5 psia to 22.7 psia HPCI injection delay time was updated from 45 sec to 55 sec Time delay for HPCI turbine to reach rated speed has been updated from 48 sec to 55 sec HPCI injection temperature was updated from 160°F to 140°F Capacities of safety valves 1 to 8 were updated from 645,000 Ibmlbr to 644,543 lbmlbr Core shroud repair leakage was updated from 180 gpm to 137 gpm Time delay to load pump to bus for LPCS has been updated from 14 sec to 12 sec Relief valve re-opening maximum delay time increase to 18.0 sec SRV open stroke time decrease to 0.1 sec Corrected sum for QC2 total LPCI leakage Level 3 (13) is updated from 488 in to 500 in Reactor Vessel Pressure (RVP) time delay is updated from 1 sec to 0.55 sec DWP time delay is updated from 1 sec to 0.65 sec MSIV delay is now provided as 0.06 sec Total outside shroud leakage is updated from 325 gpm to 351.6 gpm (QCl) and 173 gpm (QC2);

the more conservative QC 1 value is used in the break spectrum analysis Total LPCI leakage is updated from 811.10 gpm to 1033.8 gpm (QCl) and 782 gpm (QC2); the more conservative QC 1 value is used in the break spectrum analysis NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 50 of 90 B.1.1 Historical Items The previous AOR of Reference 3-12 addressed the following:

Break spectrum analysis - The break spectrum analysis addresses break location, break type, break size and limiting ECCS single failure. In addition to recirculation line breaks, steam line breaks outside containment (validating Automatic Depressurization System (ADS) instrumentation setpoints), steam line breaks inside containment, core spray line breaks and FW line breaks are considered. Justification of breaks that are not analyzed is provided. Technical justification of key inputs is included. The break spectrum analysis is performed by providing boundary conditions from the GOBLIN hot assembly analysis to the heat-up code CHACHA, which determines PCT and maximum oxidation. The same lattice, nodal exposure and nodal peaking are used for all CHACHA cases to ensure that the system response to the LOCA can be compared in a meaningful way.

Sensitivity studies - Sensitivity studies were performed to investigate ECCS injection temperature and current plant flexibility options such as 1) Single Loop Operation (SLO), 2) FW Temperature Reduction (FWTR) with a maximum reduction of 120°F, and 3) currently analyzed Equipment Out Of Service (EOOS) options as applied to LOCA analysis. Multipliers to Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits are calculated as needed so that the licensing basis PCT is not exceeded for operation in each specific flexibility option.

PowerlFlow Map - The LOCA analysis is performed at 102% power and 108% core flow and 102% power and MELLLA minimum flow condition. These power/flow conditions are the most limiting for the LOCA analysis and bound other powerlflow conditions on the current QC power/flow map including the MELLLA boundary.

Core performance parameters used in the LOCA analysis are listed and justified. The list of parameters includes MAPLHGR, R-factor, exposure point analyzed, initial Minimum Critical Power Ratio (MCPR) and axial peaking factor.

ADS out of service - The analysis evaluates one of five ADS valves out of service.

Long-term cooling The LOCA analysis demonstrated applicability of the current long-term cooling requirement as specified in the QC UFSAR for SVEA-96 Optima2 fuel.

LOCA analysis report content - The report includes LOCA accident description and acceptance criteria, LOCA analysis description (blowdown, refilVreflood and heatup analyses), break spectrum analysis description, single-loop analysis description and long-term cooling analysis and conclusions.

Axial power distribution - The LOCA analysis considers axial power shapes and peaking that are bounding for all the licensing core designs, which also supports spectral shift strategy.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 51 of90 Mixed cores of legacy fuel and SVEA -96 Optima2 fuel-The LOCA analysis evaluates the effect of the mixed core on the limiting break to determine the limiting core configuration.

SVEA-96 Optima2 perforation temperature - LOCA analysis determines the minimum temperature that SVEA-96 Optima2 fuel could perforate and release its fuel pin fission gasses under long-term LOCA conditions. This is to satisfy the UFSAR changes to include this limit for SVEA-96 Optima2 fuel.

Adjustable Speed Drive (ASD) - Reference 3-12 concluded that the predicted coastdown time constant for the QC ASD modification of > 5.1 seconds is reasonable for applications to ECCS performance analysis. Reference 3-12 and the current AOR both use a modeled coastdown time constant of 4.45 seconds which therefore remains conservative to the predicted time constant with the ASD modification.

Also, in order to provide less restrictive MAPLHGR limits, the original analysis approach went through a series of improvements in the following areas:

The above items and improvements are incorporated into the current break spectrum analysis.

