ML16188A332

License Amendment Request and Exemption Request to Support the Implementation of Next Generation Fuel

ML16188A332
Person / Time
Site:	Palo Verde
Issue date:	07/01/2016
From:	Lacal M Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
102-07277-MLL/GWA
Download: ML16188A332 (75)

Text

Enclosure Attachment 8 contains PROPRIETARY information To be withheld under 10 CFR 2.390 10 CFR 50.90 10 CFR 50.12

()aps MARIA L. LACAL Senior Vice President, Nuclear Regulatory & Oversight 102-07277-M LL/GWA Palo Verde July 1, 2016 Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072 Mail Station 7605 U.S. Nuclear Regulatory Commission Tel 623.393.6491 ATIN: Document Control Desk Washington, DC 20555-0001

Subject:

Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528,59-529, and 50-530 License Amendment Request and Exemption Request to Support the Implementation of Next Generation Fuel In accordance with the provisions of Title10 of the Code of Federal Regulations (10 CFR) 50.90, Arizona Public Service Company (APS) is submitting a License Amendment Request (LAR) to revise the Technical Specifications (TS) for Palo Verde Nuclear Generating Station (PVN~9) ,

Units 1, 2, and 3 to support the implementation of Next Generation Fuel (NGF). In additrb*h ~to""

this request and in accordance with the provisions of 10 CFR 50.12, APS is requesting an exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50, Appendix K, EGGS Evaluation Models, for the use of Optimized ZIRLO' cladding.

The Enclosure to this letter provides a description and assessment of the proposed changes including a technical evaluation, a regulatory evaluation, a no significant hazards consideration, and an environmental evaluation. The Enclosure also includes eight attachments.*

Attachment 1 provides a proposed license condition and a new regulatory commitment (as defined by NEI 99-04, Guidelines for Managing NRG Commitment Changes, Revision 0) to. be implemented. Attachment 2 provides marked-up existing TS pages. Attachment 3 provides

• revised (clean) TS pages. Attachment 4 provides marked-up TS Bases pages to show the conforming changes for information only. Attachment 5 provides an assessment of limitations and conditions contained in the Safety Evaluations for NRG-approved Topical Reports related to this license amendment request.

Attachment 6 is an affidavit signed by Westinghouse Electric Company LLC (Westinghouse) that sets forth the basis on which the proprietary information in Attachment 8 may be withheld .

from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4). Correspondence with respect to the proprietary aspects of Attachment 8 or the supporting Westinghouse affidavit should reference Westinghouse letter number CAW-16-4439 and be addressed to James A. Gresham, Manager, Regulatory Attachment 8 transmitted herewith contains PROPRIETARY information.

When separated from Attachment 8, this transmittal is decontrolled.

102-07277-MLL/GWA ATTN: Document Control Desk U. S. Nuclear Regulatory Commission LAR and Exemption Request to Support Implementation of Next Generation Fuel Page2 Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066. provides a non-proprietary version of the technical analysis supporting this LAR. provides a proprietary version of the technical analysis supporting this LAR, which contains information proprietary to Westinghouse.

Two pre-submittal meetings were held with the NRC on January 20, 2016 (Agency Document Access and Management System [ADAMS] accession number ML16028A394) and on March 24, 2016 (ADAMS accession number ML16088A060) to discuss various aspects of the NGF LAR.

A license condition identified in Attachment 1 to the Enclosure is included to address Information Notice (IN) 2012-09, Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength (ADAMS accession number ML113470490). As discussed in the NRC public meeting held with APS on March 24, 2016, this license condition would require APS to incorporate NRC-approved guidance into the current licensing basis regarding fuel assembly integrity under externally applied forces as described in IN 2012-09 within 2 cycles following Mode 4 entry with the first NGF transition core. APS will notify the NRC when this action is completed.

In accordance with the PVNGS Quality Assurance Program, the Plant Review Board and the Offsite Safety Review Committee have reviewed and approved the proposed amendment. By copy of this letter, this license amendment request is being forwarded to the Arizona Radiation Regulatory Agency in accordance with 10 CFR 50.91(b)(1).

APS requests approval of the proposed license amendment within 18 months of this LAR submittal to support planned implementation during the fall 2018 Unit 2 refueling outage (2R21).

Should you have any questions concerning the content of this letter, please contact Michael Dilorenzo, Section Leader, Nuclear Regulatory Affairs, at (623) 393-3495.

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

Executed on --~J~u'"""'ly<--=1_.=20~1~6~-

(Date)

Sincerely, MLL/GWA

102-07277-MLL/GWA ATTN: Document Control Desk U. S. Nuclear Regulatory Commission LAR and Exemption Request to Support Implementation of Next Generation Fuel Page 3

Enclosure:

Description and Assessment of Proposed License Amendment cc: K. M. Kennedy NRC Region IV Regional Administrator S. P. Lingam NRC NRR Project Manager for PVNGS M. M. Watford NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS A. V. Godwin Arizona Radiation Regulatory Agency (ARRA)

. T. Morales Arizona Radiation Regulatory Agency (ARRA)

Enclosure Description and Assessment of Proposed License Amendment

Enclosure Description and Assessment of Proposed License Amendment TABLE OF CONTENTS

1.

SUMMARY

DESCRIPTION .................................................................................................... 2

2. PROPOSED CHANGES TO PVNGS LICENSING BASIS ............................................................ 2

3. TECHNICAL SPECIFICATIONS CHANGES .............................................................................. 2 3.1. TS 4.2.1- Design Features (Optimized ZIRLO') .............................................................. 2 3.2. TS 5.6.5.b-Core Operating Limits Report (COLR) Analytical.Methods ........................... 3

4. LICENSE CONDITION TO ADDRESS INFORMATION NOTICE 2012-09 ................................. 4

5. REGULATORY ANALYSIS FOR TECHNICAL SPECIFICATIONS CHANGES ............................... 5 5.1. Applicable Regulatory Requirements .....*........................*...............*................................ 5 5.2. Precedent ......................................................................................................................... 5 5.3. No Significant Hazards Consideration Determination ..................................................... 6 5.4. Conclusions ..................................................................................................................... 11

6. ENVIRONMENTAL EVALUATION ....................................................................................... 11

7. PERMANENT EXEMPTION -10 CFR 50.46 AND 10 CFR PART 50 APPENDIX K FOR OPTIMIZED ZIRLO' .............................................................. ~ ............................................ 11

8. REFERENCES ........................ ,.............................................................................................. 15 ATTACHMENTS:

ATTACHMENT 1 - License Condition and Regulatory Commitment ATTACHMENT 2 - Technical Specifications Page Mark-ups ATTACHMENT 3- Clean Technical Specifications Pages ATTACHMENT 4- Technical Specification Bases Page Mark-ups (provided for information only)

ATTACHMENT 5- Assessment of Topical Report Limitations and Conditions ATTACHMENT 6 - Affidavit from the Westinghouse Electric Company Submitted in Accordance with 10 CFR 2.390 to Consider.Attachment 8 as a Proprietary Document ATTACHMENT 7- Technical Analysis [NON-PROPRIETARY VERSION]

ATTACHMENT 8- Technical Analysis [PROPRIETARY VERSION]

1

Enclosure Description and Assessment of Proposed License Amendment

1.

SUMMARY

DESCRIPTION This evaluation supports a request to amend Operating Licenses NPF-41, NPF-51, and NPF-74, for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, respectively.

The license amendment request (LAR) will revise the Technical Specifications (TS) for PVNGS Units 1, 2, and 3 to support the implementation of Next Generation Fuel (NGF). In addition to this request, PVNGS is requesting an exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50, Appendix K, EGGS Evaluation Models, to allow the use of Optimized ZIRLO' as a fuel rod cladding material.

A license condition as identified in Attachment 1 to this enclosure is included to address Information Notice (IN) 2012-09, Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength (ADAMS acce_ssion number ML113470490).

The proposed change will allow for the implementation of NGF including the use of Optimized ZIRLO' fuel rod cladding material. The NGF assemblies contain advanced features to enhance fuel reliability, thermal performance, and fuel* cycle economics.

This change is planned to be first implemented with the fall 2018 Unit 2 refueling outage (2R21).

2. PROPOSED CHANGES TO PVNGS LICENSING BASIS Implementation of NGF at PVNGS requires the following changes to the PVNGS licensing basis:

• Technical Specification (TS) Section 4.2.1 -Addition of Optimized ZIRLO' fuel rod cladding material as an acceptable fuel rod cladding material

• TS Section 5.6.5 - Addition to the Core Operating Limits Report (COLR) of analytical methods previously reviewed and approved by the NRC

• A License Condition to address Information Notice (IN) 2012-09

• A permanent exemption from certain requirements of 10 CFR 50.46, Acceptance

-Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50 Appendix K, EGGS Evaluation Models

3. TECHNICAL SPECIFICATIONS CHANGES The proposed use of NGF requires changes to TS 4.2.1 and TS 5.6.5. The following subsections address these changes to the PVNGS licensing basis.

Mark-ups of the affected TS pages are provided in Attachment 2 to this enclosure. Clean TS pages are provided in Attachment 3 to this enclosure.

Mark-ups of the affected Technical Specification Bases pages are provided in Attachment 4 for information in support of the TS changes.

3.1. TS 4.2.1 - Design Features (Optimized ZIRLO')

The proposed change will revise the PVNGS Units 1, 2, and 3 Operating Licenses to allow the use of Optimized ZIRLO' fuel rod cladding material. Acceptable fuel rod cladding material is identified in Technical Specification 4.2.1, Fuel Assemblies. The proposed change will add 2

Enclosure Description and Assessment of Proposed License Amendment Optimized ZIRLO' fuel rod cladding material as an acceptable material consistent with the permanent exemption request presented in Section 7 of this license amendment request.

3.2. TS 5.6.5.b - Core Operating Limits Report (COLR) Analytical Methods Technical Specification 5.6.5.b, Core Operating Limits Report (COLR), lists the documents which describe the COLR analytical methods used to determine the core operating limits presented in each PVNGS unit specific COLR.

The proposed change will revise TS 5.6.5.b to add the following NRC approved topical reports to the list of referenced core operating analytical methods for consistency with the analytical methods that will be used to determine the core operating limits following NGF, VIPRE-W Code, Critical Heat Flux (CHF) Correlation, and Zirconium Diboride burnable absorber methodology implementation:

• WCAP-16500-P-A, CE 16x16 Next Generation Fuel Core Reference Report

• WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis

• CENPD-387-P-A, ABB Critical Heat Flux Correlations for PWR Fuel

• WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes

• WCAP-16072-P-A, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs Sections 1 through 5 of Attachment 5 to this enclosure provide an assessment as to how the limitations and conditions contained in the NRC Safety Evaluations for these topical reports are met.

Topical report CENPD-404-P-A, Implementation of ZIRLO' Cladding Material in CE Nuclear Power Fuel Assembly Designs, is currently addressed in TS 5.6.5.b as Item 13. Section 6 of to this enclosure addresses how CENPD-404-P-A, Addendum 1-A and Addendum 2-A will be implemented with the introduction of NGF.

Topical report CENPD-183-A, Loss of Flow, C-E Methods for Loss of Flow Analysis, is currently addressed in TS 5.6.5.b as Item 19. Section 7 of Attachment 5 to this enclosure addresses how topical report CENPD-183-A will be implemented differently with the introduction of NGF.

Clean TS pages are provid~d in Attachment 3. PVNGS has adopted Technical Specification Task Force (TSTF) 363, Revise Topical Report References in ITS 5.6.5, COLR, (Reference 8) in Amendment No. 137 to the PVNGS Operating Licenses; therefore, the proposed change to TS 5.6.5.b, identifies the documents by number and title.

Each PVNGS unit specific COLR specifies the complete identification (i.e., report number, title, revision, date, and any supplements) for each of the Technical Specification referenced topical reports used to prepare the COLR. The following table describes each technical specification reference number, document, title, revision, and date for the additional proposed COLR reference additions to TS 5.6.5.b. Each PVNGS unit specific COLR will be updated as part of the implementation of the approved license amendment.

3

Enclosure Description and Assessment of Proposed License Amendment T.S.

Title Re~ort No. Rev Date §!mm Ref#

Optimized ZIRLOTM CENPD-404-P-A, 0 July 2006 N.A.

13 (N001-0203-00611) Addendum 1-A Westin~house Clad Corrosion Model for CENPD-404-P-A, 0 October N.A.

13 ZIRLO and Optimized ZIRLO' Addendum 2-A 2013 (N001-0205-00006)

CE 16x16 Next Generation Fuel Core WCAP-16500-P-A 0 August N.A.

22 Reference Report (N001-0203-00614) 2007 Application of CE Setpoint Methodology for WCAP-16500-P-A 1 December 1 22 CE 16x16 Next Generation Fuel (NGF) 2010 (N001-0205-00063)

Evolutionary Design Changes to CE 16x16 WCAP-16500-P-A 1 June 2016 2-P-A Next Generation Fuel and Method for 22 Addressing the Effects of End-of-Life Properties on Seismic and Loss of Coolant Accident Analyses (N001-0205-00048)

VIPRE-01 Modeling and Qualification for WCAP-14565-P-A 0 October N.A.

Pressurized Water Reactor Non- LOCA 1999 23 Thermal-Hydraulic Safety Analysis (NOO 1-0205-00002)

Addendum 1 to WCAP-14565-P-A WCAP-14565-P-A, 0 August N.A.

Qualification of ABB Critical Heat Flux Addendum 1-A 2004 23 Correlations with VI PRE-01 Code (N001-0205-00003)

Addendum 2 to WCAP-14565-P-A, Extended WCAP-14565-P-A, 0 April 2008 N.A.

Applications of ABB-NV Correlation and Addendum 2-P-A 23 Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications (N001-0205-00004)

ABB Critical Heat Flux Correlations for PWR CENPD-387-P-A 0 May 2000 N.A.

24 Fuel (MN725-A02044)

Westinghouse Correlations WSSV and WCAP-16523-P-A 0 August N.A.

WSSV-T for Predicting Critical Heat Flux in 2007 25 Rod Bundles with Side-Supported Mixing Vanes (N001-0203-00615)

Implementation of Zirconium Diboride WCAP-16072-P-A 0 August N.A.

Burnable Absorber Coatings in CE Nuclear 2004 26 Power Fuel Assembly Designs (N001-0205-00226)

4. LICENSE CONDITION TO ADDRESS INFORMATION NOTICE 2012-09 Appendix A to NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition (SRP), Section 4.2, Fuel System Design (Reference 6),

provides NRC review guidance for the evaluation of fuel assembly structural response to '

externally applied forces. The review guidance contained in SRP 4.2 indicates it is acceptable to assume that fuel spacer grid strength at the beginning-of-life is most limiting. However, NRC Information Notice (IN) 2012-09, Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength (Reference 7), states that Operating Experience (OE) regarding the effects of in-reactor service on fuel assembly component response to externally applied forces (i.e.,

4

Enclosure Description and Assessment of Proposed License Amendment earthquakes and postulated pipe breaks in the reactor coolant system) challenge this existing NRC staff guidance. Specifically, OE shows that the crush strength of fuel assembly spacer grids may decrease during the life of a fuel assembly due to the effects of irradiation.

The Pressurized Water Reactor Owners Group (PWROG) is proposing a resolution for the IN 2012-09 issue for Westinghouse and Combustion Engineering fuel designs that will be submitted for NRC review and approval. As discussed in the NRC public meeting held with APS on March 24, 2016 (ADAMS accession number ML16088A060), since a resolution has not yet been approved, APS is requesting the following text for an IN 2012-09 license condition:

Information Notice 2012-09 APS will incorporate NRG-approved guidance into the current licensing basis regarding fuel assembly integrity under externally applied forces as described in IN 2012-09 within 2 cycles following Mode 4 entry with the first NGF transition core. APS will notify the NRG when this action is completed.

5. REGULATORY ANALYSIS FOR TECHNICAL SPECIFICATIONS CHANGES 5.1. Applicable Regulatory Requirements The proposed changes have been evaluated to determine whether applicable regulations and requirements continue to be met.

Title 10 of the Code of Federal Regulations (1 O CFR) Paragraph 50.36(c)(2)(ii) requires that TS limiting conditions for operation be established for process variables, design features, or operating restrictions for which a value is assumed as an initial condition of a design basis accident or transient analysis in the licensee's safety analyses. To eliminate the need for an amendment to update the cycle-specific parameter limits for each fuel cycle while complying with 10 CFR 50.36(c)(2)(ii) requirements, the cycle-specific parameter limits are incorporated in the PVNGS unit specific COLR.

