ML18296A466

Supplemental Information Regarding License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel

ML18296A466
Person / Time
Site:	Palo Verde
Issue date:	10/18/2018
From:	Lacal M Arizona Public Service Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML18296A464	List: ML18296A466
References
102-07807-MLL/MDD
Download: ML18296A466 (35)

Text

Enclosure Attachment 3 contains PROPRIETARY information To be withheld under 10 CFR 2.390 10 CFR 50.90 10 CFR 50.12 Qaps MARIA L. LACAL Senior Vice President Nuclear Regulatory and Oversight Palo Verde Nuclear Generating Station P.O. Box 52034 102-07807-MLiyMDD Phoenix, AZ 85072 Mail Station 7605 October 18, 2018 Tei 623.393.6491 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001

References:

1. Arizona Public Service Company (APS) letter number 102-07727, License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel, dated July 6, 2018 [Agencywide Documents Access and Management System (ADAMS) Accession Number ML18187A417]

2. NRC letter Palo Verde Nuclear Generating Station, Units 1, 2, and 3

- Supplemental Information Needed for Acceptance of Requested License Amendments and Exemptions RE: Implementation of Framatome High Thermal Performance Fuel, dated October 2, 2018 (ADAMS Accession No. ML18271A039)

Subject:

Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos. STN 50-528, 50-529, and 50-530 Supplemental Information Regarding License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel In Reference 1, Arizona Public Service Company (APS) submitted a license amendment request (LAR) to revise the Technical Specifications (TS) for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 to support the implementation of Framatome (formerly AREVA, Inc.) Advanced Combustion Engineering 16x16 (CE-

16) High Thermal Performance (HTP^") fuel design with M5 as a fuel rod cladding material and gadolinia as a burnable absorber. In addition to the license amendment, pursuant to 10 CFR 50.12, Specific exemptions, APS requested an 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 A member of the STARS Alliance, LLC Callaway

• Diablo Canyon

• Palo Verde

• Wolf Creek Attachment 3 transmitted herewith contains PROPRIETARY information.

When separated from Attachment 3, this transmittal is decontrolled.

102-07807-MLL7MDD ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Suppiementai Information Regarding License Amendment Request for Impiementation of Framatome Fuel Page 2 50, Appendix K, ECCS Evaluation Models, to ailow the use of the Framatome M5 alioy as a fuei rod ciadding materiai. In Reference 2, the Nuclear Regulatory Commission (NRC) staff requested supplemental information to support the acceptance review of the LAR.

The enclosure to this letter provides supplemental information in response to the NRC request. The enclosure also contains three attachments. Attachments 1 and 3 provide non-proprietary and proprietary responses to the NRC request, respectively. is an affidavit signed by APS that sets forth the basis on which the proprietary information in Attachment 3 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 3 or the supporting APS affidavit should be addressed to Bruce Rash, Vice President, Nuclear Engineering, of APS.

Further, during preparation of responses to the NRC request, three editorial errors were discovered in the original LAR. These include an incorrect docket number for Unit 2 on the cover letter, incorrectly numbered references in Section 6 of Attachments 8 and 10 of the enclosure, and an incorrect example provided in Section 5.4.3 of Attachments 8 and 10 of the enclosure stating that the NRC has previously approved the use of the CE-1 critical heat flux correlation with the VIPRE-W thermal-hydraulic code at PVNGS. These errors are discussed further in the enclosure to this letter.

APS has reviewed the information supporting a finding of no significant hazards consideration previously provided to the NRC in Reference 1. APS has concluded that the information provided in this response does not affect the basis for concluding that the proposed license amendment does not involve a significant hazards consideration under the standards set forth in 10 CFR 50.92.

By copy of this letter, this supplemental information is being forwarded to the Arizona Department of Health Services Bureau of Radiation Control in accordance with 10 CFR 50.91(b)(1).

Should you have any questions concerning the content of this letter, please contact Matthew S. Cox, Licensing Section Leader, at (623) 393-5753.

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

Executed on October 18, 2018 (Date)

Sincerely, MLI7MDD/SMM/sa

102-07807-MLL/MDD ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Supplemental Information Regarding License Amendment Request for Implementation of Framatome Fuel Page 3

Enclosure:

Supplemental Information Regarding License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel cc: K. M. Kennedy NRC Region IV Regional Administrator M. D. Orenak NRC NRR Project Manager for PVNGS M. M. O'Banion NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS B. GoretzkI Arizona Department of Health Services (ADHS) - Bureau of Radiation Control

ENCLOSURE Supplemental Information Regarding License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel - Responses to NRC Questions [NON-PROPRIETARY VERSION] - Affidavit from Arizona Public Service Company Submitted in Accordance with 10 CFR 2.390 to Consider Attachment 3 as a Proprietary Document - Responses to NRC Questions [PROPRIETARY VERSION]

Enclosure Supplemental Information Regarding License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel Introduction By letter dated July 6, 2018 [Agencywide Documents Access and Management System (ADAMS) Accession Number ML18187A417], Arizona Public Service Company (APS) submitted a license amendment request (LAR) to revise the Technical Specifications for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3 to support the implementation of Framatome (formerly AREVA, Inc.) Advanced Combustion Engineering 16x16 (CE-16) High Thermal Performance (HTP') fuel design with M5 as a fuel rod cladding material and gadolinia as a burnable absorber. In addition to the license amendment, pursuant to 10 CFR 50.12, Specific exemptions, APS requested an 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, ECCS Evaluation Models, to allow the use of the Framatome M5 alloy as a fuel rod cladding material.

During the acceptance review of the license amendment and exemption applications, the U.S.