B.l.2 Description of LOCA Event The LOCA event is postulated as a rupture of piping connected to the reactor pressure vessel wi thin the primary containment. A spectrum of piping breaks is considered to encompass all sizes and locations of breaks up to and including the circumferential failure of the largest connected pipe. A LOCA inside containment would result in the heating and pressurization of containment, a challenge to the ECCS, and the potential release of radioactive material to the environment. By evaluating the entire spectrum of postulated breaks, the most severe challenge to the ECCS and primary containment can be determined.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRlETARY CLASS 3 Page 52 of 90 This report evaluates the fuel thennal response and the BCCS perfonnance as a result of a postulated LOCA. The function of each BCCS subsystem is to ensure adequate core cooling across the entire spectrum of line break accidents when operated with other available BCCS subsystems with consideration of the Appendix K single failure criterion and without reliance on external sources of electrical power.

The LOCA event described below is for a large break in one of the two external recirculation loops.

Other break locations have slightly varying transient characteristics similar to that outlined here.

Following the postulated pipe rupture, coolant discharges rapidly through both sides of the break, with greater flow from the vessel side. The pump side flow is restricted by the reduced flow area of the jet pump nozzles and friction losses in the recirculation loop and pump. Loss of all offsite electrical power is assumed coincident with the break, resulting in a coastdown of all recirculation pumps and a rapid closure of the turbine stop valves. The closure of the turbine stop valves will cause a momentary increase in reactor vessel pressure and a potential power increase due to void collapse. Automatic reactor scram occurs as a result of high DWP, high reactor pressure or low reactor water level. Following reactor shutdown, the steam production in the core is reduced and the reactor pressure decreases rapidly. After several seconds, the water level in the downcomer falls to the jet pump suction elevation, which allows steam to flow to the break, and the break mass flow rate decreases significantly.

Flashing in the jet pumps and subsequently in the lower plenum occurs as the pressure continues to decrease. This results in a short-tenn two-phase mixture level rise in the core and downcomer. Following

• this mixture level swell, the continued inventory decrease results in decreasing reactor mixture level and system pressure. The core two-phase mixture level will drop exposing the fuel rods to a mostly steam environment and the heat transfer mode in the core transitions from nucleate boiling to film boiling and finally to steam cooling. The transition from nucleate boiling results in a fuel rod cladding heat up. By the time this occurs, the reactor will have scrammed. However, fission product decay heat will cause both the fuel and cladding temperatures to increase. The cladding temperature increase is terminated when two-phase cooling conditions are restored in the core by the BCCS equipment.

The low pressure BCCS equipment is actuated by either a high DWP signal or the combination of a low-low reactor water level signal and a low reactor pressure signaL For most breaks inside containment, the high DWP signal occurs first. Results for a given analysis are dependent on the BCCS equipment availability and actuation timing.

Since the HPCI system is turbine driven, it may not be actuated for large breaks as the reactor pressure decreases below the pressure required by the HPCI turbine.

B.l.3 Acceptance Criteria The Code of Federal Regulations Title 10 Part 50.46 provides five specific design acceptance criteria for the plant BCCS. The acceptance criteria are:

1.

Peak cladding temperature shall not exceed 2200°F.

2.

The calculated local oxidation of the cladding shall nowhere exceed 0.17 times the local cladding thickness before oxidation.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 53 of90

3.

The calculated total amount of hydrogen generated from the chemical reaction ofthe cladding with water or steam shall not exceed 0.01 times the hypothetical amount that would be generated if all of the metal in the cladding cylinders surrounding the fuel, except the cladding surrounding the plenum volume, were to react.

4.

Calculated changes in core geometry shall be such that the core remains amenable to cooling.

5.

After any calculated successful operation of the ECCS, the calculated core temperature shall be maintained at an acceptably low value and decay heat shall be removed for the extended period of time required by the long-lived radioactivity remaining in the core.

Demonstration that the first three criteria are satisfied ensures that the fourth criterion is satisfied.

B.2 WESTINGHOUSE BOILING WATER REACTOR (BWR) LOCA METHODOLOGY The Westinghouse BWR LOCA methodology is described in References 3-1 through 3-6 and 3-13. The latter describes the new USA6 Evaluation Model EM used in support of the LOCA analysis for the QC units in this report. This methodology makes use of the GOBLIN series of computer codes to calculate the BWR transient response to both large and small break LOCAs.

B.2.1 Analysis Codes The GOBLIN series of computer codes is comprised of two major computer codes GOBLIN and CHACHA-3D.

GOBLIN performs the analysis of the LOCA blowdown and reflood thermal hydraulic transient for the entire reactor, including the interaction with various control and safety systems. The hot assembly analysis is performed as a parallel calculation by running a two-channel GOBLIN model in which one of the channels represents the hot assembly. The GOBLIN code is described in detail in Reference 3-1.