The proposed change adds to TS 5.6.5.b several Westinghouse topical reports which have been reviewed and approved by the NRC for licensing application. The addition of these topical reports to the PVNGS Units 1, 2, and 3 TS is consistent with the current TS 5.6.5.b practice established in Amendment 137 to the PVNGS Operating Licenses (Reference 5, Section 2.5) to identify the documents by number and title. The proposed change to the PVNGS unit specific COLR specifies the complete identification (i.e., report number, title, revision, date, and any supplements) for each of the TS referenced topical reports used to prepare the COLR.

5.2. Precedent Precedent has been established for the use of Next Generation Fuel (NGF) in pressurized water reactors (PWRs). Specifically, Arkansas Nuclear One (ANO) Unit 2 was approved for the use of NGF as documented in ADAMS accession numbers ML092460118, ML080840015, ML080840023, and ML080370016. Similarly, Waterford Steam Electric Station Unit 3 was approved for the use of NGF as documented in ADAMS accession numbers ML092880237, ML080880014, ML080880015, ML081260523, and ML080380005.

5

Enclosure Description and Assessment of Proposed License Amendment 5.3. No Significant Hazards Consideration Determination The proposed changes will modify the Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 Technical Specifications (TS) with changes to TS Section 5.6.5 and TS Section 4.2.1.

TS Section 5.6.5.b lists the documents which describe the analytical methods used to determine the core operating limits. The proposed change will revise TS 5.6.5.b by updating the references to be consistent with the analytical methods that will be used to determine the core operating limits following Next Generation Fuel (NGF) implementation. The proposed change will add the following topical reports to the list of referenced core operating analytical methods to be consistent with the analytical methods that will be used to determine the core operating limits following NGF, VIPRE-W Code, Critical Heat Flux (CHF) Correlation, and Zirconium Diboride burnable absorber methodology implementation.

• WCAP-16500-P-A, CE 16x16 Next Generation Fuel Core Reference Report

• WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis

• CENPD-387-P-A, ABB Critical Heat Flux Correlations for PWR Fuel

• WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes

• WCAP-16072-P-A, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs WCAP-16500-P-A (Next Generation Fuel)

Topical report WCAP-16500-P-A describes the 16x16 lattice NGF assembly mechanical design for the CE nuclear steam supply system (NSSS) and the methods and models used for evaluating its acceptability. In addition to the reference product description, this topical report describes the fuel mechanical and.reload design methodology intended to support fuel design and licensing applications up to a rod average burnup of 62 Gigawatt Days per Metric Ton Uranium (GWd/MTU). APS has demonstrated that the limitations and conditions con~ained in the NRC Safety Evaluation for this topical report will be met as described in Section 1.1 of Attachment 5 to this enclosure.

Topical report WCAP 16500-P-A, Supplement 1, Revision 1, describes a revised analytical process for calculating Core Operating Limit Supervisory System (COLSS) and Core Protection Calculator System (CPCS) addressable constants and database constants for plant reloads with CE 16x16 NGF (CE16NGF') assemblies. Per Section 1.2 of Attachment 5 to this enclosure, there are no limitations and conditions contained in the NRC Safety Evaluation for this topical report.

Topical report WCAP 16500-P-A, Supplement 2, describes three evolutionary changes to the CE16NGF' spacer grid designs being pursued that, in combination, will further improve the grid-to-rod fretting (GTRF) resistance of the fuel and its resistance to crud formation while improving the fabricability of the spacer grid straps. The changes to the design are all related to the mid and intermediate flow-mixing (IFM) spacer grids. All other features of the CE 16x16 NGF design remain the same as described in WCAP-16500-P-A.

6

Enclosure Description and Assessment of Proposed License Amendment APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be met as described in Section 1.3 of Attachment 5 to this enclosure.

WCAP-14565-P-A (VIPRE-W Code)

Topical report WCAP-14565-P-A, describes Westinghouse VIPRE (i.e., VIPRE-W) modeling and qualification for PWR thermal hydraulic (T-H) design and non-LOCA safety analysis. APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be met as described in Section 2.1 of Attachment 5 to this enclosure.

Topical Report WCAP-14565-P-A, Addendum 1-A, has installed ABB-NV and ABB-TV CHF correlations into its version of the VIPRE (i.e., VIPRE-W) code. The ABB-NV correlation is for Combustion Engineering designed PWR (CE-PWR) 14x14 and 16x16 fuels with non-mixing vane grids, and ABB-TV is for the 14x14 Turbo fuel with mixing vane grids. APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be met as described in Section 2.2 of Attachment 5 to this enclosure.

Topical report WCAP-14565-P-A, Addendum 2-P-A, presents a modification to the ABB-NV correlation based on rod bundle data at low pressure and low flow conditions. The modified low pressure ABB-NV correlation is designated as the Westinghouse Low Pressure (WLOP) correlation in this report. The WLOP correlation predicts more accurate DNBR than either W-3 or MacBeth correlation at the low pressure and low flow conditions. APS has demonstrated that the limitations and conditions contained.in the NRC Safety Evaluation for this topical report will be met as described in Section 2.3 of Attachment 5 to this enclosure.

CENPD-387-P-A (ABB CHF Correlation)

Topical report CENPD-387-P-A provides a description of the PWR CHF correlations for ABB Combustion Engineering (ABB-CE) PWR 14x14 and 16x16 fuels. The ABB-NV correlation is for ABB-CE PWR 14x14 and 16x16 fuels with non-mixing vane grids, and the ABB-TV correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS. APS has demonstrated that the limitations and conditions contained in the NRC Safety .Evaluation for this topical report will be met as described in Section 3 of Attachment 5 to this enclosure.

WCAP-16523-P-A (WSSV CHF Correlation)

Topical report WCAP-16523-P, describes the development of CHF correlations for PWR fuel designs containing structural mixing vane grids and intermediate flow mixer grids with side-supported vanes. The correlations, WSSV and WSSV-T, are for 14x14 and 16x16 fuel designs containing side-supported vane grids for CE-PWRs. Both correlations utilize the same form, but with different coefficients. The WSSV correlation coefficients were derived with the Westinghouse version of the VIPRE-01 (i.e., VIPRE-W) subchannel code. The WSSV-T correlation coefficients were derived with the CE TORC subchannel code. APS has demonstrated that the limitations and conditions 7

Enclosure Description and Assessment of Proposed License Amendment contained in the NRC Safety Evaluation for this topical report will be met as described in Section 4 of Attachment 5 to this enclosure.

WCAP-16072-P-A CIFBA)

Topical report WCAP-16072-P-A, describes the use of Zirconium Diboride (ZrB 2 ) Integral Fuel Burnable Absorber (IFBA) in Westinghouse fuel assembly designs. APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be met as described in Section 5 of Attachment 5 to this enclosure.

CENPD-404-P-A CZIRLO and Optimized ZIRLO')

Topical report CENPD-404-P-A is currently addressed in TS 5.6.5.b as Item 13. The proposed change addresses how CENPD-404-P-A, Addendum 1-A, and Addendum 2-A, will be implemented with the introduction of NGF.

Topical report CENPD-404-P-A, Addendum 1-A, describes an extension of the regulatory definition of ZIRLO to allow for the optimization of ZIRLO for enhanced corrosion resistance in more adverse in-reactor primary chemistry environments and at higher fuel duties with higher burnups. APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be n:iet as described in Section 6.1 of Attachment 5 to this enclosure.

Topical report CENPD-404-P-A, Addendum 2-A, describes the Westinghouse ZIRLO fuel rod cladding corrosion model and Optimized ZIRLO' corrosion model. The existing ZIRLO corrosion model is based on a corrosion model originally developed for Zircaloy-4 cladding. Comparison of the measured cladding oxide thickness on ZIRLO fuel rods to the measured cladding oxide thickness ofZircaloy-4 fuel rods with similar power histories indicated that the ZIRLO cladding corrosion rate was a specific fraction of the Zircaloy-4 cladding corrosion rate. Thus, the existing ZIRLO best estimate fuel rod corrosion model is a multiplier applied to the Zircaloy-4 fuel rod corrosion model.

APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be met as described in Section 6.2 of Attachment 5 to this enclosure.

CENPD-183-A Closs of Flow Analysis)

Topical report CENPD-183-A is currently addressed in TS 5.6.5.b as Item 19. The proposed change addresses how topical report CENPD-183-A will be implemented differently with the introduction of NGF.

Topical report CENPD-183-A describes the assumptions, conservatisms and basic methods used for analyzing loss of reactor coolant forced flow events. The main body of the report describes a loss of flow analysis method for use with a computer code having transient core thermal hydraulic capabilities (referred to as the dynamic method). The appendix describes a similar loss of flow analysis method for use with a steady state core thermal hydraulic code (referred to as the static method). APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for this topical report will be met as described in Section 7 of Attachment 5 to this enclosure.

8

Enclosure Description and Assessment of Proposed License Amendment Assumptions used for accident initiators and/or safety analysis acceptance criteria are not altered by the addition of these topical reports.

Use of the referenced methodologies will support implementation of CE16NGF' fuel. The fuel design is intended to provide improved fuel reliability by reducing grid-to-rod fretting issues, improved fuel performance for high duty operation, and enhanced operating margin.

TS Section 4.2.1 lists acceptable fuel rod cladding material. The proposed change adds Optimized ZIRLO' fuel rod cladding material as an acceptable material to allow the use of NGF consistent with the permanent exemption request presented in Section 7 of this license amendment request (LAR).

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

Response: No.

The proposed change to TS Section 4.2.1 adds Optimized ZIRLO' fuel rod cladding material as an acceptable material consistent with the permanent exemption request presented in Section 7 of this LAR.

The NRC approved topical report CENPD-404-P-A, Addendum 1-A and Addendum 2-A addresses Optimized ZIRLO' and demonstrates that Optimized ZIRLO' has essentially the same properties as currently licensed ZIRLO. The fuel cladding itself is not an accident initiator and does not affect accident probability. Use of Optimized ZIRLO' fuel cladding has been shown to meet all 10 CFR 50.46 design criteria and, therefore, will not increase the consequences of an accident.

Therefore, the proposed change to TS Section 4.2.1 does not involve a significant increase in the probability or consequences of an accident previously evaluated.

The proposed changes to TS Section 5.6.5 have no impact on any plant configuration or system performance. Changes to the calculated core operating limits may only be made using NRC approved methodologies, must be consistent with all applicable safety analysis limits, and are controlled by the 10 CFR 50.59 process. The proposed changes to TS Section 5.6.5 will add the NRC approved topical reports, as described, to the list of referenced core operating analytical methods. APS has demonstrated that the limitations and conditions contained iri the NRC Safety Evaluation for these topical reports, and their various supplements and revisions will be met as described in Attachment 5 to this enclosure.

Therefore, the proposed change to TS Section 5.6.5 does not involve a significant increase in the probability or consequences of an accident previously evaluated.

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

Response: No.

The proposed change to TS Section 4.2.1 adds Optimized ZIRLO' fuel rod cladding material as an acceptable material consistent with the permanent exemption request presented in Section 7 of this LAR.

9

Enclosure Description and Assessment of Proposed License Amendment Use of Optimized ZIRLO' clad fuel will not result in changes in the operation or configuration of the facility. Topical report CENPD-404-P-A demonstrated that the material properties of Optimized ZIRLO' are similar to those of standard ZIRLO. Therefore, Optimized ZIRLO' fuel rod cladding will perform similarly to those fabricated from standard ZIRLO thus precluding the possibility of the fuel becoming an accident initiator and causing a new or different type of accident.

Therefore, the proposed change to TS Section 4.2.1 does not create the possibility of a new or different kind of accident from any previously evaluated.

The proposed changes to TS Section 5.6.5 have no impact on any plant configuration or system performance. Changes to the calculated core operating limits may only be made using NRC approved methodologies, must be consistent with all applicable safety analysis limits, and are controlled by the 10 CFR 50.59 process. The proposed changes to TS Section 5.6.5 Will add the NRG-approved topical reports, as described, to the list of referenced core operating analytical methods. APS has demonstrated that the limitations and conditions contained in the NRC Safety Evaluation for these topical reports, and their various supplements and revisions as identified in Attachment 5 to this enclosure, will be met as described in Section 3.2.

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

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

Response: No.

Tlie proposed change to TS Section 4.2.1 adds Optimized ZIRLO' fuel rod cladding material as an acceptable material consistent with the permanent exemption request presented in Section 7 of this LAR.

The proposed change will not involve a significant reduction in the margin of safety because it has been demonstrated that the material properties of the Optimized ZIRLO' are not significantly different from those of standard ZIRLO. Optimized ZIRLO' is expected to perform similarly to standard ZIRLO for all normal operating, transient, and accident scenarios, including both loss of coolant accident (LOCA) and non-LOCA scenarios. For LOCA scenarios, where the slight difference in Optimized ZIRLO' material properties relative to standard ZIRLO could have some impact on the overall accident scenario, plant-specific LOCA analyses using Optimized ZIRLO' properties were performed. These LOCA analyses demonstrate that the acceptance criteria of 10 CFR 50.46 are satisfied when Optimized ZIRLO' fuel rod cladding is implemented.

Therefore, the proposed change to TS Section 4.2.1 does not involve a significant reduction in a margin of safety.

The proposed changes to TS Section 5.6.5 have no impact on any plant configuration or system performance. The proposed changes to TS Section 5.6.5 will add the NRG-approved topical reports, as described, to the list of referenced core operating analytical methods. The proposed changes do not amend the cycle specific parameter limits located in the PVNGS unit specific COLR from the values presently required by the TS. The individual specifications continue to require operation of the plant within the bounds of the limits specified in PVNGS unit specific COLR.

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

10

Enclosure Description and Assessment of Proposed License Amendment 5.4. Conclusions APS concludes that operation of the facility in accordance with the proposed amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92(c),

and, accordingly, a finding of "no significant hazards consideration" is justified. 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.

6. ENVIRONMENTAL EVALUATION The proposed amendment does not involve (i) a significant hazards consideration, (ii) a significant change in the types or a significant increase in the amounts of any effluents that may be released offsite, or (iii) a significant increase in individual or cumulative occupational radiation exposure. Accordingly, tile proposed amendment meets the eligibility criterion for categorical exclusion set forth in 10 CFR 51.22(c)(9). Therefore, pursuant to 10 CFR 51.22(b),

no environmental impact statement or environmental assessment needs to be prepared in connection with the proposed amendment.

7. PERMANENTEXEMPTION-10 CFR 50.46AND10 CFR PART 50 APPENDIX K FOR OPTIMIZED ZIRLO' The proposed change will revise the PVNGS Units 1, 2, and 3 Operating Licenses to allow the use of Optimized ZIRLO' fuel rod cladding material. Acceptable fuel rod cladding material is identified in PVNGS TS Section 4.2.1, Fuel Assemblies. The proposed change will add Optimized ZIRLO' fuel rod cladding material as an acceptable material. A permanent exemption from certain requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50 Appendix K, EGGS Evaluation Models, is required to support this change. By letter dated August 26, 2010, the NRC staff approved a temporary exemption from these requirements to support the PVNGS NGF Lead Fuel Assembly (LFA) program (ADAMS Accession No. ML 101,900254). The requested permanent exemption will replace the approved temporary exemption.

Part 50.46(a)(l)(i) of Title 10 of the Code of Federal Regulations (10 CFR 50.46(a)(l)(i)) states in part:

"Each boiling or pressurized light-water nuclear power reactor fueled with uranium oxide pellets within cylindrical Zircaloy or ZIRLO cladding must be provided with an emergency core cooling system (EGGS) that must be designed so that its calculated cooling performance following postulated loss-of-coolant accidents conforms to the criteria set forth in paragraph (b) of this section. EGGS cooling performance must be calculated in accordance with an acceptable evaluation model and must be calculated for a number of postulated loss-of-coolant accidents of different sizes, locations, and other properties sufficient to provide assurance that the most severe postulated loss-of-coolant accidents are calculated."

11

Enclosure Description and Assessment of Proposed License Amendment 1 O CFR 50.46 continues with a delineation of specifications for peak cladding temperature, maximum hydrogen generation, coolable geometry, and long-term cooling. Since 10 CFR 50.46 specifically refers to fuel with Zircaloy or ZIRLO cladding and does not list Optimized ZIRLO' cladding, the use of Optimized ZIRLO' cladding requires a permanent exemption from this section of the regulations.

10 CFR 50, Appendix K, paragraph l.A.5, states in part:

"The rate of energy release, hydrogen generation, and cladding oxidation from the metal/water reaction shall be calculated using the Baker-Just equation."

The Baker-Just equation presumes the use of Zircaloy or ZIRLO cladding. The routine use of Optimized ZIRLO' cladding requires a permanent exemption from this section of the regulations.

Pursuant to 10 CFR 50.12, Specific Exemptions, APS is requesting a permanent exemption from the requirements of 10 CFR 50.46, Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Nuclear Power Reactors, and 10 CFR 50, Appendix K, EGGS Evaluation Models, for PVNGS Units 1, 2 and 3.