Nuclear Regulatory Commission (NRC) staff identified the need for supplemental information to support the acceptance review of the LAR. These questions were provided to APS by letter dated October 2, 2018 (ADAMS Accession No. ML18271A039). Each of the NRC staff information requests are provided first, followed by the APS response to each request.

During preparation of responses to the aforementioned NRC requests, three editorial errors were discovered in the original LAR. These include an incorrect docket number for Unit 2 on the cover letter (59-529 instead of 50-529), incorrectly numbered references in Section 6 of attachments 8 and 10 of the enclosure, and an incorrect example provided in Section 5.4.3 of attachments 8 and 10 of the enclosure. The incorrect example (bullet 2) stated that the NRC has previously approved the use of the CE-1 critical heat flux (CHF) correlation with the VIPRE-W thermal-hydraulic code at PVNGS. The CE-1 CHF correlation had not been previously approved for use with the VIPRE-W code. Bullet 2 should be reworded as follows:

• Implementing the CE-1 CHF correlation, which the NRC has previously approved for use with the VtPRE-W, CETOP-Dr and TORC codes, for use with the VIPRE-01 and/or VIPRE-W codes; and This enclosure provides supplemental information in response to the NRC request and contains three attachments. Attachments 1 and 3 provide non-proprietary and proprietary responses to the NRC request, respectively. Attachment 2 is an affidavit signed by APS that sets forth the basis on which the proprietary information in Attachment 3 may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in 10 CFR 2.390(b)(4).

ATTACHMENT 1 Responses to NRC Questions

[NON-PROPRIETARY VERSION]

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 1 In Section 5.4.1.1, of Attachment 10 of the application, APS discusses its use of the VIPRE-01 computer code. VlPRE-01 has been previously reviewed and approved by the NRC staff.

However, as noted by the NRC staff in its safety evaluation, Each organization using VIPRE-01 for licensing calculations should submit separate documentation describing how they intend to use VIPRE-01 and providing justification for their specific modeling assumptions, choice of particular two-phase flow models and correlations, heat transfer correlations, CHF [critical heat flux] correlation and DNBR [departure from nucleate boiling ratio] limit, input values of plant specific data such as turbulent mixing coefficient, slip ratio, grid loss coefficient, etc., including defaults."

In Section 5 of Enclosure 10 of the application, APS did not provide the details necessary for the NRC staff to approve the use of VIPRE-01. Please submit the information that describes how APS intends to use VIPRE-01.

Response to NRC Question 1 APS intends to use the Electric Power Research Institute (EPRI) version of VIPRE-01 in the same manner as the NRC approved Westinghouse version of VIPRE-01, herein referred to as VIPRE-W" [Reference (1)], for core reload thermal-hydraulic analyses that are the basis for Departure from Nucleate Boiling Ratio (DNBR) driven fuel failure analyses for Chapter 15 Updated Final Safety Analysis Report (UFSAR) events [Reference (2)]. These events include but are not limited to:

• Pre-trip steam line break

• Control Element Assembly (CEA) withdrawal from subcritical or at power

• Anticipated Operation Occurrence (AOO) from Specified Acceptable Fuel Design Limit (SAFDL)

• Loss of forced reactor coolant flow

• Locked reactor coolant pump rotor or shaft break

• Startup of an inactive reactor coolant pump (cold water accident)

In addition to the transients listed above, VIPRE-01 will be used to establish analytical limits for the DNBR SAFDL, for analysis of fuel failure via statistical convolution, and in the CETOP benchmarking and setpoints determination processes.

Section 1.3 and Figure 1-5 of Attachment 10 of the application [Reference (3)] present an overview of the APS reload analysis process and methodology including changes to interfacing computer programs. Certain inputs such as boundary conditions for APS VIPRE-01 calculations will be obtained from other PVNGS licensing basis computer codes, including a reactor systems analysis code such as CENTS, reactor physics codes such as CASMO and SIMULATE, and a fuel performance code such as (following approval of the APS application [Reference (3)])

COPERNIC. CENTS provides time-dependent parameters such as reactor coolant system pressure, core power, core coolant inlet flow rate, and core coolant inlet enthalpy. CASMO and SIMULATE provide data such as core power distribution and peaking factors. COPERNIC provides data such as temperature and enthalpy in the fuel rods.

1

Enclosure Attachment 1 NON-PROPRIETARY APS is approved to use the TORC and VIPRE-W codes for reload core thermal-hydraulics.

VIPRE-01 use allows a vendor independent code option that is consistent with our overall strategy.

Because VIPRE-W has already been approved for use at APS [Reference (2)], many of the modeling assumptions used in VIPRE-01 are based on those used in the APS implementation of VIPRE-W. Table 1 presents VIPRE-01 with Framatome fuel model choices that are based on those used for VIPRE-W with Westinghouse fuel. Table 2 presents VIPRE-01 with Framatome fuel model choices that are different from VIPRE-W with Westinghouse fuel.

The key modeling options, including the additional assumptions and correlations, are consistent with the recommendations of the VIPRE-01 code vendor based on the results of previous VIPRE-01 sensitivity studies (for example, EPRI NP-2511-CCM-A [Reference (4)]) and the vendors experience with the VIPRE-01 code.

The key modeling options for the VIPRE-W analysis of Framatome CE16HTP fuel are provided in Section 5.4.2.1 and Table 5-7 of Attachment 10 of the application [Reference (3)]. In general, the VIPRE-W analysis utilizes the same key modeling options, including assumptions and correlations, as modeled in the VIPRE-01 analysis.