The GOBLIN code can be divided into four main sections as described below.

The hydraulic model solves the mass, energy and momentum conservation equations together with the equation of state for each control volume. This model includes empirical constitutive correlations for the calculation of pressure drops, two-phase energy flow (drift flux), two-phase level tracking, spray-fluid interaction, and critical flow rate. Thermal equilibrium between phases is assumed.

The system models contain detailed models of the various reactor components, and the safety systems that are activated after a LOCA. They include the steam separators and dryers; reactor level measurement; reactor trip and depressurization systems; recirculation and jet pumps; and ECCS.

The thermal model calculates the heat conduction and heat transfer from the fuel rods, pressure vessel, and internals to the coolant. This model solves the material heat conduction equation and calculates the heat transfer from the fuel and structures to the coolant.

The power generation models calculate the heat generation due to fission, decay heat, and metal-water reaction. Fission power is determined by a point kinetics model.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 54 of90 These models are described in detail in References 3-1 through 3-6 and 3-13. The most recent change to the EM, described in detail in Reference 3-13, updates the methodology for defming the end of lower plenum flashing. This results in an improved convective cooling boundary condition for the fuel heat-up evaluation.

CHACHA-3D performs detailed fuel rod mechanical and thermal response calculations at a specified axial level within the hot assembly. All necessary fluid boundary conditions are obtained from the hot assembly thermal hydraulic analysis described above. CHACHA-3D determines the temperature distribution of each rod at the axial elevation analyzed. These results are used to determine the PCT and cladding oxidation at the axial plane under investigation. CHACHA-3D also provides input for the calculation of total hydrogen generation.

The major components of the CHACHA-3D code are the fuel rod conduction model, the channel temperature model, the heat generation model, the metal-water reaction model, the thermal radiation model, the gas plenum temperature and pressure model, the channel rewet model, the pellet/cladding gap heat transfer model and the cladding strain and rupture model. These models are described in detail in References 3-1,3-4 and 3-6.

The most recent changes to the CHACHA-3D computer code, which are described in more detail in Reference 3-6, are the addition of a new fuel rod plenum model for part-length rods and the addition of applicable fuel performance models from the approved STAV7.2 fuel performance code (Reference 3-7).

B.2.2 Analysis Process The Westinghouse analysis process as described below is in accordance with NRC approved methodology and codes. There are no deviations from the approved methodology.

The flow of information between the various analyses is shown in Figure B-1. In the case shown in the figure, the hot assembly analysis is done in conjunction with the system analysis, and the information provided to CHACHA-3D is derived from the hot assembly results.

The system analysis determines the overall response of the system to the event analyzed. The discussion in Section B.I.2 provides an example for one such scenario. In addition to the boundary conditions that are used by the CHACHA-3D heat-up analysis, the GOBLIN system analysis determines the actuation of the ECCS components that provide cooling to the core after the system is depressurized.

The hot assembly analysis determines the thermal-hydraulic conditions in the hot channel. The hot assembly initial power level is established at a conservative operating limit MCPR. This ensures that the conditions in the analyzed hot assembly bound conditions that would be seen during operation. In addition to providing the transient thermal-hydraulic channel boundary conditions that are used by the CHACHA-3D heat-up analysis, the GOBLIN system and hot assembly analysis provide the time of end of lower plenum flashing and time of two-phase recovery at the elevation of the plane analyzed by the CHACHA-3D heat-up analysis.

The Westinghouse BWR ECCS Evaluation Methodology is approved to use code predicted heat transfer conditions before the end oflowerplenum flashing, as documented in item I.D.6 of the SER of the NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 55 of90 original topical report RPB 90-94-P-A (Reference 3-2). In previous analyses, the end of lower plenum flashing time for switching to adiabatic convective heat transfer was based on [

]R.C As described in Section B.l, the purpose of this report is to document the QC LOCA analysis with the updated methodology of Reference 3-13. The latter defines the end of lower plenum flashing mechanistically and credit is taken for the convective cooling due to flashing in the lower plenum as long as the plenum continues to flash until the core spray flow reaches rated conditions.

As shown in Figure B-1, the convective HTCs predicted by the hot assembly analysis are provided to the CHACHA-3D heat-up calculation. For the time period between the start of the event and the time of end of lower plenum flashing, the convective heat transfer coefficients applied are those predicted by the hot assembly analysis at the studied location. However, the predicted heat transfer coefficients beyond the time of rated spray flow are replaced by values that are required by Appendix K. The spray cooling and two-phase heat transfer coefficients are consistent with Appendix K requirements and described in Reference 3-6.

Figure B-1 also shows fuel performance data being applied in the form of gap coefficients for the system and hot channel analyses. A core-average gap heat transfer model is applied in the core-average channel of the system analysis and hot-assembly gap heat transfer model is applied in the hot assembly channel.