The permanent exemption will allow the use of fuel assemblies manufactured by Westinghouse with Optimized ZIRLO' alloy clad fuel rods, consistent with NRC approved APS and Westinghouse design and analysis methodologies.

Currently, numerous pressurized water reactors (PWRs) in the United States have used Optimized ZIRLO' clad fuel assemblies either in full batch or LFA programs, including Arkansas Nuclear One Unit 2 and Waterford Steam Electric Station Unit 3 (which are also Combustion Engineering Nuclear Steam Supply System 16x16 plants).

Background

With the NRC's proposed changes to 10 CFR 50.46, the corrosion performance requirements for the nuclear fuel cladding become more demanding. Optimized ZIRLO' provides enhanced corrosion resistance in more adverse in-reactor primary chemistry environments and at higher fuel duties with higher burnups.

Optimized ZIRLO' fuel cladding is different from standard ZIRLO in two respects: 1) the tin content is lower; and 2) the microstructure is different. This difference in tin content and microstructure can lead to differences in some material properties. Most of the material properties of standard ZIRLO' and Optimized ZIRLO' are the same within the uncertainty of the data and therefore use of standard ZIRLO properties for safety analyses is acceptable.

However, the NRC Safety Evaluation for Optimized ZIRLO' suggests that the computer codes used to perform fuel design safety analyses incorporate the material properties of Optimized ZIRLO'.

Technical Justification of Acceptability Optimized ZIRLOTM is described in topical report CENPD-404-P-A, Addendum 1-A, and Addendum 2-A, which have been reviewed and approved by the NRC (References 2 and 3).

The NRC staff approved Optimized ZIRLO' fuel cladding based on: 1) similarities with standard ZIRLO; 2) demonstrated material performance; and 3) a commitment to provide irradiated data and validated fuel performance models ahead of burnups achieved in batch application. The NRC Safety Evaluations for the addenda to this topical report included limitations and conditions. These limitations and conditions are addressed in Section 6 of to this enclosure.

12

Enclosure Description and Assessment of Proposed License Amendment The core reload evaluations will ensure that acceptance criteria are met for the insertion of assemblies with fuel rods clad with Optimized ZIRLOTM. These assemblies will be evaluated using NRC approved methods and models to address the use of Optimized ZIRLO'.

Justification of Exemption The standards set forth in 10 CFR 50.12 provide that the Commission may grant exemptions from the requirements of the regulations for reasons consistent with the following:

• The exemption is authorized by law;

• The exemption will not present an undue risk to the public health and safety;

• The exemption is consistent with the common defense and security; and

• $pecial circumstances are present.

This exemption is authorized by law. *The selection of a specified cladding material in 10 CFR 50.46 and implied in 10 CFR Part 50, Appendix K, was adopted at the discretion of the Commission consistent with its statutory authority. No statute required the NRC to adopt this specification. Additionally, the NRC has the authority under Section 50.12 to grant exemptions from the requirements of Part 50 upon showing proper justification. Further, it should be noted that, by submitting this exemption request, APS does not seek an exemption from the acceptance and analytical criteria of 10 CFR 50.46 and 10 CFR Part 50, Appendix K. Ttie intent of the request is solely to allow the use of Optimized ZIRLO' as fuel rod cladding material in lieu of Zircaloy or ZIRLO. Therefore, as required by 10 CFR 50.12 (a)(1), this requested exemption is "authorized by law."

The exemption will not present an undue risk to public health and safety. The NRC staff approved Westinghouse topical report CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO' (Reference 2), and Addendum 2-A, Westinghouse Clad Corrosion Model for ZIRLO and .

Optimized ZIRLO' (Reference 3).

The Optimized ZIRLO' topical report demonstrates that predicted chemical, mechanical, and material performance characteristics of the Optimized ZIRLO' alloy cladding are within those approved for Zircaloy and ZIRLO under anticipated operational occurrences and postulated accidents. Topical report CENPD-404-P-A, Addendum 1-A, demonstrates that Optimized ZIRLO' has essentially the same properties as currently licensed ZIRLO. The safety analysis for PVNGS Units 1, 2, and 3 is supported by site specific Technical Specifications (TS). Reload cores are required to be operated in accordance with the operating limits specified in the TS.

Normal reload design and analysis methodologies in use at APS and at the fuel vendor will evaluate the Optimized ZIRLO' clad fuel susceptibility to failure during normal operation,.

anticipated operational occurrences and postulated accidents for each core design. Optimized ZIRLO' fuel performance (as well as any co-resident Zircaloy and/or ZIRLO fuel) in each core design will be evaluated and any predicted fuel failures will be limited such that dose .

consequence impacts are within the applicable regulatory limits. Therefore, the use of Optimized ZIRLO' clad will not present an undue risk to the public health and safety.

The exemption is consistent with the common defense and security. As previously noted, the exemption request is only to allow the application of the aforementioned regulations to Optimized ZIRLO', an improved fuel rod cladding material. All the requirements and acceptance criteria will be maintained. The special nuclear material in these assemblies is required to be handled and controlled in accordance with approved procedures. Use of Optimized ZIRLO' fuel rod cladding will not affect plant operations and is consistent with 13

Enclosure Description and Assessment of Proposed License Amendment common defense and security. Therefore, the common defense and security are not impacted by this exemption request.

Special circumstances are present. 10 CFR 50.12(a)(2) states that the NRC will not consider granting an exemption to the regulations unless special circumstances are present. The requested exemption meets the special circumstances set forth in 10 CFR 50.12(a)(2)(ii), which states that special circumstances are present whenever:

"Application of the regulation in the particular circumstances would not seNe the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule ... "

10 CFR 50.46 identifies acceptance criteria for ECCS system performance at nuclear power facilitie~. The effectiveness of the ECCS will not be affected by the use of Optimized ZIRLO' clad fuel assemblies. Due to the similarities in the material prop~rties of the Optimized ZIRLOTM alloy to ,Zircaloy or ZIRLO as identified in the Optimized ZIRLO' topical report it can be concluded that the ECCS effectiveness would not be adversely affected. Westinghouse has performed evaluations using approved Loss of Coolant Accident (LOCA) methods to ensure that assemblies with Optimized ZIRLO' fuel rod cladding material meet all LOCA safety criteria.

The intent of paragraph l.A.5 of Appendix K to 10 CFR 50 is to apply an equation for rates of energy release, hydrogen generation, and cladding oxidation from the metal-water reaction that conservatively bounds all post-LOCA scenarios (i.e., the Baker-Just equation). The supporting documentation for the Optimized ZIRLO' topical report shows that due to the similarities in the composition of the Zircaloy or ZIRLO and Optimized ZIRLO', the application of the Baker-Just equation will continue to conservatively bound all post-LOCA scenarios.

The regulations of 10 CFR 50.46 and 10 CFR Part 50, Appendix K, make no provision for use of fuel rods clad in a material other than Zircaloy or ZIRLO. Since the chemical composition of the Optimized ZIRLO' alloy differs from the specifications for Zircaloy or ZIRLO, a plant-specific exemption is required to allow the use of the Optimized ZIRLO' alloy as a cladding material at PVNGS. The expected performance of Optimized ZIRLOTM clad material meets the intent of the regulations, as discussed in the Optimized ZIRLO' topical report.

Therefore, application of these regulations in this particular circumstance would not serve the underlying purpose of the rule and is not necessary to achieve the underlying purpose of the

• rule, so special circumstances exist.

Conclusion In order to support the use of Optimized ZIRLO' fuel rod cladding material, an exemption from the requirements of 10 CFR 50.46 and 10 CFR Part 50, Appendix K is requested. As required by 10 CFR 50.12, the requested exemption is authorized by law, does not present undue risk to public health and safety, and is consistent with common defense and security. Approval of this exemption request does not violate the underlying purpose of the rule. In addition, special circumstances do exist to justify the approval of an exemption from the subject requirements.

14

Enclosure Description and Assessment of Proposed License Amendment

8. REFERENCES

1. CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties, May 1988

2. CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO', July 2006

3. CENPD-404-P-A, Addendum 2-A, Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO', October 2013

4. WCAP-16072-P-A, Revision 0, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, August 2004

5. Letter from L. Raynard Wharton (USNRC) to Gregg R. Overbeck (APS) of October 15, 2001, Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments Re:

Various Administrative Controls (TAC Nos. MB1668, MB1669, and MB1670) (ADAMS Accession No. ML012880473)

6. NUREG-0800, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, (SRP) Section 4.2, Fuel System Design, Revision 3, March 2007 (ADAMS Accession No. ML070740002)

7. Information Notice (IN) 2012-09, Irradiation Effects on Fuel Assembly Spacer Grid Crush Strength, June 28, 2012 (ADAMS Accession No. ML113470490)

8. Technical Specification Task Force (TSTF) Traveler Number 363, Revision 0, Revise Topical Report References in ITS 5.6.5, COLR, (ADAMS Accession No. ML040630088) 15

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 1 License Condition and Regulatory Commitment The following table identifies the License Condition requested in this document.

TYPE SCHEDULED COMPLETION LICENSE CONDITION continuing one-time DATE compliance (if applicable)

[1.] APS will incorporate NRG-approved guidance x Within 2 cycles into the current licensing basis regarding fuel following assembly integrity under externally applied forces as Mode 4 entry described in IN 2012-09 within 2 cycles following with the first Mode 4 entry with the first NGF transition core. APS NGF transition will notify the NRC when this action is completed. core.

The following table identifies the regulatory commitment in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes arid are not considered to be regulatory commitments.

TYPE SCHEDULED COMPLETION REGULATORY COMMITMENT continuing one-time DATE compliance (if applicable)

[1.] A fuel centerline temperature allowance at high x Upon burn up will be set aside to account for the burnup implementation dependent effects of Thermal Conductivity ofNGF Degradation (TCD) when using the FATES3B code to determine input for NGF non-LOCA and LOCA safety analyses.

ATTACHMENT 1, Page 1

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 2 Technical Specifications Page Mark-ups Affected Pages: 4.0-1 and 5.6-7

Design Features 4.0 4.0 DESIGN FEATURES 4.1 Site Location The Palo Verde Nuclear Generating Station is located in Maricopa County, Arizona. approximately 50 miles west of the Phoenix metropolitan area.

The site is comprised of approximately 4,050 acres. Site elevations range from 890 feet above mean sea level at the southern boundary to 1,030 feet above mean sea level at the northern boundary. The minimum distance from a containment building to the exclusion area boundary is 871 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor shall contain 241 fuel assemblies. Each assembly shall consist of a matrix of Zircaloy or ZIRLO or Optimized ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide (U02 ) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods. in accordance with approved applications of fuel rod configurations. may be used. Fuel assemblies shall be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions. Other cladding material may be used with an approved exemption.

4.2.2 Control Element Assemblies The reactor core shall contain 76 full strength and 13 part strength control element assemblies CCEAs).

The control section for the full strength CEAs shall be either boron carbide with Alloy 625 cladding, or a combination of silver-indium-cadmium and boron carbide with Alloy 625 cladding.

The control section for the part strength CEAs shall be solid Alloy 625 slugs with Alloy 625 cladding.

(continued)

PALO VERDE UNITS 1.2.3 4.0-1 AMENDMENT NO. +/--7-2-,

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report CCOLR) (continued) 20 . CENPD-382-P-A. "Methodology for Core Designs Containing Erbium Burnable Absorbers ." [Methodology for Specifications 3.1.1. Shutdown Margin-Reactor Trip Breakers 0Den: 3.1.2. Shutdown Margin-Reactor Trip Breakers Closed: and 3.1.4. Moderator Temperature Coefficient.]

21. CEN-386-P-A. "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16 x 16 PWR Fuel." [Methodology for Specifications 3.1.1. Shutdown Margin-Reactor Trip Breakers Open:

3.1.2.Shutdown Margin-Reactor TriR Breakers Closed: and 3.1.4. Moderator Temperature Coefficient.]

22. WCAP-16500-P-A. "CE 16x16 Next Generation Fuel Core Reference Report. [Methodolo)y tor Specifications 2.1.1.

Reactor Core SLs: 3.2.4. DNBR

23. WCAP-14565-P-A. "VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LUCA lhermal-Hydraulic Safety Analysis. [Methodology tor Specifications 2.1.1.

Reactor Core SLs: 3.2.4. DNBRJ

24. CENPD-387 -P-A. "ABB Critical Heat Flux Correlations for PWR Fuel. [Methodolo)y tor Specifications 2.1 .1. Reactor Core SLs: 3.2.4. DNBR

25. WCAP-16523-P-A. "Westinghouse Correlations WSSV and WSSV-1 tor Predictin~ Critical Heat Flux in Rod Bundles with Side-Supported ixing vanes . [Methodology tor Specifications 2.1.1. Reactor Core SLs: 3.2 .4. DNBRJ

26. WCAP -16072-P-A. "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Desi5ns . [Methodolo)y tor Specifications 2.1.1.

Reactor Core Ls: 3.2.4. DNBR

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 CECCS) limits. nuclear limits such as SOM. transient analysis limits. and accident analysis limits) of the safety analysis are met .

d. The COLR. including any mid cycle revisions or supplements.

shall be provided upon issuance for each reload cycle to the NRC.

(continued)

PALO VERDE UNITS 1.2.3 5.6-7 AMENDMENT NO. +/-+4.

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 3 Clean Technical Specifications Pages Affected Pages: 4.0-1 and 5.6-7

Design Features 4.0 4.0 DESIGN FEATURES

4. 1 Site Locati on The Palo Verde Nuclear Generating Station is located in Maricopa County, Arizona. approximately 50 miles west of the Phoenix metropolitan area.

The site is comprised of approximately 4.050 acres . Site elevations range from 890 feet above mean sea level at the southern boundary to 1.030 feet above mean sea level at the northern boundary. The mi nimum distance from a containment building to the exclusion area boundary is 871 meters.

4.2 Reactor Core 4.2.1 Fuel Assemblies The reactor sha l l contain 241 fuel assemblies . Each assembly sha l l consist of a matrix of Zircaloy or ZIRLO or Optimized ZIRLO fuel rods with an initial composition of natural or slightly enriched uranium dioxide CU02 ) as fuel material. Limited substitutions of zirconium alloy or stainless steel filler rods for fuel rods. in accordance with approved applications of fuel rod confi gurati ons. may be used. Fue l assemblies sha ll be limited to those fuel designs that have been analyzed with applicable NRC staff approved codes and methods and shown by tests or analyses to comply with all fuel safety design bases. A limited number of lead test assemblies that have not completed representative testing may be placed in nonlimiting core regions . Other cladding material may be used with an approved exemption.

4.2.2 Control Element Assemblies The reactor core shall contain 76 full strength and 13 part strength control element assemblies CCEAs).

The control section for the full strength CEAs shall be either boron carbide with Alloy 625 cladding, or a combination of silver- i ndium-cadmi um and boron carbide with All oy 625 cl adding.

The control section for the part strength CEAs shall be solid Alloy 625 slugs with Alloy 625 cladding.

(continued)

PALO VERDE UNITS 1.2.3 4.0-1 AMENDMENT NO . ~.

Reporting Requirements 5.6 5.6 Reporting Requirements 5.6.5 Core Operating Limits Report CCOLR) (continued)

20. CENPD -382-P-A. "Methodology for Core Designs Containing Erbium Burnable Absorbers." [Methodology for Specifications 3.1.1. Shutdown Margin-Reactor Trip Breakers Open; 3.1.2. Shutdown Margin-Reactor Tr i p Breakers Closed; and 3.1 .4. Moderator Temperature Coefficient.]

21. CEN-386-P-A. "Verification of the Acceptabi l ity of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Eng i neering 16 x 16 PWR Fuel." [Methodology for Specifications 3.1.1. Shutdown Margin-Reactor Trip Breakers Open ;

3.1.2.Shutdown Margin-Reactor TriR Breakers Closed; and 3.1.4. Moderator Temperature Coefficient .]

22 . WCAP-16500-P-A. "CE 16x16 Next Generation Fuel Core Reference Report ." [Methodology for Specifications 2.1.1.

Reactor Core SLs; 3.2 .4. DNBRJ

23. WCAP -14565-P-A. "VIPRE-01 Mode l ing and Qualifica ti on for Pressurized Water Reactor Non -LOCA Thermal-Hydraulic Safety Ana lysis ." [Methodo l ogy for Specifi cati ons 2.1. 1.

Reactor Core SLs; 3.2.4. DNBRJ

24. CENPD-387-P-A. "ABB Critical Heat Flux Correlations for PWR Fuel . .. [Methodology for Specifications 2.1.1. Reactor Core SLs; 3.2.4. DNBRJ

25. WCAP-16523-P-A. "Westinghouse Correlations WSSV and WSSV -T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mi xing Vanes." [Methodology for Specifications 2.1.1. Reactor Core SLs; 3.2 .4. DNBRJ

26. WCAP -16072-P-A. "Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs." [Methodology for Specifications 2.1.1.

Reactor Core SLs; 3.2.4. DNBRJ

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 CECCS) limits. nuclear limits such as SOM. transient analysis limits. and acci dent ana lysis limits) of the safety ana lysis are met.

d. The COLR . including any mid cycle revisions or supplements.

shall be provided upon issuance for each reload cycle to the NRC .