Section 5.4.2 of Attachment 10 of the application [Reference (3)] shows a comparison of VIPRE-01 results using the selected modeling assumptions listed in Section 5.4.1.1 of that attachment, to VIPRE-W results using the selected modeling assumptions that are listed in Section 5.4.2.1 of that attachment. As noted in Section 5.4.2 of that attachment, the benchmarking of the VIPRE-W code with the BHTP Critical Heat Flux (CHF) correlation to the VIPRE-01 code with the BHTP CHF correlation provides acceptable results. This benchmarking provides additional justification that the modeling assumptions made for Framatome fuel modeling with VIPRE-01, and VIPRE-W, provides accurate and acceptable results.

The determination of the DNBR limit for Framatome CE16HTP fuel was developed consistent with the Modified Statistical Combination of Uncertainties (MSCU) process as approved by the NRC in CEN-356(V)-P-A [Reference (5)] and WCAP-16500-P-A, Supplement 1, Revision 1

[Reference (6)]. This method was upgraded by the NRC in Amendment No. 205 to the PVNGS Operating Licenses [Reference (2)]. System parameters uncertainties specific to the CE16HTP fuel manufacturing and design were addressed.

Enclosure Attachment 1 NON-PROPRIETARY Table 1 VIPRE-01 with Framatome Fuel Model Choices That Are Based on Those Used for VIPRE-W with Westinghouse Fuei

Enclosure Attachment 1 NON-PROPRIETARY Table 2 VIPRE-01 with Framatome Fuel Model Choices That Are Different from VIPRE-W with Westinghouse Fuel

Enclosure Attachment 1 NON-PROPRIETARY Table 2 (continued)

VIPRE-01 with Framatome Fuel Model Choices That Are Different from VIPRE-W with Westinghouse Fuel Table 2 Notes The key modeling options for the VIPRE-01 analysis of Framatome CE16HTP fuel (which is not covered under WCAP-14565-P-A [Reference (1)]) are provided in Section 5.4.1.1 and Table 5-5 of Attachment 10 of the application [Reference (3)]. The following provides additional detail for some of these key modeling options and for additional assumptions and correlations.

((

Enclosure Attachment 1 NON-PROPRIETARY

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 2 Since the publication of Regulatory Guide (RG) 1.77, Assumptions Used for Evaluating a Control Rod Ejection Accident for Pressurized Water Reactors (ADAMS Accession No. ML003740279), the acceptance criteria of 280 calories per gram of deposited enthalpy has been found inadequate to ensure fuel rod geometry and long-term coolability. The NRC staff documented its position on RG 1.77 in a letter dated April 3, 2015, Results of Periodic Review of Regulatory Guide 1.77 (ADAMS Accession No. ML15075A311). The position is supported by a guidance document dated January 19, 2007, titled Technical and Regulatory Basis for the Reactivity-Initiated Accident Interim Acceptance Criteria and Guidance (ADAMS Accession No. ML070220400).

Please demonstrate that the control element assembly ejection transient analysis results meet General Design Criterion 28, Reactivity Limits, of Title 10 of the Code of Federal Regulations Part 50, Appendix A, by showing that fuel rod geometry and long-term coolability is maintained and that the effects of burn-up are considered in the analysis.

Response to NRC Question 2 APS cited the 280 cal/g criterion in its application of July 6, 2018 [Reference (3)], because it is part of the PVNGS Current Licensing Basis (CLB) as defined in 10 CFR 54.3. The criterion appeared in NRC Safety Evaluations for PVNGS power uprates in 2003 [Reference (9)] and 2005 [Reference (10)] and has not been modified via subsequent license amendments. For the Reference (3) application, APS will use interim criteria published by NRC staff in Standard Review Plan, Section 4.2, in 2007 [Reference (11)] with plant specific adjustments.

With regard to the proposed implementation of Framatome CE16HTP fuel at PVNGS, APS offers the following to demonstrate that CEA ejection analysis results meet General Design Criterion (GDC) 28, Reactivity Limits, of 10 CFR Part 50 Appendix A, and consider the effects of burnup:

• Core coolability - Peak fuel temperature remains below incipient fuel melting conditions as shown in Table 6-3 of Attachment 10 of the APS application [Reference (3)].

• Core coolability - Peak radial average fuel enthalpy remains below 230 cal/g as shown in Table 6-3 of Attachment 10 of the APS application [Reference (3)].

• Core coolability - There is no loss of coolable geometry due to fuel rod ballooning during a postulated CEA ejection event, as evidenced by the Departure from Nucleate Boiling (DNB) propagation analysis described in Sections 6.2 and 6.5.4.2 of Attachment 10 of the APS application [Reference (3)].

• Fuel cladding failure - Current APS methodology determines the limiting ejected CEA worths at 20%, 50%, 65%, and 95% of rated thermal power, as explained in a letter dated October 11, 2002 [Reference (12)]. Supplemental analysis at FlotZero Power (HZP) conditions yields results less than 150 cal/g.

• Fuel cladding failure - For intermediate (that is, greater than 5% rated thermal power) and full power conditions, fuel cladding failure is determined from the number of fuel

Enclosure Attachment 1 NON-PROPRIETARY rods experiencing DNB, as calculated by statistical convolution. The DNB statistical convolution methodology is described in Section 6.3 of Attachment 10 of the APS application [Reference (3)].

Fuel cladding failure - The change in radial average fuel enthalpy for Framatome CE16HTP fuel is less than the corrosion-dependent limit depicted in Figure B-1 of the NRC interim guidance document [Reference (11)]. The Framatome COPERNIC fuel performance code corrosion model used in this evaluation has been reviewed by the NRC [Reference (13)].