In both cases, a conservatively low gap heat transfer model is chosen.

Figure B-1 also shows that fuel performance data are provided to the CHACHA-3D heat-up analysis.

These data are comprised of initial conditions for each type of fuel rod (e.g., gap dimensions, fission gas composition, initial oxide thickness, initial crud thickness, etc) as a function of nodal exposure. These data are generated for a series of six segmented power histories for each type of fuel rod being analyzed consistent with the methodology described in Reference 3-8. Analyses are performed for each power history and the most conservative result is selected.

Core neutronics, core-average Hgap, reactor vessel geometry, ECCS

+---- performance, break type, single failure Coolant temperature, pressure, information, hot channel geometry, hot channel Hgap, hot channel power void fraction, heat transfer coefficients, part-length rod plenum surface temperature CHACHA03Dy***.

FuelH8at-upAnalysis,*

Rod geometry. local planar power, fuel r+----

performance, pin-ta-pin peaking Clad temperature & oxidation

+

Figure B-1 Flow of Information Between Computer Codes for a Two-Channel GOBLIN Model NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 56 of90 B.3 BREAK SPECTRUM ANALYSIS The importance of various single failures depends on break location as the location of the break can disable an ECCS component. Each ECCS subsystem is designed to ensure adequate core cooling across the entire spectrum ofline break accidents when operated with other available ECCS subsystems determined from the Appendix K single active failure criterion. Table B-llists the break locations considered and the ECCS equipment available under different postulated single active failures. Figure B-2 shows the GOBLIN nodalization diagram for the two-channel model that is used for the updated break spectrum study.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 57 of90 Table B-1 Break Location, Single Failure and ECCS Availability Equipment Available Break Locations to Consider Recirc Steam FW Case LPCS LPCI HPCI ADS Line(l)

Line(2)

LPCS Line Failure/Comments 1

2 0

I 5

X X

~Nfail""

2 1

2 I

5 X

X ailure 3

2 4

0 5

X X

HPCI failure 4

2 4

I 5

X X

Loop select failure 5

2 4

I 4

X X

ADS failure -

important only for small breaks 6

1 0

I 5

X LPCI IV failure 7

0 2

I 5

X EDG failure + break in powered LPCS line 8

I 4

0 5

X HPCI failure 9

1 4

1 5

X Loop select failure has no effect due to break location to 1

4 1

4 X

ADS failure -

important only for small breaks 11 2

0 0

5 X

LPCI IV failure 12 1

2 0

5 X

EDG failure 13 2

4 0

5 X

HPCI failure has no effect since break location disables HPCI 14 2

4 0

5 X

Loop select failure has no effect due to break location 15 2

4 0

4 X

ADS failure -

important only for small breaks Notes:

I.

The breaks in the recirculation line were located in the pump suction line unless the break area was less than or equal to 0.15 ttl and the LPCI system was operable. In the latter case, the breaks were located in the pump discharge line so that some of the water injected by LPCI would be lost to the break.

2.

The steam line breaks both inside and outside containment are considered since the location affects the availability of high DWP signal.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Figure B-2 GOBLIN Two Channel Model Noding Diagram NF-BEX-13-143-NP Page 58 of90 Revision 2 November 2013 a,c

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 59 of90 B.4 BREAK SPECTRUM CONFIRMATION The modeling change to implement the revised definition of the end-of lower plenum flashing makes it necessary to either confirm the conclusion of the previous break spectrum of Reference 3-12, or if it does not remain applicable, to determine the limiting break under the revised conditions. From the previous break spectrum and as described in Section I and Reference 2-9, the following breaks are selected for the confirmati on:

LPCI IV Failure - Recirculation Line Break: 1.0DEGPS (7.23 ttl), 0.8DEGPS (5.78 ft2),

0.6DEGPS (4.34 ttl), 2.5FT2PS, 1.0FT2PS EDG Failure - Recirculation Line Break: 1.0DEGPS (7.23 ttl), 0.8DEGPS (5.78 ft2)

LSL Failure Recirculation Line Break: l.5FT2PD, 1.0FT2PD HPCI Failure Recirculation Line Break: 0.15FT2PD, 0.lOFT2PD This subset of the spectrum covers the peaks observed in large break, small break, and intermediate break sizes. As described below, several other break cases were selected for analysis based on the results of the above cases: LSL Failure for 2.0FT2PD, 0.5FT2PD and HPCI Failure for 0.05FT2PD.