(continued)

PALO VERDE UNITS 1.2.3 5.6-7 AMENDMENT NO . +/-74.

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 4 Technical Specification Bases Page Mark-ups (provided for Information Only)

Affected Pages:

82 .1.1-3 82 .1.1-4 83 .2.1-2 83.2.1-4 83.2.1-8 83.2.2-2 83.2.2-4 83.2.2-7 83.2.3-2 83 .2.3-4 83.2.3-10 83.2.4-2 83.2.4-4 83.2.4-9 83.2 .5-2 83.2 .5-4 83.2 .5-7

Reactor Core SLs B 2.1.1 BASES APPLICABLE h. Log Power Level - High trip; SAFETY ANALYSES (continued) i. Reactor Coolant Flow - Low trip; and

j. Steam Generator Safety Valves.

The limitation that the average enthalpy in the hot leg be less than or equal to the enthalpy of saturated liquid also ensures that the ~T measured by instrumentation used in the protection system design as a measure of the core power is proportional to core power.

The SL represents a design requirement for establishing the protection system trip setpoints identified previously.

LCO 3.2.1. "Linear Heat Rate (LHR)." and LCO 3.2.4.

"Departure From Nucleate Boiling Ratio (DNBR)." or the assumed initial conditions of the safety analyses (as indicated in the UFSAR. Ref. 2) provide more restrictive limits to ensure that the SLs are not exceeded.

SAFETY LIMITS SL 2.1.1.1 and SL 2.1.1 .2 ensure that the minimum DNBR is not less than the safety analyses limit and that fuel centerline temperature remains below melting.

Prior to NGF implementation:

The minimum value of the DNBR during normal operation and design basis AOOs is limited to 1.34. based on a statistical combination of CE-1 CHF correlation and engineering factor uncertainties. and is established as an SL . Additional factors such as rod bow and spacer grid size and placement will determine the limiting safety system settings required to ensure that the SL is maintained.

Following NGF implementation:

The minimum value of the DNBR during normal operation and desiTn basis Anticipated Operational Occurrences AOOs is imited to 1.34 using the ABB-NV correlation for the first NGF transition core. This value is based on a statistical combination of CHF correlation and engineering factor uncertainties. and is established as a SL for the first NGF transition core. For the second NGF transition core and subsequent cores with NGF. the minimum value of the DNBR during normal operation and design basis AOOs is limited to 1.25 (continued)

PALO VERDE UNITS 1.2.3 B 2.1.1-3 REVISION 37

Reactor Core SLs B 2.1.1 BASES SAFETY LIMITS usina the WSSV and ABB-NV correlations. This value is (continued) base on a stat1st1cal comb1nat1on of CHF correlat1on and eng1neer1ng factor uncertaint1es. Add1t1onal factors such as rod bow and placement w111 determine the l1m1t1ng safety system sett1ngs requ1red to ensure that the SL lS ma1nta1ned .

The WSSV and ABB-NV correlations are used in the safety and setpoint analyses. However because of ex1st1ng hardware limitat1ons. the CPC al(or1thm will reta1n the CE-1 correlat1on and the DNBR- ow trip setpo1nt and Allowable Value of 1.34. To ma1ntain cons1stency w1th the CPC setpo1nt. the safett l1m1t va 1ue w1 11 remain at 1. 34 after the f1 rst NG trans1t1on core. The adJustment to the lower DNBR I 1mi t wi 11 be made w1 th1 n the safety and setp01 nt analyses.

Maintaining the dynamically adjusted peak LHR to~ 21 kW/ft or peak fuel centerline temperature < 5080°F (decreasing by 58°F per 10.000 MWD/MTU for burnup and adjusting for burnable poisons per CENPD-382-P-A). ensures that fuel centerline melt will not occur during normal operating conditions or design AOOs.

The design melting point of new fuel with no burnable poison is 5080°F. The melting point is adjusted downward from this temperature depending on the amount of burnup and amount and type of burnable poison in the fuel. The 58°F per 10.000 MWD/MTU adjustment for burnup was accepted by the NRC in Topical Report CEN-386-P-A. "Verification of the Acceptability of a 1-Pin Burnup Limit of 60 MWD/kgU for Combustion Engineering 16xl6 PWR Fuel." August 1992.

Adjustments for burnable poisons are established based on NRC approved Topical Report CENPD-382-P-A. "Methodology for Core Designs Containing Erbium Burnable Absorbers." August 1993.

(continued)

PALO VERDE UNITS 1.2.3 B 2.1.1-4 REVISION 37

LHR B 3.2.1 BASES BACKGROUND Power distribution is a product of multiple parameters.

(continued) various combinations of which may produce acceptable power distributions. Operation within the design limits of power distribution is accomplished by generating operating limits on the LHR and Departure from Nucleate Boiling (DNB).

Proximity to the DNB condition is expressed by the Departure from Nucleate Boiling Ratio (DNBR). defined as the ratio of the cladding surface heat flu x required to cause DNB to the actual cladding surface heat flux. The minimum DNBR value during both normal operation and AOOs is the DNBR Safety Limit as calculated by the ff-..+/--applicable DNB Correlation (Ref. 3) and corrected for such factors as rod bow and grid spacers. It is accepted as an appropriate margin to DNB for all operating conditions.

There are two systems that monitor core power distribution online: the Core Operating Limit Supervisory System (COLSS) and the Core Protection Calculators (CPCs). The COLSS and CPCs that monitor the core power distribution are capable of verifying that the LHR and the DNBR do not exceed their l imits. The COLSS performs this function by continuously monitoring the core power distribution and calculating core power operating limits corresponding to the allowable peak LHR and DNBR . The CPCs perform this function by continuously calculating an actual value of DNBR and Local Power Density (LPD) for compari son with the respective trip setpoints .

The COLSS indicates continuously to the operator how far the core is from the operating limits and provides an audible alarm if an operating limit is exceeded. Such a condition signifies a reduction in the capability of the plant to withstand an anticipated transient. but does not necessarily imply an immediate violation of fuel design limits. If the margin to fuel design limits continues to decrease. the RPS ensures that the specified acceptable fuel design limits are not exceeded by initiating a reactor trip.

The COLSS continually generates an assessment of the calculated margin for specified LHR and DNBR limits. The data required for these assessments include measured incore neutron flux. CEA positions. and Reactor Coo l ant System (RCS) inlet temperature. pressure. and flow.

(continued)

PALO VERDE UNITS 1.2.3 B 3.2.1-2 REVISION +/-.Q.

LHR B 3.2.1 BASES APPLICABLE The power density at any point in the core must be limited SAFETY ANALYSES to maintain the fuel design criteria (Refs. 4 and 5). This (continued) is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the correlations between measured quantities. the power distribution. and uncertainties in determining the power distribution.

Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate CLHGR) so that the peak cladding temperature does not exceed 2200°F (Ref . 5). Peak cladding temperatures exceeding 2200°F cause severe cladding failure by oxidation due to a Zircaloy Zirconium water reaction.

The LCOs governing the LHR. ASI. CEAs. and RCS ensure that these criteria are met as long as the core is operated within the ASI and F~ limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core . Operation within the limits for these variables ensures that their actual values are within the ranges used in the accident analyses (Ref. 1).

Fuel cladding damage does not occur from conditions outside the limits of these LCOs during normal operation. However.

fuel cladding damage could result if an accident occurs from initial conditions outside the limits of these LCOs . This potential for fuel cladding damage exists because changes in the power distribution can cause increased power peaking and can correspondingly increase local LHR.

The LHR satisfies Criterion 2 of 10 CFR 50.36 (c)(2)(ii).

LCD The power distribution LCD limits are based on correlations between power peaking and certain measured variables used as inputs to the LHR and DNBR operating limits. The power distribution LCD limits are provided in the COLR. The limitation on LHR ensures that in the event of a LOCA the peak temperature of the fuel cladding does not exceed 2200°F .

(continued)

PALO VERDE UNITS 1.2.3 B 3.2.1-4 REVISION 0

LHR B 3.2.1 BASES (continued)

REFERENCES 1. UFSAR. Section 15 .

2. UFSAR, Section 6.

3. CE 1 Correlation for D~BR. UFSAR Section 4.4 .

4. 10 CFR 50. Appendix A. GOC 10.

5. 10 CFR 50.46.

6. Regu l at ory Guide 1.77, Rev. 0. May 1974.

7. 10 CFR 50, Appendix A. GDC 26 .

PALO VERDE UNITS 1.2.3 B 3.2.1-8 REVISION w

F~

B 3.2 .L BASES BACKGROUND which may produce acceptable power distributions. Operation (continued) within the design li mits of power di stribution is accomp l ished by genera ting opera ting limits on Linear Heat Rate CLHR) and Departure from Nucleate Boiling CDNB) .

Proximity to the DNB condition is expressed by the Departure from Nucleate Boiling Ratio CDNBR). defined as the ratio of the cladding surface heat f l ux required to cause DNB to the actual cladding surface heat flux. The minimum DNBR value during both normal operation and AOOs i s the DNBR Safety Limit as calculated by t he bf.--+/-a~plicable DNB Cor relation (Ref . 3) and corrected for suchact ors as rod bow and gri d spacers. and it is accepted as an appropriate margin to DNB for all operating conditions.

There are two systems that monitor core power distributi on online: the Core Operating Limit Supervisory System CCOLSS) and the Core Protection Calculators CCPCs). The COLSS and CPCs th at monitor the core power distribution are capable of verifying that the LHR and the DNBR do not exceed thei r limits. The COLSS performs thi s function by continuously monitoring the core power distribution and calculating core power operati ng limits correspond ing to the allowable peak LHR and DNBR values. The CPCs perform this function by continuously calculating actual values of DNBR and Local Power Density CLPD ) for comparison with the respective trip setpoints.

DNBR penalty factors are included in both the COLSS and CPC DNBR calculations to accommodate the effects of rod bow.

The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly. Fuel assemb lies t hat incur higher than average burnup experi ence greater rod bow. Conversely , fuel assemblies that receive lower than average burnup experience less rod bow. In design calculations for a reload core. each batch of fuel is assigned a penalty applied to the maximum integrated planar radial power peak of the batch. Thi s penalty is correlated with the amount of rod bow determined from the maximum average assembly burnup of the batch. A single net penalty for the COLSS and CPCs is then determined from the penalties associated with each batch th at comprises a core reload.

accounting for the offsetting margins due to the l ower radial power peaks in the higher burnup batches.

The COLSS indicates continuously to the operator how far the core is to the operating limits and provides an audib le (continued)

PALO VERDE UNITS 1.2.3 B 3.2.2-2 REVISI ON ~

F~

B 3.2.L BASES APPLICABLE b. During CEA misoperation events or a loss of flow accident. there must be at least SAFETY ANALYSES 95%probability at the 95% confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref . 4);

c. During an ejected CEA accident. the fission energy input to the fuel must not exceed 280 cal/gm (Ref . 6);

and

d. The control rods (excluding part strength rods) must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref . 7).

The power density at any point in the core must be limited to maintain the fuel design criteria (Refs. 4 and 5). This result is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the correlations between measured quantities. the power distribution. and the uncertainties in the determination of power distribution.

Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate CLHGR) so that the peak cladding temperature does not exceed 2200°F (Ref . 5). Peak cladding temperatures exceeding 2200°F cause severe cladding failure by oxidation due to a Zircaloy Zirconium water reaction.

The LCOs governing LHR. ASI. CEAs. and RCS ensure that these criteria are met as long as the core is operated within the ASI and Fx limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the ranges used in the accident analyses (Ref. 1).

Fuel cladding damage does not occur because of conditions outside the limits of these LCOs for ASI. F~ . and Tq during normal operation . However. fuel cladding damage results if an accident occurs from initial conditions outside the limits of these LCOs. This potential for fuel cladding damage exists because changes in the power distribution can (continued)

PALO VERDE UNITS 1.2 .3 B 3.2 .2-4 REVISION 8-2

F X)'.

B 3.2.L BASES (continued)

SURVEILLANCE SR 3.2 .2.1 REQUIREMENTS This periodic Surveillance is for determining, using the Incore Detector System. that F~ values are~ f~ values used in the COLSS and CPCs. It ensures that the Fx values used remain valid throughout the fuel cycle. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program. Determining the F~y values after each fuel loading when THERMAL POWER is > 40% RTP. but prior to its exceeding 70% RTP, ensures that the core is properly loaded.

REFERENCES 1. UFSAR. Section 15.

2. UFSAR. Section 6.

3. CE 1 Correlation for DNBRUFSAR Section 4.4 .

4. 10 CFR 50, Appendi x A, GDC 10 .

5. 10 CFR 50 .46.

6. Regulatory Guide 1. 77' Rev. 0. May 1974.

7. 10 CFR 50. Appendix A. GDC 26.

PALO VERDE UNITS 1.2.3 B 3.2 .2-7 REVISION ~

T~

B 3.2.3 BASES BACKGROUND Power distribution is a product of multipl e parameters.

(continued) various combinations of which may produce acceptable power distributions . Operation within the design limits of power distribution is accomplished by generating operating limits on the Linear Heat Rate (LHR) and the Departure from Nuc l eate Boi l ing (DNB).

Proximity to the DNB condition is expressed by the Departure from Nucleate Boiling Ratio (DNBR ) . defined as the ratio of the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux. The minimum DNBR value during both normal operation and AOOs is the DNBR Safety Limit as calculated by the bf-+/-a~plicable DNB Correlati on (Ref. 3) and corrected for suchactors as rod bow and grid spacers. and it is accepted as an appropriate margin to DNB for all operating conditions.

There are two systems that monitor core power distribution online: the Core Operating Limit Supervisory System (COLSS) and the Core Protection Calculators (CPCs). The COLSS and CPCs that monitor the core power distribution are capable of verifying that t he LHR and the DNBR do not exceed their limits. The COLSS performs this function by continuously monitoring the core power distribution and calculating core power operating limits corresponding to the allowable peak LHR and DNBR. The CPCs perform this function by continuously calculating actual values of DNBR and Loca l Power Density (LPD) for comparison with the respective trip setpoints .

A DNBR penalty factor is included in the COLSS and CPC DNBR ca l culation t o accommodate the effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by the assembly. Fuel assemblies that incur higher than average burnup experience greater magnitude of rod bow . Conversely, fuel assemblies that receive lower than average burnup experience less rod bow.

In design calculations for a reload core. each batch of fuel is assigned a penalty applied to the maximum integrated planar radial power peak of the batch. This penalty is correlated with the amount of rod bow that is determined from the maximum average assembly burnup of the batch. A single net penalty for the COLSS and CPCs is then determined from the penalties associated wi th each batch that comprises a core reload. accounting for the offsetting margins caused by the lower radial power peaks in the higher burnup batches .

(continued)

PALO VERDE UNITS 1.2.3 B 3.2 .3-2 REVISION +/-Q.

Tg B 3.2.3 BASES APPLICABLE b. During CEA misoperation events or a loss of flow accident. there must be at least SAFETY ANALYSES 95% probability at the 95%confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4);

c. During a CEA ejection accident. the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6);

and

d. The control rods (excluding part strength rods) must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 7).

The power density at any point in the core must be limited to maintain the fuel design criteria (Ref. 1). This result is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analysis (Ref. 2) with due regard for the correlations between measured quantities. the power distribution. and uncertainties in the determination of power distribution.

Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate CLHGR) so that the peak cladding temperature does not exceed 2200°F (Ref . 1). Peak cladding temperatures exceeding 2200°F cause severe cladding failure by oxidation due to a Zircaloy Zirconium water reaction .

The LCOs governing LHR. ASI. CEAs. and RCS ensure that these criteria are met as long as the core is operated within the ASI and Fx limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core . Operation within the limits of these variables ensures that their actual values are within the range used in the accident analyses (Ref. 1).

(continued)

PALO VERDE UNITS 1.2.3 B 3.2.3-4 REVISION ~

Tg B 3.2 .3 BASES REFERENCES 1. UFSAR. Section 15.

2. UFSAR. Section 6.

3. CE 1 Correlation for DNBRUFSAR Section 4.4 .

4. 10 CFR 50. Appendi x A. GDC 10.

5. 10 CFR 50 .46.

6. Regulatory Guide 1.77. Rev . 0. May 1974 .

7. 10 CFR 50. Appendi x A. GDC 26 .

PALO VERDE UNITS 1.2.3 B 3.2.3 -10 REVISION tt

DNBR B 3.2.4 BASES BACKGROUND Power di st ri buti on i s a product of mu l tip l e parameters.