Burnup effects - As indicated in the preceding, the proposed PVNGS CEA ejection analysis methodology for Framatome CE16HTP fuel utilizes the Framatome COPERNIC fuel performance code, which models cladding corrosion as a function of burnup. The COPERNIC code explicitly models other effects of burnup including but not limited to Thermal Conductivity Degradation (TCD). Furthermore, the PVNGS reload cycle design process imposes a thermal-mechanical radial falloff curve that limits fuel rod powers (that is, reduces peaking factors and linear heat rates) at intermediate and high levels of burnup. As stated in Section 6.5.2 of Attachment 10 of the APS application [Reference (3)], certain CEA ejection analysis inputs are derived at the knee of the radial falloff curve to bound fuel performance at all levels of burnup.

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 3 of the application, Insert A to B.2.1.1, indicates that there will be three fuel types which will be used at Palo Verde (CE16STD, CE16NGF, and CE16HTP). However, in Section 5.5 of Attachment 10 of the application, APS seems to limit the mixed core assessment to (1) the inlet flow distribution, and (2) crossflow velocity analysis for mechanical calculations.

Provide a summary of the analysis that describes how mixed cores are treated from a thermal-hydraulic standpoint, especially noting any DNBR penalties which are needed due to the different fuel types. This should be provided for both the VIPRE-W and VIPRE-01 methodologies.

Response to NRC Question 3 The mixed core methodology is summarized in Section 1.4 of Attachment 10 of the application

[Reference (3)]. The following paragraphs provide an overview of the analyses that assess mixed cores from a thermal-hydraulic perspective.

Mixed Core Compatibility Evaluations Several mixed core compatibility evaluations have been performed by both Framatome and Westinghouse. A summary of the mixed core thermal-hydraulic compatibility evaluations and results are discussed in Section 5.7 of Attachment 10 of the application [Reference (3)]. In addition, the impact due to the presence of CE16HTP fuel on Westinghouse fuel was evaluated.

The evaluation found assembly bow force, crossflow velocity, uplift force, and scram differential pressure to be acceptable.

CE16HTP Thermal-Hydraulic Characterization The Framatome assembly design is compared to both types of Westinghouse designs in Table 2-1 of Attachment 10 of the application [Reference (3)]. The CE16HTP fuel is geometrically similar to CE16STD and CE16NGF fuel. The primary hydraulic differences between the three fuel types result from the different bottom grid and intermediate spacer grids each design employs. Additionally, CE16NGF has a slightly smaller fuel rod outer diameter compared to CE16HTP and CE16STD fuel. The CE16HTP and CE16STD assemblies were tested by Framatome and hydraulic loss coefficients were developed. The CE16STD and CE16NGF assemblies were tested by Westinghouse.

To ensure consistency in mixed core VIPRE analyses, the hydraulic loss coefficients for CE16STD, CE16NGF, and CE16HTP fuel types were developed on a consistent basis. The hydraulic loss coefficients for spacer grids, inlet and outlet regions, and bare rod friction factors for all three fuel types were adjusted as required to ensure consistency, based on testing of the CE16STD at Framatome.

Enclosure Attachment 1 NON-PROPRIETARY Mixed Core Thermal Margin Performance Mixed core effects are considered for thermal margin evaluations by recalculation of the inlet flow distribution as described in Section 5.5.1 of Attachment 10 of the application [Reference (3)]. This will be performed each cycle, until full transition to one type of fuel, to capture the mixed core impact on inlet flow distribution.

After inlet flow calculations, the thermal-hydraulic evaluation proceeds with the determination of potentially limiting assemblies. In keeping with the existing process for selecting the limiting fuel assembly, the limiting assembly selection process is unchanged from the process used by APS for many years. Typical DNB limiting assemblies are from Potentially limiting assemblies are further screened in the CETOP benchmark analysis with consideration given to mixed core effects by analyzing each assembly type with its specific geometry and grid loss coefficients. Once the final limiting assemblies are chosen, CETOP benchmarking is conducted to obtain bounding CETOP to VIPRE adjustment factors that cover each assembly type at all operating conditions allowed by the Limiting Conditions for Operation (LCO) and Limiting Safety Systems Settings (LSSS). This process ensures that CETOP is conservative relative to VIPRE. The APS CETOP benchmark process ((l CETOP benchmark methodology is inherently a mixed core analysis process, and no additional mixed core DNBR penalty is necessary. An example of CETOP benchmarking using a CETOP model containing the Combustion Engineering CE-1 CHF correlation is described in Section 5.5.2 of Attachment 10 of the application [Reference (3)]. A branched process is used to develop these adjustment factors. The branched process utilizes a detailed VIPRE model, which includes every fuel assembly in the quarter core, and is used with modeling requirements consistent with WCAP-14565-P-A [Reference (1)] as approved for PVNGS [Reference (2)] to develop adjustment factors for CE16STD or CE16NGF type fuel. A second evaluation is performed using VIPRE as specified in Section 5.4.2.1 of Attachment 10 of the application [Reference (3)] to obtain adjustment factors for CE16HTP fuel.

For each type of DNBR limiting fuel, whether CE16HTP, CE16NGF, or CE16STD, a set of CETOP adjustment factors is selected to bound all fuel of that type. When the same CETOP model is used for multiple fuel types, then a composite set of limiting adjustment factors may be applied to the mixed core.

Cycle specific validation of the DNBR SAFDL and associated probability distribution function is performed following the methods described in Section 5.6 of Attachment 10 of the application

[Reference (3)]. Potentially limiting assemblies are again selected as previously described for evaluation, typically using a 25-channel VIPRE model. The location of the limiting fuel assembly in the core is reflected in the 25-channel model by application of the limiting assembly specific radial power and the inlet flow distribution. This is an acceptable modeling technique approved by the NRC in WCAP-14565-P-A [Reference (1)] and approved for PVNGS [Reference (2)].