B.4.1 Break Spectrum Results and Conclusions The previous break spectrum analysis of Reference 3-12 considered breaks in the recirculation line, the steam line, the feedwater line and the core spray injection line. Since that analysis predicted no core uncovery for the breaks in the steam line, FW line, and the core spray line, this break spectrum confirmation analysis only considers breaks in the recirculation line. Different break sizes were evaluated to identify the limiting break size for each single failure considered. As shown in Figure B-3, break configurations between the full DEG break of the recirculation line and split breaks having an area of 0.05 ttl were evaluated. Table B-2 shows the numerical results for the cases shown in Figure B-3.

Generally, the system responses for both large and small breaks with the improved LOCA methodology, as expected, are very similar to the previous analysis of Reference 3-12.

As shown in Figure B-3, the limiting break is a large double-ended guillotine break in the pump suction line with the single failure of the LPCI IV. PCT results for these large break cases are lower versus the comparable cases of Reference 3-12. Section B.5 provides the PCT for the limiting large break with application of the Appendix K HTC along with a sequence of events table.

There are differences for the small breaks from the previous break spectrum assessment. The differences that would impact the small breaks are an increase of 10 seconds in HPCI actuation delay and an increase in the assumed delay time for opening the lowest pressure setpoint ERY. As a result, the HPCI failure cases, impacted by the latter and not the former, showed a different ranking than previously analyzed in that the HPCI failure 0.10 ttl break is now the limiting small break. The limiting LSL failure case, impacted by both delay increases, has a higher PCT than previously calculated. However, these remain NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 60 of 90 non-limiting small breaks. Based on small break results, extra cases were analyzed for the confirmation:

the O.05f1? HPCI failure and the O.5ft2 LSL failure.

The break spectrum confmnation results confirm the conclusions of the previous break spectrum. The conclusions are that large breaks are more limiting than small breaks and that the LPCI IV failure is more limiting than the EDG failure. The limiting LPCI IV failure is the 1.0DEG break size which is slightly more limiting than the O.8DEG break.

A more direct comparison is made of margin differences in MAPLHGR operating limits for QC Unit 1 Cycle 23 (QCIC23). Section B.6 compares a MAPLHGR based on the current QCIC23 MAPLHGRs with a comparable case analyzed with the improved methodology. Margin improvements of [

t,e, depending on the exposure, are seen for this example.

Sections BA.2 through B.4.5 provide more detail from the recirculation line break spectrum confirmation runs.

Figure B-3 Summary of Revised Break Spectrum Confirmation NF-BEX-13-143-NP Revision 2 November 2013 a,c

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 61 of90 Table B-2 Break Spectrum Confirmation Results (Recirculation Line Breaks)

Single Break Size Break Type Break Current Break Previous Break Failure (ff)

Location Spectrum PCT Spectrum PCT eF),

eF)

Ref. 3-12 LPCIIV 7.23 DEG Suction r-

- a,c 5.78 i

DEG Suction 4.34 DEG Suction 2.50 Split Suction 1.00 Split Suction EDG 7.23 DEG Suction 5.78 DEG Suction HPCI 0.15 Split Discharge 0.10 Split Discharge 0.05 Split Discharge LSL 2.0 Split Discharge 1.50 Split Discharge 1.00 Split Discharge 0.50 Split Discharge B.4.2 Case 1: LPCI IV Failure Table B-3 Case 1 (LPCI IV Failure): PCT Results for Recirculation Line Breaks Break Spectrum Confirmation 1.0 DEG PS 0.8DEGPS 0.6 DEG PS 2.5 ff PS 1.0 ff PS This case investigated the single failure of the LPCI IV. In this situation., no coolant injection from the LPCI system occurs. However, two LPCS pumps, one HPCI pump and five ADS valves are assumed to be operable. For large breaks, the system depressurizes very rapidly and neither HPCI nor ADS are actuated.

Several break sizes in the suction line were evaluated. Breaks in the discharge line are generally not as limiting due to the flow restriction of the jet pump nozzles on the vessel side of the break. As shown in NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 62 of90 Table B-3, the limiting break is the large double-ended guillotine break in the pump suction line. These results have the same relative ordering as the assessment of Reference 3-12.

Figures B-4 through B-7 show the graphical results for the limiting full double-ended guillotine break. As shown, the dome pressure increases at the beginning of the event due to the closure of the turbine stop valves.

However, reactor trip occurs very rapidly resulting in a decrease in reactor power and a rapid decrease in dome pressure. The pressure decreases below the minimum pressure at which HPCI can operate before the HPCI pump can start. As a result, there is no HPCI injection for this break. At approximately 23 seconds, the pressure reaches the pressure permissive for the LPCS IV. Flow from the LPCS pumps enters the spray spargers in the upper plenum where it can flow downward through the core or the bypass channels. The water that flows into the core provides cooling directly to the fuel rods. The water that flows down the bypass channels refills the lower plenum until the water level reaches the core inlet. After this time, the flow through the core switches from counter-current flow to co-current upward flow.