(continued) various combi nations of which may produce acceptable power distributions . Operation with i n the design limits of power distribution i s accomplished by generating operating limits on the Linear Heat Rate (LHR) and the Departure from nucleate boi l i ng (DNB).

Proximity to the DNB condition is expressed by the DNBR .

defined as t he ratio of the cladd i ng surface heat flux required t o cause DNB to the actual cl adding surface heat f l ux. The minimum DNBR va l ue duri ng both normal operat ion and AOOs i s t he DNBR Safety Limit as ca l culated by the Gf-+/-

applicable DNB Correlation (Ref . 3) and corrected for such factors as rod bows and grid spacers and it is accepted as an appropriate margin to DNB for all operating conditions .

There are two systems that monitor core power distribution online : the Core Operating Limits Supervisory System (COLSS) and the Core Protection Calculators (CPCs). The COLSS and CPCs that mon i tor the core power distribution are capabl e of verifyi ng that the LHR and DNBR do not exceed t hei r li mits . The COLSS performs th i s funct i on by contin uous ly mon itoring the core power distribution and calcul ating core power operating l imi ts corresponding to the allowable peak LHR and DNBR. The CPCs perform this function by continuously calculating an actual value of DNBR and LPD for comparison with the respective trip setpoints.

A DNBR penalty factor is included in both the COLSS and CPC DNBR ca l cula t ion to accommodate t he effects of rod bow. The amount of rod bow in each assembly is dependent upon the average burnup experienced by t hat assemb ly. Fuel assemblies that incur higher t han average burnup exper i ence a greater magn i tude of rod bow. Conversely, fuel assembl ies that receive lower than average burnup experience less rod bow. In design calculations for a reload core. each batch of fuel is assigned a penalty that is applied to the maximum integrated planar radial power peak of the batch. This penalty is correlated with the amount of rod bow that is determined from the maximum average assembly burnup of the batch. A si ng le net penalty for the COLSS and CPCs i s then determi ned f rom t he penalti es assoc i ated with each batch that comprises a core reload. accounting for the offse t t ing margins due to t he lower radia l power peaks i n the hi gher burnup batches.

(continued)

PALO VERDE UNITS 1.2.3 B 3.2 .4-2 REVISION +/-.Q.

DNBR B 3.2.4 BASES APPL ICABL E b. During CEA misoperation events or a loss of flow accident, there must be at l east SAFETY ANALYSES 95% probability at the 95% confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 3);

c. During an ejected CEA accident, the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6);

and

d. The control rods (excluding part strength rods) must be capable of shutting down the reactor with a minimum requ i red SOM with the highest worth control rod stuck fully withdrawn (Ref. 7).

The power density at any point in the core must be limited to maintain the fuel design criteria (Ref. 4). This is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the corre l ations between measured quantities, the power distribution, and uncertainties i n the determination of power distribution.

Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate CLHGR) so that the peak cladding temperature does not exceed 2200°F (Ref. 4) . Peak cladding temperatures exceeding 2200°F may cause severe cladding failure by oxidation due to a Zircaloy zirconium water reaction.

The LCOs governing LHR, ASI, CEAs. and RCS ensure that these criteria are met as long as the core is operated within the ASI and Fx limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the range used in the accident analyses (Ref. 1).

(continued)

PALO VERDE UNITS 1,2,3 B 3.2.4-4 REVISION ~

DNBR B 3.2.4 BASES SURVEI LLANCE SR 3.2.4.2 REQUIREMENTS (continued) Verification that the COLSS margin alarm actuates at a power level equal to or less than the core power operating limit.

as calculated by the COLSS. based on the DNBR . ensures that the operator is alerted when operating conditions approach the DNBR operating limit. The Surveillance Frequency is controlled under the Surveillance Frequency Control Program .

REFERENCES 1. UFSAR. Chapter 15.

2. UFSAR. Chapter 6.

3. CE 1 Correlation for DNBR . UFSAR Section 4.4.

4. 10 CFR 50. Appendi x A. GDC 10 .

5. 10 CFR 50 . 46.

6. Regulatory Guide 1.77. Rev. 0. May 1974.

7. 10 CFR 50. Appendix A. GDC 26.

PALO VERDE UNITS 1.2.3 B 3.2 .4-9 REVISION oo

ASI B 3.2.5 BASES BACKGROUND Power distribution is a product of multiple parameters.

(continued) various combinations of which may produce acceptable power distributions . Operation within the design limits of power distribution is accomplished by generating operating limits on the Linear Heat Rate (LHR) and the Departure from Nucleate Boiling (DNB).

Proximity to the DNB condition is expressed by the Departure from Nucleate Boiling Ratio (DNBR), defined as the ratio of the cladding surface heat flux required to cause DNB to the actual cladding surface heat flux . The minimum DNBR value during both normal operation and AOOs is the DNBR Safety Limit as calculated by the bf-+/----a ~plicable DNB Correlation (Ref. 3). and corrected for suchactors as rod bow and grid spacers. and it is accepted as an appropriate margin to DNB for all operating conditions.

There are two systems that monitor core power distribution online: the Core Operating Limit Supervisory System (COLSS) or the Core Protection Calculators (CPCs). The COLSS and CPCs monitor the core power distribution and are capable of verifying that the LHR and DNBR do not exceed their limits.

The COLSS performs this function by continuously monitoring the core power distribution and calculating core power operating limits corresponding to the allowable peak LHR and DNBR. The CPCs perform this function by continuously calculating actual values of DNBR and local power density (LPD) for comparison with the respective trip setpoints.

A DNBR penalty factor is included in both the COLSS and CPC DNBR calculations to accommodate the effects of rod bow.

The amount of rod bow in each assembly is dependent upon the average burnup experienced by that assembly. Fuel assemblies that incur higher than average burnup experience greater rod bow. Conversely, fuel assemblies that receive lower than average burnup experience less rod bow. In design calculations for a reload core. each batch of fuel is assigned a penalty that is applied to the maximum integrated planar radial power peak of the batch . This penalty is correlated with the amount of rod bow that is determined from the maximum average assembly burnup of the batch. A single net penalty for the COLSS and CPC is then determined from the penalties associated with each batch that comprises a core reload. accounting for the offsetting margins due to the lower radial power peaks in the higher burnup batches.

(continued)

PALO VERDE UNITS 1.2.3 B 3.2.5-2 REVISION +/--0

ASI B 3.2.5 BASES APPLICABLE b. During CEA misoperation events or a loss of flow accident. there must be at least SAFETY ANALYSES 95% probability at the 95% confidence level (the (continued) 95/95 DNB criterion) that the hot fuel rod in the core does not experience a DNB condition (Ref. 4):

c. During an ejected CEA accident. the fission energy input to the fuel must not exceed 280 cal/gm (Ref. 6):

d. The control rods (excluding part strength rods) must be capable of shutting down the reactor with a minimum required SOM with the highest worth control rod stuck fully withdrawn (Ref. 7) .

The power density at any point in the core must be limited to maintain the fuel design criteria (Refs. 4 and 5). This is accomplished by maintaining the power distribution and reactor coolant conditions so that the peak LHR and DNB parameters are within operating limits supported by the accident analyses (Ref. 1) with due regard for the correlations among measured quantities. the power distribution. and uncertainties in the determination of power distribution.

Fuel cladding failure during a LOCA is limited by restricting the maximum Linear Heat Generation Rate CLHGR) so that the peak cladding temperature does not exceed 2200°F (Ref . 5). Peak cladding temperatures exceeding 2200°F may cause severe cladding failure by oxidation due to a Zircaloy Zirconium water reaction.

The LCOs governing LHR. ASI, and RCS ensure that these criteria are met as long as the core is operated within the ASI and Fx limits specified in the COLR. and within the Tq limits. The latter are process variables that characterize the three dimensional power distribution of the reactor core. Operation within the limits for these variables ensures that their actual values are within the range used in the accident analysis (Ref. 1).

Fuel cladding damage does not occur from conditions outside these LCOs during normal operation. However, fuel cladding damage results when an accident occurs due to initial conditions outside the limits of these LCOs. This potential for fuel cladding damage exists because changes in the power distribution can cause increased power peaking and correspondingly increased local LHRs.

(continued)

PALO VERDE UNITS 1,2.3 B 3.2.5-4 REVISION 6,'.?

ASI B 3.2 .5 BASES SURVEILLANCE SR 3.2.5.1 (continued)

REQUIREMENTS (continued) This SR is modified by a Note that states that the SR is not requ i red to be performed until 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after MODE 1 with THERMAL POWER > 20%RTP. During plant startup (increase from 15-18% RTP). the plant dynamics associated with the downcomer to economizer swapover may result in a temporary power increase above 20%RTP . The 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> after reaching 20%RTP is required for plant stabilization.

REFERENCES 1. UFSAR. Chapter 15.

2. UFSAR. Chapter 6.

3. CE 1 Correlation for DNBR. UFSAR Section 4.4 .

4. 10 CFR 50. Appendi x A. GDC 10.

5. 10 CFR 50 .46.

6. Regulatory Guide 1. 77. Rev . 0. May 1974.

7. 10 CFR 50. Appendix A. GDC 26.

PALO VERDE UNITS 1.2.3 B 3.2 .5-7 REV ISION ij

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 5 Assessment of Topical Report Limitations and Conditions

Enclosure Description and Assessment of Proposed License Amendment Assessment of Topical Report Limitations and Conditions Technical Specification (TS) 5.6.5, Section "b," Core Operating Limits Report (COLR), lists the documents which describe the Core Operating Limits Report (COLR) analytical methods used to determine the core operating limits presented in each PVNGS unit specific COLR.

The proposed changes to TS 5.6.5.b will add the following topical reports to the list of referenced core operating analytical methods to be consistent with the analytical methods that will be used to determine the core operating limits following NGF, VIPRE-W Code, Critical Heat Flux (CHF) Correlation, and Zirconium Diboride burnable absorber methodology implementation:

• WCAP-16500~P-A, CE 16x16 Next Generation Fuel Core Reference Report

• WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis

• CENPD-387-P-A, ABB Critical Heat Flux Correlations for PWR Fuel

• WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes, August 2007

• WCAP-16072-P-A, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs Sections 1 through 5 of this Attachment provide an assessment as to how the Limitations and Conditions contained in the NRC Safety Evaluations for these topical reports are met.

Topical report CENPD-404-P-A is currently addressed in TS 5.6.5.b as Item 13. Section 6 of this attachment addresses how CENPD-404-P-A, Addendum 1-A, and Addendum 2-A, will be implemented with the introduction of NGF.

Topical report CENPD-183-A is currently addressed in TS 5.6.5.b as Item 19. Section 7 of this attachment addresses how topical report CENPD-183-A will be implemented differently with the introduction of NGF.

1. WCAP-16500-P-A (Next Generation Fuel) 1.1. WCAP-16500-P-A, Revision 0 Topical report WCAP-16500-P-A, CE 16x16 Next Generation Fuel Core Reference Report, describes the 16x16 lattice NGF assembly mechanical design for the Combustion Engineering (CE) nuclear steam supply system (NSSS) and the methods and models used for evaluating its acceptability. In addition to the reference product. description, this topical report describes the fuel mechanical and reload design methodology intended to support fuel design and licensing applications up to a rod average burnup of 62 Gigawatt Days per Metric Ton Uranium (GWd/MTU). .

The limitations and conditions contained in the NRC Safety Evaluation (SE) for this topical report are met as follows:

Condition 1:

Using approved methods, the licensee must ensure that all of the design criteria specified in TR WCAP-16500,.p are satisfied on a cycle-specific basis (SE Section 3.3.1).

ATIACHMENT 5, Page 1

Enclosure Description and Assessment of Proposed License Amendment Condition 1 Response:

Fuel temperature and rod internal pressure criteria are confirmed on a reload basis as part of the approved reload process Condition 2:

Fuel assembly component design and configuration (e.g., type and distribution of spacer grids and /FM grids) are limited to the five designs described in TR WCAP-16500-P and in response to RA/ 2 (SE Section 3.2).

Condition 2 Response:

The NGF grid as described in WCAP-16500-P-A, WCAP-16500-P-A, Revision 0 (Reference 1.10), and WCAP-16500-P-A, Supplement 2 (Reference 1. 7) complies with the condition described. -

Condition 3:

The reference fuel assembly design, CE 16x16 NGF, its fuel mechanical design methodology and design criteria, are approved up to a peak rod average bumup of 62 GWd/MTU. A fuel burnup limit may exist, however, either explicitly or implicitly, in other portions of a plant's licensing basis. The NRG staff's approval of this topical report allows the CE 16x16 NGF assembly to reach a rod average bumup of 62 GWd/MTU. However, a licensing amendment request, specifically addressing each plant's licensing basis including radiological consequences, is required prior to extending bumup beyond current levels. Further, the NRG staff's SE for Optimized ZIRLO' (Addendum 1 to TR WCAP-12610-P-A and TR CENPD-404-P-A) specified a 60 MWdlkgU bumup limit and this limitation must be revised prior to extending the peak rod average bumup for the NGF design (SE Section 3.4).

Condition 3 Response:

Calculations are performed up to 62 GWd/MTU, although the rod average burnup limit remains at 60 GWd/MTU (licensed) for NGF.

Condition 4:

Licensees shall demonstrate the accuracy of their growth predictions based upon measured data and this validation shall be ahead of the bumups achieved by batch implementation. The growth model validation (e.g., measured versus predicted) should be documented in a letter(s) to the NRG (SE Section 3.2.1).

Condition 4 Response:

Westinghouse demonstrated the accuracy of growth predictions based upon measured data and provided this growth model validation to the NRC in Westinghouse document LTR-NRC 40 (Reference 1.8).

Condition 5:

To compensate for NRC staff concerns related to the digital setpoints process, an interim margin penalty of 6% must be applied to the final addressable constants (e.g., BERR1*1.06,

[(1+EPOL2)*1.06-1.0]) calculated following the 1164 hypercube setpoints process.... Removal of this interim margin penalty will be considered after the digital setpoints methods have been formalized, documented (e.g., revision to TR WCAP-16500-P), and approved by the NRG (SE Section 3. 7).

ATTACHMENT 5, Page 2

Enclosure Description and Assessment of Proposed License Amendment Condition 5 Response:

The digital setpoints methods utilized for NGF introduction at PVNGS Units 1, 2 and 3 have been formalized and approved by the NRC in WCAP-16500-P-A, Supplement 1, Revision 1 (Reference 1.9). Therefore, a 6% interim margin penalty is not applied to the final addressable constants.

Condition 6:

Licensees are required to demonstrate that during transition cores, DNB margin gains associated with the NGF design offset (1) any impacts to flow starvation due to the increased pressure drop and (2) uncertainty associated with predicting local flow characteristics. Further, licensees must detail the analytical methods and results oftheir transition core LOCA and non-LOCA analyses (SE Sections 3.7 and 3.10).

Condition 6 Response:

NGF assemblies will likely experience some flow starvation due to flow diversion in the mixing vane grid region in transition cores. Thermal margin gains from the WSSV CHF correlation will offset any flow penalty in that region. The thermal margin gains associated with the NGF design are realized through the CETOPNIPRE-W benchmarking analysis performed in accordance with the methodology described in WGAP-16500-P-A, Revision 0 (Reference 1.10).

The analytical methods used to analyze transition cores are discussed in Attachments 7 and 8 to the enclosure, Sections 5, 7, and 8.

Condition 7:

Implementation of CE 16x16 NGF assemblies necessitates re-analysis of the plant-specific LOCA analyses. Licensees are required to submit a license amendment containing the revised LOCA analyses for NRG review. Upon approval, the revised LOCA analyses constitute the analysis-of-record and baseline for which future changes will be measured in accordance with 10 CFR 50.46(a)(3) (SE Sections 3.7 and 3.10).

Condition 7 Response:

. Revised LOCA analyses are provided for NRC review in Attachments 7 and 8 to the enclosure, Sections 8 and 9.

Condition 8: .

Using approved models and methods, Westinghouse will continue to limit peak local power experienced during Condition I and II events to ensure that fuel temperature remains below melting temperature at all bumups. This evaluation may be both plant and cycle-specific (SE Section 3.3.4).

Condition 8 Response:

Calculations have been performed to ensure that fuel melt does not occur below the steady-state and transient peak local heat generation rate. Additionally, the limiting peak local power was evaluated to ensure the fuel melt acceptance criterion is met for transient events. The peak power limits are to be verified to be applicable for the reload cycle per the approved reload process.

Condition 9:

The NRG staff's approval of TR WCAP-16500-P establishes the licensing basis for batch implementation of the CE 16x16 NGF assembly design. Licensees wishing to implement this ATTACHMENT 5, Page 3

Enclosure Description and Assessment of Proposed License Amendment fuel design are required to submit a license amendment request, where applicable, updating their Core Operating Limits Report list of methodologies with the A" version of this TR.