Again, a branched process is used to develop the probability distribution function and validate

Enclosure Attachment 1 NON-PROPRIETARY the DNBR SAFDL. A 25-channel VIPRE model is typically used with modeling requirements consistent with WCAP-14565-P-A [Reference (1)] and approved for PVNGS [Reference (2)] for CE16STD or CE16NGF type fuel. A second evaluation is performed using VIPRE as specified in Section 5.4.2.1 of Attachment 10 of the application [Reference (3)] for CE16HTP. As conservatism, the more limiting DNBR limit and associated probability distribution function are then used in the Core Operating Limits Supervisory System (COLSS) and Core Protection Calculator System (CPCS) setpoint process described in Section 11 of Attachment 10 of the application [Reference (3)], that is explained further in the Response to NRC Question 7 in this enclosure.

VIPRE-01 and VIPRE-W Codes The VIPRE-01 and VIPRE-W codes are considered interchangeable for the preceding applications, with the CHF correlation and modeling selections remaining consistent with the licensing of each respective CHF correlation, and VIPRE code modifications described in the application [Reference (3)] and the Response to NRC Question 6 in this enclosure.

Mixed Core DNBR Penalties The preceding CETOP benchmark methodology explicitly accounts for mixed core effects.

Additional conservatism is added by selection of the most limiting DNB limit; therefore, no additional mixed core DNBR penalty is applied.

Thermal-Hydraulic Instability CE16HTP fuel has been designed to be very similar to the resident CE16STD fuel as illustrated in Table 2-1 of Attachment 10 of the application [Reference (3)]. Fuel rod pitch, diameter, heated length, spacer grid elevations, and core flow area are essentially unchanged relative to resident fuel. The geometry is also very similar to CE16NGF fuel, with the exception of rod diameter and the intermediate flow mixing vanes. Additionally, the overall pressure drop of CE16HTP fuel assembly falls between the pressure drops of the CE16STD and CE16NGF designs. By keeping the geometry essentially unchanged from resident fuel, the design of CE16HTP fuel precludes thermal-hydraulic instabilities.

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 4 Pages 12 and 13 of Attachment 5 of the application discuss the use of CENPD-183-A (ADAMS Accession No. ML16224A358) for loss of flow analysis using several codes, such as COAST, QUIX, COSMO/W3, TORC/CE1 and CESEC. The methodology and codes are approved for Combustion Engineering (CE) or Westinghouse type of fuels. Please provide a detailed justification as to how these codes can be applied to the Palo Verde mixed core with Framatome CE 16x16 high thermal performance (HTP) fuel design.

Response to NRC Question 4 This response will be provided in two parts. The first part addresses the history of the computer codes approved for this method expanding on the information provided in Section 6 of of the application [Reference (3)], and the second part addresses application of these codes to Framatome mixed cores.

CENPD-183-A (Loss of Flow Analysis) Codes Topical Report CENPD-183-A, C-E Methods for Loss of Flow Analysis [Reference (14)], is currently addressed in PVNGS Technical Specification 5.6.5, Core Operating Limits Report (COLR). This will be implemented for Framatome CE16HTP fuel in a manner consistent with the PVNGS use of CENPD-183-A for CE16STD and CE16NGF.

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).

PVNGS UFSAR Appendix 15D details the previously NRC approved deviations from the approved CENPD-183-A methodology.

• QUIX: The use of the HERMITE code instead of QUIX was approved in NUREG 0857

[Reference (15)], which directs the reader to NUREG-0852 [Reference (16)], and reaffirmed by the NRC staff in Safety Evaluations dated September 2003 [Section 4.3.0 of Reference (9)] and January 2018 [Section 3.3.12.1 of Reference (2)].

• COSMO: The use of the TORC code in place of COSMOA/V3 was approved by the NRC in NUREG 0857 [Reference (15)], which directs the reader to NUREG-0852

[Reference (16)].

Additional NRC approved changes to the computer codes delineated in CENPD-183-A are as follows:

• COAST: The use of the CENTS computer code to determine the flow coastdown curve in place of the COAST computer code was approved in an NRC Safety Evaluation

[Section 3.3.12.1 of Reference (2)].

Enclosure Attachment 1 NON-PROPRIETARY TORC; The use of either the CETOP or VIPRE-W computer codes for the thermal-hydraulic (DNBR) response as an alternate to the TORC code was approved in an NRC Safety Evaluation [Section 3.3.12.1 of Reference (2)].

CESEC; The use of the CENTS computer code as an option to the use of the CESEC computer code for the transient simulation in the reload analysis was authorized in an NRC Safety Evaluation [Reference (17)]. (Reference (18) is the 3-month notification letter of the intent to use CENTS per Generic Letter 83-11, Supplement 1.) Section 4.3.0 of Reference (9) reaffirmed the use of the CENTS computer code for plant response to non-LOCA transients. Section 3.3.12.1 of Reference (2) also reaffirmed that authorization to use the CENTS computer code as a replacement of CESEC code for the system response transient.

Applicability of the Computer Codes to Model Framatome CE16HTP Fuel

• CENTS: The CENTS computer program is a computer code developed by Combustion Engineering for the simulation of Nuclear Steam Supply System (NSSS) transient behavior under normal and off-normal conditions. The CENTS computer code is documented in Reference (19) and has been approved by the NRC for use in the licensing analyses for Pressurized Water Reactors (PWRs) originally designed by Combustion Engineering [References (20) and (21)].