Table B-3 shows how the PCT decreases with decreasing break size. For the very small breaks, the core does not uncover. Figures B-8 through B-14 show the results for the intermediate 2.5 ft2 break. As shown, HPCI is actuated at around 60 seconds and ceases shortly thereafter when the system pressure decreases below the required low pressure cutoff. The suppression pool temperature does not approach the 140°F limit until about 400 seconds, well after HPCI has been deactivated. Both LPCS pumps actuate at approximately 50 seconds when the system pressure decreases below the pressure permissive setpoint (300 psig). ADS is not actuated in this case.

a,c Figure B-4 Case 1: Dome Pressure for LPCI IV Failure l.ODEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 63 of90 a,c Figure B-S Case 1: Break Flow for LPCI IV Failure 1.0D EGPS Break a,c Figure B-6 Case 1: LPCS Flow Rate for LPCI IV Failure 1.0DEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 64 of90 a,c Figure B-7 Case 1: System Mass for LPCI IV Failure 1.0DEGPS Break a,c Figure B-8 Case 1: System Pressure for LPCI IV Failure 2.5 Ft2 Pump Suction Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 65 of90 a,c Figure B*9 Case 1: Break Flow for LPCI IV Failure 2.5 Fe Pump Suction Break a,c Figure B*10 Case 1: LPCS Flow Rate for LPCI IV Failure 2.5 Fe Pump Suction Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON~PROPRIETARY CLASS 3 Page 66 of90 a,C Figure B-11 Case 1: HPCI Flow Rate for LPCI IV Failure 2.5 Ff Pump Suction Break a,c Figure B-12 Case 1: Relief Valve, Safety Valve/ADS Flow Rate for LPCI IV Failure 2.5 Ff Pump Suction Break NF~BEX-13~143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 67 0[90 a,c Figure B-13 Case 1: System Mass for LPCI IV Failure 2.5 Fr Pump Suction Break a,c Figure B-14 Case 1: Suppression Pool Temperature for LPCI IV Failure 2.5 Fr Pump Suction Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 68 of90 B.4.3 Case 2: EDG Failure I Table B-4 Case 2 (EDG Failure): PCT Results for Recirculation Line Breaks Break Spectrum Confirmation 1.0 DEG PS 0.8 DEGPS

['---------'-----J a,c This case investigates the single failure of one of the EDGs to start. For this failure, only one train of equipment will be powered. Therefore, one LPCS pump, two LPCI pumps, one HPCI pump and five ADS valves will be operable. For large breaks, the system depressurization is rapid and neither HPCI pump nor ADS actuate.

The loop select logic will select the intact loop and align the LPCI pumps to the intact loop for break sizes greater than 0.15 ri. For breaks equal to or smaller than 0.15 ft2, it is assumed that loop select logic will not reliably select the intact loop. In these situations it is assumed that the break is located in the pump discharge of the selected loop. As a result, much of the LPCI water injected into the pump discharge leg will be spilled out the break. For recirculation line break areas greater than 0.15 ft2, the break is assumed to be located in the pump suction line as it results in the most inventory loss.

As shown in Figure B-3, for the EDG failure, the largest double-ended break in the pump suction line is limiting, although not as limiting as the large double-ended break for the LPCI IV failure. Figures B-15 through B-28 show the graphical results for the two largest EDG failure, double-ended, pump suction line breaks. For the 1.0DEGPS break, LPCS and LPCI pumps begin to inject at approximately 30 and 38 seconds respectively (the LPCS pumps have a higher shutoff head than the LPCI pumps). The system pressure decreases below the low pressure threshold for the HPCI pump before it can start. As shown in Figure B-19, the system mass begins to recover after the ECCS pumps begin to inject. Midplane uncover occurs at approximately 150 seconds.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 69 of90 a,c Figure B-15 Case 2: Dome Pressure for EDG Failure 1.0DEGPS Break a,c Figure B-16 Case 2: Break Flow for EDG Failure 1.0DEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 70 0[90 a,c Figure B-17 Case 2: LPCS Flow Rate for EDG Failure 1.0DEGPS Break a,c Figure B-18 Case 2: LPCI Flow Rate for EDG Failure 1.0DEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 71 of90 a,c Figure B-19 Case 2: System Mass for EDG Failure 1.0DEGPS Break a,C Figure B-20 Case 2: Dome Pressure for EDG Failure O.8DEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 72 of90 a,c Figure B-21 Case 2: Break Flow for EDG Failure O.8DEGPS Break a,c Figure B-22 Case 2: LPCS Flow Rate for EDG Failure O.8DEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 73 of90 a,c Figure B-23 Case 2: LPCI Flow Rate for EDG Failure O.8DEGPS Break a,c Figure B-24 Case 2: System Mass for EDG Failure O.8DEGPS Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 74 of90 B.4.4 Case 3: HPCI Failure Table B-5 Case 3 (HPCI Failure): PCT Results for Recirculation Line Breaks Break Spectrum Confirmation 0.15 ffpo 0.10 ff PO 0.05 ff PO

[

J~

This case considers the failure of the HPCI system. Since HPC} provides flow over a restricted range of system pressure, it does not have time to actuate for large double-ended breaks. Therefore, this investigation evaluates intennediate to small breaks only. In this situation it is assumed that two LPCS pumps, four LPCI pumps (two trains) and five ADS valves will be operable.