Condition 9 Response:

A listing of new methodologies to be included in Technical Specification 5.6.5, Section "b," Core Operating Limits Report (COLR), is provided in the enclosure Section 3.2.

Condition 1O:

The NRG staffs review did not include the LOCA model changes describe in Appendix A of TR WCAP-16500-P. Therefore, a licensee, will have to submit a license amendment, if they desire to use the Appendix A LOCA model changes.

Condition 10 Response:

Changes to the LOCA model outlined in Appendix A of WCAP-16500-P-A, Revision 0 (Reference 1.10) were resubmitted to the NRC by Westinghouse under CENPD-132, Supplement 4-P-A, Addendum 1-P (Reference 1.12) and have been approved for use in license amendment applications as described in WCAP-16523-P-A (Reference 1.11).

1.2. WCAP-16500-P-A, Supplement 1, Revision 1 Topical report WCAP-16500-P-A, Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF), describes a revised analytical process for calculating COLSS and CPCS addressable constants and database constants for plant reloads with CE 16x16 NGF (CE16NGF') assemblies .

. There are no limitations and conditions contained in the NRC Safety Evaluation for this topical 1

report.

1.3. WCAP-16500-P-A, Supplement 2 Topical report WCAP-16500-P-A, Supplement 2, Evolutionary Design Changes to CE 16x16 Next Generation Fuel and Method for Addressing the Effects of End-of-Life Properties on Seismic and Loss of Coolant Accident Analyses, describes three evolutionary changes to the CE 16x16 NGF spacer grid designs being pursued that, in combination, will further improve the grid-to-rod fretting (GTRF) resistance of the fuel and its resistance to crud formation while improving the fabricability of the spacer grid straps. The ch8'iges to the design are all related to the mid and intermediate flow-mixing (IFM) spacer grids. All other features of the CE 16x16 NGF design remain the same as described in WCAP-16500-P-A.

The limitations and conditions contained in the NRC Safety Evaluation for this topical report are met as follows:

Condition 1:

This condition contains proprietary information and is therefore not restated in this attachment.

It is available in the NRG staff safety evaluation, which is included in the approved topical report.

Condition 1 Response:

All CE 16x16 NGF spacer grids manufactured for PVNGS Units 1, 2, and 3 were manufactured in compliance with the requirements of this condition.

ATTACHMENT 5, Page 4

Enclosure Description and Assessment of Proposed License Amendment Condition 2:

Any changes or combinations of changes approved in this safety evaluation shall be analyzed and explicitly accounted for according to approved licensed methodologies prior to implementation.

Condition 2 Response:

The combined mechanical design changes described in WCAP-16500-P-A, Supplement 2 (Reference 1. 7), have been explicitly accounted for in the analyses performed to support implementation of the CE 16x16 NGF product.

Condition 3:

Licensees may not reference the proposed approach to address IN 2012-09 detailed in the Supplemental 2 to WCAP-16500-P-A I WCAP-16500-NP-A submittal as this approach has not been reviewed or approved by the NRC staff.

. Condition 3 Response:

No credit was taken for the approach to address IN 2012-09 outlined in WCAP-16500-P-A, Supplement 2 (Reference 1.7), and a licensing condition was established to ensure IN 2012-09 is addressed in the future following an NRG-approved approach.

2. WCAP-14565-P-A (VIPRE-W Code}

2.1. WCAP-14565-P-A, Revision 0 VIPRE-01 is a subchannel thermal-hydraulic (T-H) computer code that is typically used to describe the reactor core of a nuclear power plant. The VIPRE-01 code was developed by the Battelle Northwest National Laboratories under the sponsorship of Electric Power Research Institute (EPRI) (Reference 1.3). VIPRE-01 has been approved by the USNRC for PWR licensing applications (References 1.4 and 1.5).

Topical report WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis (Reference 1.6), describes Westinghouse VIPRE (hereafter referred to as VIPRE-W) modeling and qualification for PWR T-H design and non-LOCA safety analysis.

The limitations and conditions contained in the NRC Safety Evaluation for this topical report are met as follows:

Condition 1:

Selection of the appropriate CHF correlation, DNBR safety limit, engineered hot channel factors for enthalpy rise and other fuel-dependent parameters for a specific plant application should be justified with each submittal.

Condition 1 Response:

The WSSV critical heat flux (CHF) correlation with a 95/95 correlation DNBR safety limit of 1.12 approved with VIPRE-W in WCAP-16523-P-A (Reference 1.11) was used in the DNBR calculations for the mixing vane grid regions of the NGF design.

The ABB-NV CHF correlation with a 95/95 correlation limit of 1.13 approved with VIPRE-W in WCAP-14565-P-A Addendum 1-A (Reference 1.13) and WCAP-14565-P-A Addendum 2-P-A (Reference 1.14) was used in the DNBR calculations for the non-mixing vane grid regions for ATTACHMENT 5, Page 5

Enclosure Description and Assessment of Proposed License Amendment the NGF design, as well as .for the Standard Fuel (STD), also known as Value-Added Fuel (VAF), design in place of the currently licensed CE-1 CHF correlation.

The WLOP CHF correlation can be used as an alternative to the MacBeth CHF correlation for the analysis of both STD and NGF in PVNGS when the primary CHF correlation is not applicable. In WCAP-14565-P-A Addendum 2-P-A, the WLOP 95/95 correlation limit o'f 1.18 was approved for use with VI PRE-W.

The correlation limits used in the NGF safety analysis DNBR calculations for the loaded fuel types in PVNGS are consistent with the approved values in WCAP-16523-P-A for the WSSV CHF correlation and WCAP-14565-P-A Addendum 2-P-A for the ABB-NV and WLOP CHF correlations. The engineered hot channel factors and other fuel-dependent parameters in the PVNGS DNBR calculations for the STD and NGF designs were justified as part of the statistical combination of uncertain,ties (SCU) DNBR safety limit calculations.

Condition 2:

Reactor core boundary conditions determined using other computer codes are generally input into VIPRE for reactor transient analyses. These inputs include core inlet coolant flow and enthalpy, core average power, power shape and nuclear peaking factors. These inputs should be justified as conservative for each use of V/PRE.

Condition 2 Response:

The core boundary conditions used in the VIPRE-W DNBR calculations are all generated from NRG-approved codes and analysis methodologies. Continued applicability of the core boundary conditions as VIPRE-W input is verified on a cycle-by-cycle basis using the CE NSSS reload methodology.

Condition 3:

The NRG Staff's generic SER for V/PRE set requirements for use of new CHF correlations with VIPRE. Westinghouse has met these requirements for using the WRB-1, WRB-2 and WRB-2M correlations. The DNBR safety limit for WRB-1 and WRB-2 is 1.17. The WRB-2M correlation has a DNBR safety limit of 1.14. Use of other CHF correlations not currently included in VIPRE will require additional justification.

Condition 3 Response:

As discussed in response to Condition 1, the WSSV CHF correlation with a 95/95 correlation limit of 1.12 was approved in WCAP-16523-P-A (Reference 1.11). The ABB-NV CHF correlation with a 95/95 correlation limit of 1.13 and the WLOP CHF correlation with a 95/95 correlation limit of 1.18 were previously approved in WCAP-14565-P-A Addendum 2-P-A (Reference 1.14) for use with the VIPRE-W code.

Condition 4:

Westinghouse proposes to use the VIPRE code to evaluate fuel performance following postulated design-basis accidents, including beyond-CHF heat transfer conditions. These evaluations are necessary to evaluate the extent of core damage and to ensure that the core maintains a coo/able geometry in the evaluation of certain accident scenarios. The NRG Staff's generic review of VIPRE did not extend to post CHF calculations. V/PRE does not model the time-dependent physical changes that may occur within the fuel rods at elevated temperatures.

Westinghouse proposes to use conservative input in order to account for these effects. The NRG Staff requires that appropriate justification be submitted with each usage of VIPRE in the post-CHF region to ensure that conservative results are obtained.

ATTACHMENT 5, Page 6

Enclosure Description and Assessment of Proposed License Amendment Condition 4 Response:

For the PVNGS NGF safety analysis, application of the VIPRE-W code does not model the time-dependent physical changes that may occur within fuel rods at elevated temperatures in the post-CHF region.

2.2. WCAP-14565-P-A, Addendum 1-A Topical report WCAP-14565-P-A, Addendum 1-A, Addendum 1 to WCAP-14565-P-A Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code, has installed ABB-NV and ABB-TV CHF correlations into its version of the VIPRE (i.e., VIPRE-W) code. The ABB-NV correlation is for CE-PWR 14x14 and 16x16 fuels with non-mixing vane grids. The ABB-TV correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

The limitations and conditions contained in the NRC Safety Evaluation for this topical .report, including Safety Evaluation Section 3.1, are met as follows:

Section 3.1 :

Westinghouse will apply the VIPRE-01 code with the ABB CHF correlations under the following.

conditions consistent with the requirements in the CENPD-387-P safety evaluation:

1. The 95195 DNBR limits of the ABB-NV and ABB-TV correlations are not lower than the current NRG-approved limit of 1.13 for the CE-PWR fuels.

2. The ABB-NV and ABB-TV correlations are used with the VIPRE-01 code, in addition to the TORC and CETOP-0 codes currently used for CE-PWRs. This addendum demonstrates the V/PRE-01 equivalency to TORC for DNBR calculations.

3. The ABB-NV and ABB-TV correlations are used with the optimized Fe shape factor to account for effects of non-uniform axial power shapes.

4. The current range of applicability for the ABB-NV and ABB-TV correlations as shown in Table 2-1 of the TR remains applicable.

5. The ABB-NV and ABB-TV correlations are used only for the CE-PWR fuel designs with NRG-approved methodology for PWR safety analysis.

6. Technology transfer is accomplished through a process that meets the guidance of GL 83-11, Supplement 1.

Section 3.1 Response:

The VIPRE modelling used for this analysis meets these conditions. Detailed responses to the specific CENPD-387-P-A safety evaluation conditions are addressed in Section 3 of this Attachment.

Condition 1:

Addendum 1 to the WCAP-14565-P-A VIPRE-01 model must remain consistent with that for the DNB data analysis described in WCAP-14565-P-A V/PRE-01.

Condition 1 Response:

The VIPRE-W model used for this analysis was performed as described in WCAP-14565-P-A.

Condition 2:

The current 95195 DNBR safety limit of 1.13 remains unchanged.

ATTACHMENT 5, Page 7

Enclosure Description and Assessment of Proposed License Amendment Condition 2 Response:

The ABB-NV 95/95 DNBR safety limit of 1.13 remains unchanged for the thermal hydraulic analysis performed for this application.

Condition 3:

DNBR calculations for CE-PWR fuels are within the current applicable range defined in Table 2-1 of the TR.

• Table 2-1: Applicable Range of ABB CHF Correlations

[Note: Table 2-1 is taken verbatim from WCAP-14565-P-A, Addendum 1-A]

Parameter ABB-NV Range ABB-TV Range Pressure (psia) 1750 to 2415 1500 to 2415 Local mass velocity (Mlbmlhr-tf) 0.8 to 3.16 0.9 to 3.40 Local quality (fraction) -0.14 to 0.22 -0.10 to 0.225 Heated length, inlet to CHF location (in) 48 to 150 48 to 136. 7 Grid spacing (in) 8 to 18.86 8 to 18.86 Heated hydraulic diameter ratio (!!.hml!!.h) 0.679 to 1.08 0.679 to 1.00 Condition 3 Response:

The ABB-TV CHF correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS. The DNBR calculations using the ABB-NV CHF correlation were performed in accordance with the limitations and conditions described in the Safety Evaluation to WCAP-14565-P-A Addendum 2-P-A Condition 1, which provides updated parameters for the same correlation. Section 2.3 of this Attachment addresses how WCAP-14565-P-A Addendum 2-P-A Condition 1 is met.

• 2.3. WCAP-14565-P-A, Addendum 2-P-A Topical report WCAP-14565-P-A, Addendum 2-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, presents a modification to the ABB-NV correlation based on rod bundle data at low pressure and low flow conditions. The modified low pressure ABB-NV correlation is designated as the Westinghouse Low Pressure (WLOP) correlation in thisreport. The WLOP correlation predicts more accurate DNBR than either W-3 or MacBeth correlation at the low pressure and low flow conditions.

The limitations and conditions contained in the NRC Safety Evaluation for this topical report are met as follows:

.Condition 1:

The applicable range of the ABB-NV and WLOPcorrelations are presented in Table 1 and Table 2, respectively, of this SE.

ATTACHMENT 5, Page 8

Enclosure Description and Assessment of Proposed License Amendment Table 1: Applicable Range of the ABB-NV Correlation

[Note: Table 2-1 is taken verbatim from WCAP-14565-P-A, Addendum 2-P-A]

Parameter ABB-NV Pressure (psia) 1750 to 2415 Local mass velocity (Mlbmlhr-ff) 0.8 to 3.16 Local quality < 0.22 Heated length, inlet to CHF location (in.) 48* to 150 Heated hydraulic diameter ratio 0.679 to 1.08 Grid distance (in.) . 7.3 to 24

• Although the heated length below the first MV grid is below 48 in., the minimum heated length used in the correlation is conservatively maintained at 48 in.

Table 2: Applicable Range of the WLOP Correlation

[Note: Table 2-1 is taken verbatim from WCAP-14:565-P-A, Addendum 2-P-A]

Parameter WLOP Pressure (psia) 185 to 1800 Local Coolant Quality < 0.75 Local mass velocity (Mlbmlhr-ff) 0.23 to 3.07 Heated hydraulic diameter ratio 0.679 to 1.00 Heated length, HL (in.) 48* to 168 Grid spacing term 27to 115

• Set as minimum HL value, applied at all elevations below 48 inches.

Condition 1 Response:

For the NGF safety analysis DNB analyses that were based on the ABB-NV and WLOP CHF correlations, the results were confirmed to be in full compliance with the parameter ranges of the CHF correlations as specified in Table 1 and Table 2, respectively, of WCAP-14565-P-A Addendum 2-P-A (Reference 1.14).

Condition 2:

The ABB-NV correlation and the WLOP correlation must use the same Fe factor for power shape correction as used in the primary DNB correlation for a specific fuel design.

Condition 2 Response:

For the NGF safety analysis DNB analyses that were based on the ABB-NV and WLOP correlations, the Fe factor for power shape correction that was applied was the same as the power shape correction used for the WSSV correlation, which is the primary CHF correlation for ATTACHMENT 5, Page 9

Enclosure Description and Assessment of Proposed License Amendment the NGF design at PVNGS. The same optimized Fe factor was applied to the ABB-NV CHF correlation for STD fuel applications.

Condition 3:

Selection of the appropriate DNB correlation, DNBR safety limit, engineering hot channel factors for enthalpy rise, and other fuel-dependent parameters will be justified for each application of each correlation on a plant specific basis.

Condition 3 Response:

The ABB-NV CHF correlation was used for analysis of the PVNGS NGF design when the primary CHF correlation WSSV is not applicable. The WLOP CHF correlation could be used for low pressure events when the primary correlation conditions are outside of the primary correlation limits. In WCAP-14565-P-A Addendum 2-P-A (Reference 1.14), the current ABB-NV and WLOP DNBR safety limits were approved for use with VIPRE-W. The 95/95 ABB-NV CHF correlation limit is 1.13 for PWR fuel design applications. The 95/95 WLOP CHF correlation limit is 1.18. The correlation limits used in the NGF safety analysis DNBR calculations for the loaded fuel types in PVNGS are consistent with the approved values WCAP-16523-P-A (Reference 1.11) for the WSSV CHF correlation and WCAP-14565-P-A Addendum 2-P-A for the ABB-NV and WLOP CHF correlations. The engineering hot channel factors and other fuel-dependent parameters in the PVNGS DNBR calculations for the STD and NGF designs were justified as part of SCU DNBR safety limit calculations.

,Condition 4:

The ABB-NV correlation for Westinghouse PWR applications and the WLOP correlation must be used in conjunction with the Westinghouse version of the VIPRE-01 (VIPRE) code since the correlations were justified and developed based on VIPRE and the associated VIPRE modeling specifications.

Condition 4 Response:

The Westinghouse version of the VIPRE-01 (i.e., VIPRE-W) code subchannel analysis code, which has been qualified and approved with the ABB-NV and WLOP CHF correlations, was implemented for all DNB analyses of the PVNGS fuel types. See Section 2 of this Attachment for compliance with Safety Evaluation conditions on the use of the VIPRE-W code.