A review of CENTS indicated that the only models impacted by the use of Framatome fuel are the cladding material properties of thermal conductivity and specific heat.

Section 6.4.4 of Attachment 10 of the application [Reference (3)] discusses the approach to modeling Framatome fuel cladding with CENTS. This approach is consistent with what has been approved previously for other fuel claddings with the use of CENTS [References (22) and (23)]. Consequently, no changes to CENTS are needed to model Framatome fuel cladding.

Consistent with License Condition 4 of Reference (20), only the point kinetics model is used for the core neutronics. The neutronics inputs to CENTS are calculated using detail design codes (for example, SIMULATE). Consequently, no changes to CENTS are needed to model Framatome CE16HTP fuel.

As noted in Section 5.7 of Attachment 10 of the application [Reference (3], the differences in the core differential pressure drop, assembly geometry, and reactor coolant flowrate between fuel types are not substantial, and the CENTS model can be tuned to conservatively model the differences. For example, for loss of flow analyses, the presence of Framatome CE16HTP fuel does not have an adverse effect on flow coastdown modeling because the pressure drop of CE16HTP falls between that of the CE16STD and CE16NGF fuel types.

In summary, the modeling of mixed cores containing Framatome CE16HTP fuel is well within the capability of the CENTS code.

• HERMITE: HERMITE is a space-time kinetics computer code that is approved by the NRC [Reference (24)]. HERMITE was developed for the analysis of design and off-design transients in PWRs by means of a numerical solution to the multi-dimensional.

Enclosure Attachment 1 NON-PROPRIETARY few-group, time dependent neutron diffusion equation including feedback effects of fuel temperature, coolant temperature, coolant density, and control rod motion. The heat conduction equation in the fuel pellet, gap, and clad is solved by a finite difference method. Continuity and energy conservation equations are solved for the coolant enthalpy and density.

A review of HERMITE indicated that the cladding material properties employed are cladding thermal conductivity and specific heat. The HERMITE code is structured in such a way that the default model variables numerical values can be overwritten by the user. Section 6.4.5 of Attachment 10 of the application [Reference (3)] discusses the approach to modeling Framatome fuel cladding with HERMITE. This approach is consistent with what has been approved previously for other fuel claddings with the use of HERMITE [References (22) and (23)]. Consequently, no changes to HERMITE are needed to model Framatome fuel cladding.

The HERMITE code has the capability to model mixed cores containing Framatome CE16HTP fuel as the code is structured in such a way that the default model variables numerical values can be ovenwritten by the user.

• VIPRE: As discussed in Section 5.0 of Attachment 10 of the application [Reference (3)], and in the Response to NRC Question 3 in this enclosure, the VIPRE-W and VIPRE-01 codes can adequately model Framatome CE16HTP fuel in either a full core or mixed core configuration.

• CETOP: As discussed in Section 5.4.3 of Attachment 10 of the application [Reference (3)], and in the Response to NRC Question 5" in this enclosure, CETQP may be modified to allow use of the BHTP correlation. Mixed core configurations will not, however, be explicitly modeled with CETOP because of CETOP modeling limitations.

The appropriate, limiting fuel assembly model and its associated CHF correlation will be used with CETOP.

• TORC; As discussed in Section 5.4.3 of Attachment 10 of the application [Reference (3)], and in the Response to NRC Question 5 in this enclosure, TORC may be modified to allow use of the BHTP correlation. Mixed core configurations may be modeled in TORC in a manner analogous to VIPRE. TORC models will be based on using the appropriate fuel assembly models and their associated CHF correlations.

• Fuel Failure by DNB Statistical Convolution: Section 6.3 of Attachment 10 of the application [Reference (3)] provides the discussion on the use of fuel failure prediction using DNB statistical convolution and its applicability to Framatome CE16HTP fuel and mixed core application.

Based on the preceding, the current PVNGS methodology can be applied to Framatome CE16HTP with only minor input changes as discussed.

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 5 Section 5.4.3 of Attachment 10 (proprietary) of the application discusses the proposed implementation of the BHTP CHF correlation with the VIPRE-01 and VIPRE-W thermal-hydraulic codes and with several other thermal-hydraulics codes at Palo Verde. The examples cited are:

• Implementing the ABB-NV or WSSV CHF correlations, which the NRC has previously approved for use with the VIPRE-W code, for use with the VIPRE-01 code;

• Implementing the CE-1 CHF correlation, which the NRC has previously approved for use with the VIPRE-W, CETOP-D, and TORC codes, for use with the VIPRE-01 code; and

• Implementing the BHTP CHF correlation, which is anticipated to be approved for use with the VIPRE-01 and VIPRE-W code as described in this license amendment request, with the CETOP-D code and/or the TORC code.

Please provide detailed explanations for inserting the different correlations into the above codes with respect to validation, verification, and sensitivity studies.

Response to NRC Question 5 Section 5.4.3 of Attachment 10 of the APS application [Reference (3)] proposed a process for use on a forward-going basis, whereby CHF correlations could be inserted into or coupled with thermal-hydraulic computer codes and qualified for PVNGS licensing applications, without necessarily requiring a separate License Amendment Request (LAR) pursuant to 10 CFR 50.90.

This process would be subject to the following constraints:

• It could only be used for CHF correlations previously approved for use at PVNGS by NRC and described in topical reports and other documents referenced in PVNGS Technical Specification 5.6.5, Core Operating Limits Report (COLR), namely the CE-1, ABB-NV, WSSV, WSSV-T, WLOP, and Macbeth correlations, and (following requested approval) the BHTP correlation. (Note: The APS application [Reference (3)] stated that NRC had previously approved the CE-1 correlation for use with VIPRE-W. The NRC, however, had explicitly approved the ABB-NV correlation for use with VIPRE-W, with ABB-NV developed from the same test data as CE-1.)