For break areas less than or equal to 0.15 ttl, the loop select logic is assumed to fail. In this case the break is placed in the discharge line of the selected loop, which results in much of the injected LPCI water spilling out the break.. As shown in Figure B-3, the 0.10 ttl break is the limiting break in this series.

Figures B-25 through B-3l show the graphical results for the 0.10 ttl break. As shown in Figure B-27, the safety valves control system pressure for the first part of the transient until ADS actuates at approximately 175 sec. After ADS actuation, the system depressurizes to the pennissive, which initiates the ECCS pumps. As shown in Figure B-30, the system mass begins to recover shortly after ECCS actuation. The PCT occurs at approximately 450 seconds.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 75 of90 a,c Figure B-25 Case 3: Dome Pressure for HPCI Failure 0.10 Ff Pump Discharge Break a,c Figure B-26 Case 3: Break Flow for HPCI Failure 0.10 Ff Pump Discharge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 76 of 90 a,c Figure B-27 Case 3: Relief Valve, Safety Valve/ADS Flow for HPCI Failure 0.10 Fe Pump Discharge Break a,c Figure B-28 Case 3: LPCS Flow Rate for HPCI Failure 0.10 Ftt Pump Discharge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 77 of90 a,c Figure B~29 Case 3: LPCI Flow Rate for HPCI Failure 0.10 Fr Pump Discharge Break a,c Figure B~30 Case 3: System Mass for HPCI Failure 0.10 Fr Pump Discharge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 78 of90 a,c Figure B-31 Case 3: Suppression Pool Temperature for HPCI Failure 0.10 Ft2 Pump Discharge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 79 of90 B.4.5 Case 4: Loop Select Logic Failure TableB-6 Case 4 (LSL Failure): PCT Results for Recirculation Line Breaks Break Spectrum Confirmation 2.0 fr PD 1.5 fr PD 1.0 fr PD 0.5 fr PD

[

J~

This case considers the failure of the LSL to detect and select the intact loop in the event of a break in the recirculation line. For these cases, it was assumed that the break was in the discharge leg so that the water injected by the LPCI system would be lost out the break.

Due to the location of the break, the break flow rate is smaller than for the same sized suction leg break.

As a result, the system pressure decreases more slowly and the HPCI system has time to actuate even for the largest break size. When the system pressure decreases below the permissive setpoint of the LPCI and LPCS IVs, the two LPCS pumps and four LPCI pumps begin to inject. Even though the LPCI flow is lost out the break, the system mass begins to increase shortly afterward. The trend of PCT with break size shown in Figure B-3 as well as previous break spectrum analysis indicates that small breaks would be limiting for this single failure. Figures B-32 through B-39 show the graphical results for the limiting 1.0 fe pump discharge break. A comparison of the break flow rate to the LPCI flow rate indicates that the break flow increases after LPCI actuation and that all of the water injected by LPCI is lost out the break.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON~PROPRlETARY CLASS 3 Page 80 of90 a,c Figure B-32 Case 4: Dome Pressure for LSL Failure 1.0 Fr Pump Discbarge Break a,c Figure B-33 Case 4: Break Flow for LSL Failure 1.0 Fe Pump Discharge Break NF~BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 81 of90 a,c Figure B-34 Case 4: Relief Valve, Safety Valve/ADS Flow for LSL Failure 1.0 Ff Pump Discbarge Break a,c Figure B-35 Case 4: HPCI Flow for LSL Failure 1.0 Ff Pump Discbarge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 82 of90 a,c Figure B-36 Case 4: Suppression Pool Temperature for LSL Failure 1.0 Fr Pump Discharge Break a,c Figure B-37 Case 4: LPCS Flow for LSL Failure 1.0 Fr Pump Discharge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 83 0[90 a,c Figure B-38 Case 4: LPCI Flow for LSL Failure 1.0 Fr Pump Discharge Break a,c Figure B-39 Case 4: System Mass for LSL Failure 1.0 Fr Pump Discharge Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON~PROPRIETARY CLASS 3 Page 84 of90 B.S APPENDIX K LIMITING LARGE AND SMALL BREAK As shown in Figure B~3, the limiting break. is a full double-ended guillotine break. in the pump suction line with the single failure of the LPCI IV. A heatup calculation is perfonned on this limiting case using the applicable Appendix K HTCs. This is analyzed at the same conditions as for the break. spectrum: at an assembly peaking factor of 1.75 and a bumup exposure of 12,000 MWDIMTU. The significant difference in the Appendix K approach is the HTCs employed which are illustrated in Figure B~40. The time of the end of lower plenum flashing occurs almost 30 seconds later than that seen in the limiting large break of Reference 3-12.