3. CENPD-387-P-A (ABB CHF Correlations)

Topical Report CENPD-387-P-A, ABB Critical Heat Flux Correlations for PWR Fuel, provides a description of the PWR CHF correlations for ABB Combustion Engineering PWR 14x14 and 16x16 fuels. The ABB-NV correlation is for ABB-CE PWR 14x14 and 16x16 fuels with non-mixing vane grids. The ABB-TV correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

The limitations and conditions contained in the NRC Sa'fety Evaluation for this topical report are met as follows:

Condition 1:

The ABB-NV and ABB-TV correlations indicate a minimum DNBR limit of 1. 13 will provide a 95 percent probability with 95 percent confidence of not experiencing CHF on a rod showing the limiting value.

ATTACHMENT 5, Page 10

Enclosure Description and Assessment pf Proposed License Amendment Condition 1 Response:

The current NRG-approved ABB-NV 95/95 DNBR safety limit of 1.13 remains unchanged for the thermal hydraulic analysis performed for this application. The ABB-TV CHF correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

Condition 2:

The ABB-NV and ABB-TV correlations must be used in conjunction with the TORC code since the correlations were developed on the basis of the TORC and the associated TORC input specifications. The correlations may also be used in the CETOP-D code in support of reload design calculations.

Condition 2 Response:

The ABB-NV CHF correlation will be used with the TORC and CETOP-D codes in support of reload design calculations. The use of the ABB-NV CHF correlation in conjunction with the VIPRE-01 code in support of reload design calculations has been addressed in Section 2.2 of this attachment. The ABB-TV CHF correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

Condition 3:

The ABB-NV and ABB-TV correlations must also be used with the ABB-CE optimized Fe, shape factor to correct for non-uniform axial power shapes Condition 3 Response:

For the DNB safety analyses using the ABB-NV CHF correlation, the optimized Tong Fe factor for power shape correction was applied to the non-mixing vane grid region DNBR predictions.

The ABB-TV CHF correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

Condition 4:

Range of applicability for the ABB-NV and ABB-TV correlations:

Parameter ABB-NV Range ABB-TV Range Pressure (psia) 1750 to 2415 1500 to 2415 Local mass velocity (Mlbm/hr-ff) 0.8 to 3.16 0.9 to 3.40 Local quality (fraction) -0.14 to 0.22 -0.10 to 0.225 Heated length, inlet to CHF location (in) 48 to 150 48 to 136. 7 Grid spacing (in) 8 to 18.86 8 to 18.86 Heated hydraulic diameter ratio (t:..hm/t:..h) 0.679 to 1.08 0.679 to 1.00 Condition 4 Response:

The results of DNBR calculations using the ABB-NV CHF correlation were confirmed to be in full compliance with the parameter ranges of the CHF correlations as specified in the preceding table when ABB-NV CHF correlation is used with the TORC code. The DNBR calculations using the ABB-NV CHF correlation with the VIPRE-W code were performed in accordance with the limitations and conditions described in the Safety Evaluation to WCAP-14565-P-A Addendum 2-P-A Condition 1, which provides updated parameters for the same correlation.

ATTACHMENT 5, Page 11

Enclosure Description and Assessment of Proposed License Amendment Section 2.3 of this Attachment addresses how WCAP-14565-P-A Addendum 2-P-A Condition 1 is met. The ABB-TV CHF correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

Condition 5:

The ABB-NV and ABB-TV correlation will be implemented in the reload analysis in the exact manner described in Section 7.1 of Topical Report CENPD-387-P, Revision 00-P.

Condition 5 Response:

The ABB-NV correlation is applied according to Section 7.1 of CENPD-387-P-A for non-mixing vane grid spans for CE 16 x 16 Standard and NGF assemblies. The WSSV and WSSV-T correlations are applied for the mixing vane grid spans of the NGF fuel as described in Sections 6.1and6.2 ofWCAP-16523-P-A, respectively, instead of the ABB-TV correlation. The ABB-TV CHF correlation is for the 14x14 Turbo fuel with mixing vane grids, and is therefore not being used at PVNGS.

Condition 6:

Technology transfer Will be accomplished only through the process described in Reference 5

[Letter from Ivan Rickard, ABB-CE, to NRG Document Control Desk, dated February 23, 2000]

which includes ABB-CE performing an independent benchmarking calculation for comparison to the licensee generated results to verify that the new CHF correlations are properly applied for the first application by the licensee.

Condition 6 Response:

APS has successfully participated in the CE Reload Technology Transfer Program, including independent reload core design and verification calculations, as addressed in the NRC approved PVNGS Reload Analysis Methodology Report (Reference 1_.18).

4. WCAP-16523-P-A (WSSV CHF Correlation)

Topical Report WCAP-16523-P, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes, describes the development of critical heat flux (CHF) correlations for pressurized water reactor (PWR) fuel designs containing structural mixing vane grids and intermediate flow mixer grids with side-supported vanes. The correlations, WSSV and WS$V-T, are for 14x14 and 16x16 fuel designs

\ containing side-supported vane grids for CE-PWRs. Both correlations utilize the same form, but with different coefficients. The WSSV correlation coefficients were derived with the Westinghouse version of the VIPRE-01 (i.e., VIPRE-W) subchannel code. The WSSV-T correlation coefficients were derived with the CE TORC subchannel code.

The limitations and conditions contained in the NRC. Safety Evaluation for this topical report are met as follows:

Condition 1:

The WSSV correlation must be used in conjunction with the VIPRE code since the correlation was developed based on VIPRE and the associated VIPRE input specifications. Other uses of the WSSV correlation should reference this TR and be based on appropriate benchmarking with VIPRE.

ATTACHMENT 5, Page 12

Enclosure Description and Assessment of Proposed License Amendment Condition 1 Response:

The WSSV CHF correlation with a 95/95 correlation limit of 1.12 approved in WCAP-16523-P-A (Reference 1.11) was used in the VIPRE-W DNBR calculations for the mixing vane grid regions of the NGF design. The WSSV correlation was installed into CETOP-D code and was benchmarked against the VIPRE-W code for NGF safety analysis applications.

Condition 2:

The WSSV-T correlation must be used in conjunction with the TORC code since the correlation was developed based on TORC and the associated TORC input specifications. The correlations may also be used in the CETOP-0 code in support of reload design calculations benchmarked by TORC.

Condition 2 Response:

The WSSV-T CHF correlation will only be used with the TORC and CETOP-D codes in support of reload design calculations.

Condition 3:

The WSSV and WSSV- T correlations must also be used with the optimized Tong Fe shape factor for non-mixing vane and side-supported mixing vane grids to correct for non-uniform axial power shapes.

Condition 3 Response:

For the safety analysis DNB analyses using the WSSV CHF correlation, the optimized Tong Fe factor for power shape correction was applied to the mixing vane grid region DNBR predictions.

The optimized Tong Fe factor for power shape correction was also applied in conjunction with ABB-NV CHF correlation for non-mixing vane grid region DNBR predictions.

• Condition 4:

The range of applicability for, both WSSV and WSSV-T correlations are:

Parameter Units Range Pressure psia 1,495 to 2,450 Local coolant quality --  :::; 0.34 Local mass velocity 106 lbm/hr-tt.2 0.90 to 3.46 Matrix heated hydraulic diameter, Ohm inches 0.4635 to 0.5334 Heated hydraulic. diameter ratio, Ohm/Oh -- 0.679 to 1.00 Heated length, HL inches 48. to 150 Grid spacing inches 10.28 to 18.86

• Set as minimum HL value, applied at all elevations below 48 inches Condition 4 Response:

For the safety analysis DNB analyses using the WSSV CHF correlation, the results were confirmed to be in compliance with the NRG-approved parameter ranges of the CHF correlation as specified in Condition 4. Any future DNB analyses using the WSSV-T CHF correlation will ATTACHMENT 5, Page 13

Enclosure Description and Assessment of Proposed License Amendment also confirm compliance with the NRG-approved parameter ranges of the CHF correlation as specified in Condition 4.

5. WCAP-16072-P-A (IFBA)

Topical Report WCAP-16072-P-A,lmplementation of Zirconium Diboride Bumable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, describes the use of Zirconium Diboride (Zr8 2 ) Integral Fuel Burnable Absorber (IFBA) in Westinghouse fuel assembly designs.

The limitations and conditions contained in the NRC Safety Evaluation for this topical report are met as follows:

Condition 1:

A license amendment is required to add this TR to the Core Operating Limits Report analytical methods listed in the licensee's TS.

Condition 1 Response:

A listing of new methodologies including WCAP-16072-P-A (Reference 1.16) to be included in Technical Specification 5.6.5, Section "b," Core Operating Limits Report (COLR), is provided in the enclosure Section 3.2. Additionally, Attachment 2 to this enclosure contains the marked-up TS pages for TS Section 5.6.5.b, which lists the topical reports added to the licensing basis as a result of implementing NGF.

Condition 2:

Plant-specific core design guidelines o_r cycle-specific calculations shall be used to verify that required power margins in the axial cutback regions are* maintained within safety analysis limitations.

Condition 2 Response:

APS will ensure that the required power margins in the axial cutback regions are maintained on a cycle-specific basis per the approved reload process.

Condition 3:

Plant TS SRs on MTG validate the physics predictions and ensure that plant operations remain within allowable limits. In addition to current SRs, licensees shall confirm that the peak positive HFP MTG is within the TS limits at the highest RCS soluble boron concentration predicted during full power operation. The peak positive H!=P MTG shall be derived by adjusting the measured MTG at HFP BOC conditions to the maximum HFP soluble boron concentration expected during the cycle. In order to ensure a conservative adjustment, a direct measurement of MTG is required at the highest RCS soluble boron concentration predicted during full power operation. This direct measurement is only required for the first application of ZrB2 IFBA in a CE 14x14or16x16 fuel assembly design. During the first cycle implementation, Westinghouse shall provide the staff with a letter containing the following information:

• i. Measured HFP BOC MTG (TS SR) ii. Measured HFP MTG at highest RCS soluble boron concentration iii. Calculated HFP MTG at highest RCS soluble boron concentration iv. Demonstrated accuracy of the calculated HFP MTG within current analytical uncertainties.

ATTACHMENT 5, Page 14

Enclosure Description and Assessment of Proposed License Amendment In addition, plant procedures used to perform MTG surveillances shall be updated, where appropriate, to reflect the calculated peak positive HFP MTG along with Zr8 2 IFBA 's distinctive trend in RCS critical boron concentration.

Condition 3 Response:

This is not the first application of ZrB 2 in a CE 16x16 fuel design and, as such this measurement is not required to be performed. APS will update plant procedures used to perform MTG surveillances where appropriate.

Condition 4:

Prior to startup following a Condition Ill or IV event, licensees must evaluate clad hydriding to ensure that hydrides have not precipitated in the radial direction (in accordance with Section 3.2 of this SE).

Condition 4 Response:

Hydriding will be evaluated by the appropriate functional group should a Condition Ill or IV event occur.

Condition 5:

CEN-372-P-A constraints and limitations with regard to rod internal pressure and DNB propagation must continue to be met. In addition, licensees must ensure that the following two conditions are satisfied:

a. For Condition I (normal), Condition II (moderate frequency), and Condition Ill (infrequent) events, fuel cladding burst must be precluded for Zr8 2 IFBA fuel rods. Using models and methods approved for CE fuel designs, licensees must demonstrate that the total calculated stress remains below cladding burst stress at the cladding temperatures experienced during any potential Condition II or Condition Ill event. Within the confines of the plant's licensing basis, licensees must evaluate all Condition II events in combination with any credible, single active failure to ensure that fuel rod burst is precluded.

b. For Condition IV non-LOCA events which predict clad burst, the potential impacts offue/ rod ballooning and bursting need to be specifically addressed with regard to coo/able geometry, RCS pressure, and radiological source term.

Condition 5 Response:

a. Rod internal pressure is verified for Condition I, Condition II, and Condition Ill events as part of the approved reload process. The Palo Verde licensing basis allows a limited amount of fuel failure due to DNB for selected Condition Ill events. For these events, DNB propagation (i.e. ballooning) was evaluated and shown to be bounded by the analyses discussed in Attachments 7 and 8 to the enclosure, Section 7.1.3.

b. DNB propagation (i.e. ballooning) is addressed by the analyses discussed in Attachments 7 and 8 to the enclosure, Section 7.4.8. Burst is addressed by the analyses discussed in Attachments 7 and 8 to the enclosure, Section 2.4. Sufficient DNB thermal margin is reserved by the LCOs to prevent all UFSAR Section 15 Design Basis Events (DBEs) from violating the DNB Specified Acceptable Fuel Design Limit (SAFDL), with the exception for the Sheared Shaft/Seized Rotor, Pre-trip Steam Line Break (SLB) and CEA Ejection (CEAE) events. An evaluation was performed to confirm the DNB propagation acceptance criteria are satisfied.

ATTACHMENT 5, Page 15

Enclosure Description and Assessment of Proposed License Amendment

6. CENPD-404-P-A (ZIRLO and Optimized ZIRLO')

CENPD-404-P-A is currently addressed in TS 5.6.5.b as Item 13. The implementation of NGF fuel requires the use of CENPD-404-P-A, Addendum 1-A and Addendum 2-A. The following addresses the limitations and conditions contained in the NRG Safety Evaluation for these topical report addenda.

6.1. CENPD-404-P-A Addendum 1-A Topical report CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO, describes an extension of the regulatory definition of ZIRLO to allow for the optimization of ZIRLO' for enhanced corrosion resistance in more adverse in-reactor primary chemistry environments and at higher fuel duties with higher burnups.

The limitations and conditions contained in the NRG Safety Evaluation for this topical report are met as follows:

Condition 1:

Until rulemaking to 10 CFR Part 50 addressing Optimized ZIRLO' has been completed, implementation of Optimized ZIRLO' fuel clad requires an exemption from 10 CFR 50.46 and 10 CFR Part 50 Appendix K.

Condition 1 Response:

Section 7 of this Enclosure adc;:lresses the proposed change for a permanent exemption from these requirements.

Condition 2:

The fuel rod burnup limit for this approval remains at currently established limits: 62 GWd/MTU for Westinghouse fuel designs and 60 GWd!MTU for CE fuel designs.

Condition 2 Response:

This requirement is satisfied. The bounding fuel performance analysis performs evaluations up to 60 GWd/MTU. The fuel rod burnup limit is to be verified on a cycle-specific basis per the approved reload process.

Condition 3:

This condition contains proprietary information and is therefore not restated in this attachment.

It is available in the NRG staff safety evaluation, which is included in the approved topical report.

Condition 3 Response:

This condition is superseded by the requirements and limitations approved in CENPD-404-P-A, Addendum 2-A (Reference 1.2). The maximum best estimate oxide thickness is limited to 100 microns. The maximum best estimate clac;:lding hydrogen content is limited to the value specified in Attachment 8 of the enclosure, Section 6, Fuel Corrosion Analysis. A Thermal Reaction Accumulated Duties (TRD) limit associated with 100 microns of oxide is also required.

Condition 4:

All the conditions listed in previous NRG SE approvals for methodologies used for standard ZIRLO and Zircaloy-4 fuel analysis will continue to be met, except that the use of Optimized ZIRLO' cladding in addition to standard ZIRLO and Zircaloy-4 cladding is now approved.

ATTACHMENT 5, Page 16

Enclosure Description and Assessment of Proposed License Amendment Condition 4 Response:

The methodologies that apply to ZIRLO were utilized for the STD fuel. All conditions listed in previous NRC Safety Evaluation approvals for methodologies used for ZIRLO and Zircaloy-4 fuel analysis will continue to be met for Optimized ZIRLO'.

Condition 5:

All methodologies will be used only within the range for which ZIRLO and Optimized ZIRLO' data were acceptable and for which the verifications discussed in Addendum 1 and responses to RAls were performed.

Condition 5 Response:

PVNGS fuel performance evaluations performed were within acceptable data range for ZIRLO and Optimized ZIRLO' as discussed in Optimized ZIRLO' topical report CENPD-404-P-A, Addendum 1-A (Reference 1.1).

Condition 6:

The licensee is required to ensure that Westinghouse has fulfilled the following commitment:

Westinghouse shall provide the NRG staff with a letter(s) containing the following information:

a. Optimized ZIRLO' LTA data from Byron, Calvert Cliffs, Catawba, and Millstone.

i. Visual ii. Oxidation of fuel rods iii. Profilometry iv. Fuel rod length

v. Fuel assembly length

b. Using the standard and Optimized ZIRLO' database including the most recent LTA data, confirm applicability with currently approved fuel performance models (e.g., measured versus predicted).

Confirmation of the approved models' applicability up through the projected end of cycle bum up for the Optimized ZIRLO' fuel rods must be completed prior to their initial batch loading and prior to the startup of subsequent cycles. For example, prior to the first batch application of Optimized ZIRLO', sufficient LTA data may only be available to confirm the models' applicability up through 45 GWd/MTU. In this example, the licensee would need to confirm the models up through the end of the initial cycle. Subsequently, the licensee would need to confirm the models, based upon the latest LTA data, prior to re-inserting the Optimized ZIRLO' fuel rods in future cycles. Based upon the LTA schedule, it is expected that this issue may only be applicable to the first few batch implementations since sufficient LTA data up through the bumup limit should be available within a few years.