• It could only be used with thermal-hydraulic codes previously approved for use at PVNGS by NRC and described in topical reports and other documents referenced in PVNGS Technical Specification 5.6.5, Core Operating Limits Report (COLR), namely CETOP, TORC, and VIPRE-W, and (following requested approval) the EPRI version of VIPRE-01.

• Software change control would conform to the Software Quality Assurance (SQA) provisions of the PVNGS Quality Assurance Program Description (QAPD) which is based upon the American Society of Mechanical Engineers (ASME) NQA-1-2008 Standard, the NQA-1a-2009 Addenda, and PVNGS implementing procedures. The QAPD was reviewed and approved by NRC in a Safety Evaluation dated July 22, 2016

[Reference (25)].

Enclosure Attachment 1 NON-PROPRIETARY

• Regulatory change control would conform to the requirements of 10 CFR 50.59 as well as NRC Regulatory Guide 1.187, Revision 0, Guidance for Implementation of 10 CFR 50.59 Changes, Tests, and Experiments, as specified in Section 1.8 of the PVNGS UFSAR as discussed herein. Addition of a previously approved CHF correlation to a previously approved thermal-hydraulic code would be treated as a change in an element of methodology, and would be acceptable provided that analytical results were conservative with respect to, or essentially the same as, those that the NRC had previously reviewed and approved. Thus, use of a previously approved CHF correlation into a different computer code would not adversely affect the DNBR safety limit in PVNGS Technical Specification 2.0, Safety Limits, nor would it adversely affect the DNBR statistics that are used in the licensing basis statistical convolution methodology that is used to calculate fuel failure for certain UFSAR Chapter 15 safety analyses.

• APS would docket a notification letter to the NRC prior to using a new correlation and code combination in PVNGS regulatory applications. APS would make available for NRC inspection, audit, or review any supporting data or information should the NRC staff have any questions following notification, such as occurred following the December 20, 2001, APS notification letter regarding plans for implementation of the CENTS transient analysis computer code [Reference (18)].

With respect to validation, verification, and sensitivity studies, the PVNGS software modification process would be essentially the same whether a CHF correlation was directly inserted into the source code for a thermal-hydraulic code, or coupled with a thermal-hydraulic code through the use of external files that are called during code execution. PVNGS SQA program implementing procedures require that software used for safety related licensing applications be supported by quality records such as:

Software Configuration Control Form Software Requirements Specification Software Design Description Software User Manual Verification & Validation Plan Verification & Validation Test Results Software Installation Instructions Software Installation Test Results APS recognizes that there are three primary sources of uncertainty in thermal-hydraulic analyses which must be considered during verification and validation activities, specifically the following:

• Input parameter uncertainties due to numerical solution approximations to the governing conservation equations, which may vary from one code to another. This has implications for certain code inputs such as the selection of mesh sizes and the values of transverse momentum parameters.

Enclosure Attachment 1 NON-PROPRIETARY

• System parameter uncertainties due to fuel fabrication processes and engineering factors. This has implications for the selection of input values related to fuel rod pitch, rod bowing, cladding diameter, and the engineering factor on heat flux.

• CHF correlation uncertainties, as determined from statistical analysis of test data.

Because the calculation of local coolant conditions may vary slightly from one code to another, this has implications for the comparison of predicted and measured values for parameters used in any particular CHF correlation.

Thus, APS verification and validation practices would include consideration of the following, to the extent applicable to each software modification:

• Review of topical reports and NRC Safety Evaluations related to review and approval of a particular CHF correlation, for example, the multiple topical reports related to approval of the CE-1 correlation with the TORC and CETOP-D codes. Identify how test data was evaluated and used in the development of that correlation. Also identify benchmarking and sensitivity studies that were performed, such as evaluations of mesh size selection on analytical results.

• Performance with respect to the data used in development of the particular CHF correlation. This may include consideration of poolability of data, outliers, normality distributions, comparison of various data groups, homogeneity of variance, correlation databases, validation databases, etc., as was previously reviewed and approved by the NRC. Methods may include the generation of scatter plots for each variable in the correlation, comparison of measured and predicted values as a function of correlation variables, and examination of same to identify regions of sparse data or nonconservative regions. Where statistical tests are necessary, APS would use methods that have previously been accepted by NRC staff (for example, NUREG-1475

[Reference (26)] or National Bureau of Standards Handbook 91 [Reference (27)].

• Performing code-to-code comparisons across the entire range of applicability for the particular CHF correlation. This could include, but not necessarily be limited to, benchmarking and sensitivity studies that were described in topical reports for another code and approved by the NRC. For example, examining mesh size selection for CE-1 with the VIPRE code in a manner similar to that previously performed with TORC and CETOP.

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 6 Section 1.3 of Attachment 10 (proprietary) of the application states, in part, The current APS Core Thermal Hydraulic Analysis scope will be modified to the extent that llfor compliance with Technical Specification limits on DNBR.

The specifics of this modification process in Section 5 of Attachment 10 (proprietary) do not appear to be provided. Please provide a detailed explanation of which computer codes were needed to be modified by the addition of BHTP and how that modification took place.

Response to NRC Question 6 The computer codes that utilize the BHTP CHF correlation are VIPRE-01 and VIPRE-W. The following addresses how these computer codes were modified to facilitate their use of the BHTP CHF correlation.

EPRI VIPRE-01 EPRI VIPRE-01 has the capability for adding user-defined CHF correlations by calling an external subroutine/function without modifying the VIPRE-01 code itself; therefore, no VIPRE-01 code changes were necessary.