The sequence of events for the limiting large break. is summarized in Table B-7.

Figure B-41 shows the PCT values vs. time. The maximum Appendix K PCT for the limiting large break for the conditions analyzed is [

]a,c The maximum Appendix K PCT for the limiting small break., 0.1 OFT2PD with HPCI failure, for the same analysis conditions, i.e., assembly peaking factor of 1.75 and a bumup exposure of 12,000 MWDIMTU, is [

]a,c The sequence of events for the limiting small break. is summarized in Table B-8.

NF-BEX-13-143~NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 85 0[90 Table B-7 Sequence of Events for LimitinK LarKe Break: 1.0DEGPS Event

~eaklloss of off site power occurs Turbine stop valve closes on loss of off site power High DWP occurs Reactor scram signal on high DWP Top of jet pumps uncover Suction line uncovers Reactor low-low water level (L2) signal generated

-;;;nning of LP flashing Boiling transition time LPCS pressure permissive reached Midplane uncovers TimeofLPCSpumpstart S pumps at full speed of rated spray flow End of LP flashing LPCS injection valves full open PCToccurs NF-BEX-13-143-NP Time (Sec) 0.0 0.1 0.2 1.2 3.2 4.9 5.5 6.3 17.5 23.1 25.1 29.0 34.0 45.7 57.8 76.2 180.0 Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 86 0[90 Table B-8 Sequence of Events for Limiting Small Break: O.10FT2PD Event Break/loss of offsite power occurs Turbine stop valve closes on lose of offsite power High dome pressure occurs Reactor scram on high dome pressure High DWP occurs LPCI pumps start LPCS pumps start LPCI pumps at full speed LPCS pumps at full speed Reactor low-low water level (L2) reached Recirculation discharge valve closed Lower plenum flashes ADS valves open Top of jet pumps uncover LPCS/LPCI pressure permissive reached LPCI injection valves full open LPCI injection occurs LPCS injection valves full open PCT occurs LPCS pumps deliver rated flow NF-BEX-13-143-NP Time (Sec) 0.0 0.1 0.7 1.25 5.2 24 29 31 34 58 74 111 178 305 389 417 421 442 452 N/A Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 87 of90 a,c Figure B-40 Heat Transfer Coefficient at Peak Plane for Limiting Large Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 88 of90 a,c Figure B-41 Peak Cladding Temperature for Limiting Large Break NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 89 of90 The maximum local oxidation results are reported in the cycle-specific reload licensing report for each unit.

A maximum hydrogen generation bounding calculation is performed which assumes a disproportionate number of assemblies in the core at a higher than average relative radial bundle peaking. This assumption yields a conservative total core hydrogen generation rate. This is then verified by comparison of power distributions on a cycle-to-cycle basis.

Based on the less limiting results seen in this break spectrum analysis, the maximum local oxidation and maximum hydrogen generation results are expected to be bounded by previous analysis.

B.6 MAPLHGR COMPARISON An analysis has been performed to compare the MAPLHGRs generated with the previous LOCA methodology versus the revised methodology for the end of lower plenum flashing. This was performed for the following conditions:

QCIC23 reload MAPLHGR analysis one specific lattice (152) 1.0 DEGPS break (limiting large break failure) for the failure of LPCI IV dual loop operation nominal FW temperature Note that this may not represent exactly the conditions under which the current QCIC23 MAPLHGR was generated. Typically the MAPLHGRs are found to be limited at different exposures for either the 1.0DEGPS case or the 0.8DEGPS case. But the purpose of this comparison is to compare the exact same conditions.

The graphical comparison is provided in Figure B-42. The improvements using the revised methodology for this set of conditions are [

]a.c from the previous analysis. From 10,000 MWD/MTU and beyond, the margin gain is [

]",c These margin gains are consistent with those seen in Reference 3-13.

NF-BEX-13-143-NP Revision 2 November 2013

WESTINGHOUSE NON-PROPRIETARY CLASS 3 Page 90 of90 a,c Figure B-42 MAPLHGR Comparison for QCIC23: Previous vs. Revised LOCAMethodology NF-BEX-13-143-NP Revision 2 November 2013