Condition 6 Response:

This condition has been satisfied. Westinghouse has provided the NRC with the required Lead Test Assembly (LTA) data (Reference 1.15).

Condition 7:

The licensee is required to ensure that Westinghouse has fulfilled the following commitment Westinghouse shall provide the NRG staff with a letter containing the following information:

a. Vogtle growth and creep data summary reports ATTACHMENT 5, Page 17

Enclosure Description and Assessment of Proposed License Amendment

b. Using the standard ZIRLO and Optimized ZIRLO' database including the most recent Vogtle data, confirm applicability with currently approved fuel performance models (e.g.,

level of conservatism in W rod pressure analysis, measured versus predicted, predicted minus measured vs. tensile and compressive stress).

Confirmation of the approved models' applicability up through the projected end of cycle burn up for the Optimized ZIRLO' fuel rods must be completed prior to their initial batch loading and prior to the startup of subsequent cycles. For example, prior to the first batch application of Optimized ZIRLO', sufficient LTA data may only be available to confirm the models' applicability up through 45 GWd/MTU. In this example, the licensee would need to confirm the models up through the end of the initial cycle. Subsequently, the licensee would need to confirm the models, based upon the latest LTA data, prior to re-inserting the Optimized ZIRLOTM fuel rods in future cycles. Based upon the LTA schedule, it is expected that this issue may only be applicable to the first few batch implementations since sufficient LTA data up through the burnup limit should be available within a few years.

Condition 7 Response:

This condition has been satisfied. Westinghouse has provided the NRC with the Vogtle test data (Reference 1.15).

Condition 8:

The licensee shall account for the relative differences in unirradiated strength (YS and UTS) between Optimized ZIRLO' and standard ZIRLO in cladding and structural analyses until data for Optimized ZIRLO' have been collected and provided to the NRG staff. *

a. For the Westinghouse fuel design analyses:

i. The measured, unirradiated Optimized ZIRLO' strengths shall be used for BOL analyses.

ii. Between BOL up to a radiation f/uence of 3.0 x 1021 nlcm 2 (E>1MeV), pseudo-irradiated Optimized ZIRLO' strength set equal to linear interpolation between the following two strength level points: At zero f/uence, strength of Optimized ZIRLO' equal to measured strength of Optimized ZIRLO' and a f/uence of 3.0 x 1021 nlcm 2 (E>1MeV), irradiated strength of standard ZIRLO at the f/uence of 3.0 x 1021 nlcm 2 (E>1MeV) minus 3 ksi.

iii. During subsequent irradiation from 3.0 x 1021 nlcm 2 up to 12 x 1021 nlcm2, the differences in strength (the difference at a fluence of 3 x 1a2 1 nlcm 2 due to tin content) shall be decreased linearly such that the pseudo-irradiated Optimized ZIRLO' strengths will saturate at the same properties as standard ZIRLO at 12 x 1021 nlcm 2*

b. For the CE fuel design analyses, the measured, unirradiated Optimized ZIRLO' strengths shall be used for all f/uence levels (consistent with previously approved methods).

Condition 8 Response:

The measured, unirradiated Optimized ZIRLO' strengths are used for all fluence levels (consistent with previously approved methods).

Condition 9:

As discussed in response to RA/ #21, for plants introducing Optimized ZIRLO' that are licensed with LOCBART or STRIKIN-11 and have a limiting PCT that occurs during blowdown or early reflood, the limiting LOCBART or STRIKIN-11 calculation will be rerun using the specified Optimized ZIRLO' material properties. Although not a condition of approval, the NRG staff ATTACHMENT 5, Page 18

Enclosure Description and Assessment of Proposed License Amendment strongly recommends that for future evaluations, Westinghouse update all computer models with Optimized ZIRLO' specific material properties.

Condition 9 Response:

The ECCS method of analysis has been updated to model the specific material properties.

These analyses are discussed in Attachments 7 and 8, Section 8.

Condition 10:

Due to absence of high temperature oxidation data for Optimized ZIRLO', the Westinghouse coo/ability limit on PCT during the locked rotor event shall be 2375°F.

Condition 10 Response:

This condition is specific to Westinghouse plant design safety analyses, which have a cladding temperature limit on the locked rotor analysis. This locked rotor peak clad temperature limit is not applicable to the safety analysis methodologies for Combustion Engineering designed plants such as PVNGS.

6.2. CENPD-404-P-A Addendum 2-A Topical Report CENPD-404-P-A, Addendum 2-A, Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO', describes the Westinghouse ZIRLO fuel rod cladding corrosion model that replaces the existing ZIRLO corrosion model developed when ZIRLO was first licensed.

The limitations and conditions contained in the NRC Safety Evaluation for this topical report are

• met as follows:

Condition 1:

This condition contains proprietary information and is therefore not restated in this attachment.

It is available in the NRG staff safety evaluation, which is included in the approved topical report.

Condition 1 Response:

The licensing calculations performed for the PVNGS NGF fuel performance analysis do consider a best estimate clad oxide thickness limited to a peak value of 100 microns.

Condition 2:

This condition contains proprietary information and is therefore not restated in this attachment.

It is available in the NRG staff safety evaluation, which is included in the approved topical report.

Condition 2 Response:

A hydrogen pickup content (or limit) as specified in Attachment 8 to the enclosure, Section 6, Fuel Corrosion Analysis, is used in the ZIRLO and Optimized ZIRLO' PVNGS evaluations.

Condition 3:

The NRG staff disapproves the Westinghouse assertion that a single corrosion limit could ensure cladding integrity without a separate hydrogen pickup limit. The hydrogen pickup limit in the current existing topical reports including References 9 and 1O [WCAP-12610-P-A and CENPD-404-P-A Addendum 1-A] for ZIRLO and Optimized ZIRLO' cladding shall not be replaced.

ATTACHMENT 5, Page 19

Enclosure Description and Assessment of Proposed License Amendment Condition 3 Response:

The hydrogen pickup limit as specified in Attachment 8 to the enclosure, Section 6, Fuel Corrosion Analysis, was retained and evaluated in the PVNGS NGF licensing evaluation.

Condition 4:

Condition 4 of CENPD-404-P-A SE can be removed. And, the NRG staff disapproves the use of the FOi-based corrosion model for any future licensing applications.

Condition 4 Response:

Fuel Duty Index based corrosion calculations were not performed as part of the PVNGS NGF licensing evaluation.

7. CENPD-183-A (Loss.of Flow Analysis)

Topical report CENPD-183-A, C-E Methods for Loss of Flow Analysis, is currently addressed in TS 5.6.5.b as Item 19. This will be implemented differently with the introduction of NGF.

Topical report CENPD-183-A describes the assumptions, conservatisms and basic methods used for analyzing loss of reactor coolant forced flow events. The main body of the report describes a loss of flow analysis method for use with a computer code having transient core thermal hydraulic capabilities (referred to as the dynamic method). The appendix describes a similar loss of flow analysis method for use with a steady state core thermal hydraulic code (referred to as the static method).

The limitations and conditions imposed by the NRC on the Loss of Flow Topical Report are identical to the restrictions currently in effect at PVNGS. The Limitations and Conditions contained in the NRC Safety Evaluation for this topical report are met as follows:

Condition 1:

The computer codes specifically approved by the NRG for use in conjunction with performing LOF analyses using the approach described in GENPD-183 are:

a. COAST

b. QUJX

c. COSMOIW3

d. TORC/CE-1

e. CESEC Therefore no other computer codes may be used without prior NRG approval.

Condition 1 Response:

The Non-LOCA related conditions imposed by the NRC in the CENPD-183-A Safety Evaluation have been complied with as part of the loss of flow analyses supporting implementation of the NGF design at PVNGS. The computer codes utilized in the NGF loss of flow analyses differ from those cited in Condition 1. Since the time of submittal of CENPD-183-A (Reference 1.17),

a revised set of computer codes have received NRC approval as replacements for those cited in Condition 1. Specifically, the CENTS code (Reference 1.19) was used to generate both the flow coastdown curve and the system response replacing the COAST and CESEC codes, the transient's neutronics response is modeled with the HERMITE code (Reference 1.20) rather than the QUIX/COSMO codes, and the thermal-hydraulic (DNBR) response is modeled with the ATTACHMENT 5, Page 20

Enclosure Description and Assessment of Proposed License Amendment VIPRE-01 (Reference 1.21) or CETOP-D code (Reference 1.22) replacing TORC code with DNB correlations as discussed.

Condition 2:

These assumptions will result in lower DNBR and are therefore acceptable.

a. The assumptions referred to are:

i. Most adverse initial conditions ii. Most adverse reactivity coefficients iii. Maximum system response delay Condition 2 Response:

The initial condition assumptions cited in Condition 2 are consistent with those utilized in the PVNGS Loss-of-Flow analyses.

Condition 3:

If COSMOJW-3 is used for DNBR calculations, the applicant is required to submit a fuel damage probability distribution for staff's approval.

Condition 3 Response:

The probability density function (pdf) correlation used for PVNGS has been generated with ABB-NV and WSSV CHF data. The ABB-NV and WSSV fuel failure data were generated inherently in the fuel failure calculation based on the NRG-approved CHF correlation statistics in WCAP-14565-P-A Addendum 2-P-A (Reference 1.14) and WCAP-16523-P-A (Reference 1.11).

The data interface for the rods-in-DNB calculation process no longer requires the DNB PDF data be provided in table form such as Table 2 *of CENPD-183-A.

8. Attachment 5 References 1.1. CENPD-404-P-A, Addendum 1-A, OptimizedZIRLO', July 2006 1.2. CENPD-404-P-A, Addendum 2-A, Westinghouse Clad Corrosion Model for ZIRLO and Optimized ZIRLO', October 2013 1.3. EPRI NP-2511-CCM-AM, Mod 02, Revision 3, VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores, Volume 1through4 (Revision 4, February 2001), and Volume 5 (March 1988), Electric Power Research Institute 1.4. Letter from C. E. Rossi (USNRC) to J. A. Blaisdell (UGRA Executive Committee),

Acceptance for Referencing of Licensing Topical Report, EPRI NP-2511-CCM, "VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, Volumes 1, 2, 3 and 4, May 1, 1986 1.5. Letter from A. C. Thadani (USNRC) to Y. Y. Yung (VIPRE-01 Maintenance Group),

Acceptance for Referencing of the Modified licensing Topical Report, EPRI NP-2511..:

CCM, Revision 3, VIPRE-01: A Thermal-Hydraulic Analysis Code for Reactor Cores, (TAC No. M79498), October 30, 1993 1.6. WCAP-14565-P-A, Revision 0, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non LOCA Thermal-Hydraulic Safety Analysis, October 1999 ATTACHMENT 5, Page 21

Enclosure Description and Assessment of Proposed License Amendment 1.7. WCAP-16500-P-A, Supplement 2, Evolutionary Design Changes to CE 16x16 Next Generation Fuel and Method for Addressing the Effects of End-of-Life Properties on Seismic and Loss of Coolant Accident Analyses, June 2016 1.8. Westinghouse Document, LTR-NRC-10-40, Information Satisfying WCAP-16500-P-A SER Condition 4 (Proprietary/Non-Proprietary), July, 2010 1.9. WCAP-16500-P-A Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF), December 2010 1.10. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Report, August 2007 1.11. WCAP-16523-P-A, Westinghouse Correlations WSSV and WSSV-T for Predicting Critical Heat Flux in Rod Bundles with Side-Supported Mixing Vanes, August 2007 1.12. CENPD-132, Supplement 4-P-A, Calculative Methods for the C-E Nuclear Power Large Break LOCA Evaluation Model, March 2001 [NRC ADAMS Accession No. ML071730336]

1.13. WCAP-14565-P-A, Addendum 1-A, Addendum 1 to WCAP-14565-P-A Qualification of ABB Critical Heat Flux Correlations with VIPRE-01 Code, August 2004 1.14. WCAP-14565-P-A, Addendum 2-P-A, Addendum 2 to WCAP-14565-P-A, Extended Application of ABB-NV Correlation and Modified ABB-NV Correlation WLOP for PWR Low Pressure Applications, April 2008 1.15. Westinghouse Document, LTR-NRC-13-6, SER Compliance with WCAP-12610-P-A &

CENPD-404-P-A, Addendum 1-A, Optimized ZIRLO' (Proprietary/Non-proprietary),

February 2013 1.16. WCAP-16072-P-A, Revision 0, Implementation of Zirconium Diboride Burnable Absorber Coatings in CE Nuclear Power Fuel Assembly Designs, August 2004 1.17. CENPD-183-A, Loss of Flow C-E Methods for Loss of Flow Analysis, June 1984 1.18. Letter from C. M. Trammell (USNRC) to W. F. Conway (APS), Approval of Reload Analysis Methodology Report- Palo Verde Nuclear Generating Station (TAC Nos.

M85153, M85154, and M85155), June 14, 1993 1.19. CENPD-282-P-A, Revision 2/WCAP-15996-P~A, Revision 1, Technical Description Manual for the CENTS Code, March 2005

• 1.20. CENPD-188-P-A, HERMITE A Multi-Dimensional Space-Time Kinetics Code forPWR Transients, July 1976 1.21. WCAP-14565-P-A, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis, October 1999 1.22. CEN(S)-P, Revision 1-P, CETOP-D Code Structure and Modeling Methods for San Onofre Nuclear Generating Station Units 2 and 3; September 1981 ATTACHMENT 5, Page 22

Enclosure Description and Assessment of Proposed License Amendment ATTACHMENT 6 Affidavit from the Westinghouse Electric Company Submitted in Accordance with 10 CFR 2.390 to Consider Attachment 8 as a Proprietary Document

Westinghouse Non-Proprietary Class 3

@Westinghouse Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412) 374-4643 Document Control Desk Direct fax: (724) 940-8560 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 CAW-16-4439 June 30, 2016 APPLICATION FOR WITHHOLDING,PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

Attachment 8 to the Enclosure of the License Amendment Request and Exemption Request to Support the Implementation of Next Generation Fuel (Proprietary) .

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westinghouse Electric Company LLC (Westinghouse), pursuant to the provisions of paragraph (b)(1) of Section 2.390 of the Commission's regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced report is further identified in Affidavit CAW~ 16-4439 signed by the owner of the proprietary information, Westinghouse Electric Company LLC. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (b)(4) of 10 CFR Section 2.390 of the Commission's regulations.

Accordingly, this letter authorizes the utilization of the accompanying Affidavit by Arizona Public Service Company.

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-16-4439, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite* 310, Cranberry Township, Pennsylvania 16066. *

)QJ~

James A. Gresham, Manager Regulatory Compliance

© 2016 Westinghouse Electric Company LLC. All Rights Reserved.

CAW-16-4439 June 30, 2016 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

SS COUNTY OF BUTLER:

I, James A. Gresham, am authorized to execute this Affidavit on behalf of Westinghouse Electric Company LLC (Westinghouse), and that the averments of fact set forth in this Affidavit are true and correct to the best of my knowledge, information, and belief.

2 CAW-16-4439 (1) I am Manager, Regulatory Compliance, 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) l 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 heid in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

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

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

3 CAW-.16-4439 Westinghouse's competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

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

(c) Its use by a competitor would reduce his expenditure of resources or improve his competitive position in the design, manufacture, shipment, installation, assurance of quality, or licensing a similar product.

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

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

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

(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 competitive position.

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

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

4 CAW-16-4439 (d) Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

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

(f) The Westinghouse capacity to invest corporate assets in research and development depends upon the success in obtaining and maintaining a competitive advantage.

(iv) The information is being transmitted to the Commission in confidence and, under the provisions of 10 CPR Section 2.390, 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 Attachment 8 to the Enclosure of the License Amendment Request and Exemption Request to Support the Implementation of Next Generation Fuel (Proprietary), for submittal to the Commission, being transmitted by Arizona Public Service Company letter and Application for Withholding Proprietary Information from Public Disclosure, to the Document Control Desk. The proprietary information as submitted by Westinghouse for use by Palo Verde Nuclear Generation Station Units 1, 2, and 3 to support the transition to the Next Generation Fuel product, and may be used only for that purpose.

5 CAW-16-4439 (a) This information is part of that which will enable Westinghouse to assist customers in implementing an improved fuel product to support improving their fuel performance.

(b) Further, this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of ensuring customers the highest quality fuel product.

(ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific 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 technical evaluation justifications 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 non-proprietary versions of a document, furnished to the NRC to support the transition to the Next Generation Fuel product. *

• In order to conform to the requirements of I 0 CPR 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 CPR 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 versjons of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.