To utilize the Framatome BHTP CHF correlation, Zachry Nuclear Engineering (Zachry; the organization which develops and maintains VIPRE-01) was contracted by APS to design an external CHF subroutine from the original Framatome specification presented in Topical Report BAW-10241(P)(A) [References (7) and (8)]. This work was performed under the Zachry 10 CFR Part 50 Appendix B SQA program. This work included Software Requirements Specification, Software Design Description, and Verification & Validation (V&V) documentation. A calculation was also originated to determine the BHTP DNBR correlation limit at the 95/95 level using the method of Reference (28).

Testing and qualification of VIPRE-01 with the BHTP external CHF subroutine utilized the CHF data points collected from Columbia University high pressure test loop assembly experiments as discussed in Reference (7). The VIPRE-01 BHTP CHF test matrix was generated using test specific operating conditions (for example, inlet mass flux, enthalpy, exit pressure, bundle power, and power distributions). Predicted CHF data was extracted for a known rod of interest from VIPRE-01 and compared to the measured values of Reference (7). The VIPRE-01 predictions are shown in Figure 5-1 and Table 5-4 of Attachment 10 of the APS application

[Reference (3)].

The BHTP external CHF subroutine for VIPRE-01 was implemented on the APS computer system via interfacing software that accounted for changes necessary to address the differences in computing platforms (the BHTP external CHF subroutine provided by Zachry remains unchanged). The VIPRE-01 BHTP external CHF subroutine and interfacing software were qualified for use at APS under the PVNGS SQA program and procedures. A description of

Enclosure Attachment 1 NON-PROPRIETARY the documentation required by the PVNGS SQA program is provided in the Response to NRC Question 5" in this enclosure.

Westinghouse VIPRE-W VIPRE-W was modified by Westinghouse, under their 10 CFR Part 50 Appendix B SQA program, to add the capability to call an external subroutine/function for a user-defined CHF correlation.

The previously described BFITP external CFIF subroutine software for VIPRE-01 was used with VIPRE-W via interfacing software. This interfacing software addressed the differences in computing platforms, the differences between the VIPRE-W and VIPRE-01 functional interface, and the differences in how design-specific data is read by the two codes. These interfacing software changes did not require any changes to the BFITP external CHF subroutine provided by Zachry. VIPRE-W with its call to the BHTP external CHF subroutine and interfacing software were qualified for use at APS under the PVNGS SQA program and procedures.

Testing and qualification of VIPRE-W with the BHTP external CHF subroutine utilized the CHF data points collected from Columbia University high pressure test loop assembly experiments as discussed in Reference (7). The VIPRE-W BHTP CHF test matrix was generated to be as similar as practical to the VIPRE-01 model. It was then executed using the same test specific operating conditions as for VIPRE-01. Predicted CHF data was extracted for a known rod of interest from VIPRE-W and compared to the measured values of Reference (7).

Section 5.4.2 of Attachment 10 of the APS application [Reference (3)] shows the overall average ratio of the predicted VIPRE-W/BHTP to the predicted VIPRE-01/BHTP CHF is small.

Given the small differences between VIPRE-W/BHTP and VIPRE-01/BHTP CHF results, VIPRE-W with BHTP is benchmarked and acceptable for use.

Based upon the design of the VIPRE-01 and VIPRE-W software and the qualified BHTP correlation, and the successful comparison to experimental data, it is shown either code with the BHTP CHF correlation may be used to model Framatome CE16HTP fuel.

Enclosure Attachment 1 NON-PROPRIETARY NRC Question 7 Section 11 of Attachment 10 (proprietary) of the application provides a short summary of COLSS/CPCS setpoints analysis that is unique for some CE plants. This section provides a very limited description of the setpoints analysis for Palo Verde cores, which use Framatome CE 16x16 HTP fuel.

a. Please provide justification for the CE setpoints methodology applicability to Framatome fuel.

b. Please provide details of how the CE setpoints methodology is applied to the cores with Framatome fuel design.

Response to NRC Question 7 The following response provides an overview of the standard setpoints methodology, the augmented methodology to support Westinghouse Next Generation Fuel (NGF), and the specific application for Framatome fuel.

Standard Setpoints Methodology The setpoints methodology as approved by the NRC in CEN-356(V)-P-A [Reference (5)] was designed to provide monitoring and protection for the Combustion Engineering digital core monitoring (COLSS) and protection system (CPCS). The setpoint methods discussed here were specifically designed for the digital Combustion Engineering plant types.

The reload analyses are performed in series. First the physical core is modeled using the CASMO/SIMULATE neutronics core package to generate input into the COLSS/CPCS methodology.

Then the CETOP thermal-hydraulic code is used to generate conservative DNBR calculations for((j^ It should be noted that the CETOP model used within setpoints ((l IB)]- combination of the neutronics and thermal hydraulics forms the ((l D] in the COLSS/CPCS Overall Uncertainty Analysis (OUA).

The COLSS and CPCS algorithms at the plant have simplified to generate DNBR and Linear Heat Rate (LHR)ZLinear Power Density (LPD) calculations that form the basis of the monitoring alarm and trip setpoints at the plant. The setpoint methodology utilizes to represent the actual software installed at the plant.

The COLSS/CPCS setpoint analysis performs two major functions. The first function is to tor the The second function is to perform the COLSS/CPCS OUA to provide the final overall COLSS and CPCS uncertainty factors to ensure that COLSS/CPCS DNBR and LHR/LPD calculations are conservative with a 95% probability and a 95% confidence level. These uncertainty factors are then installed as addressable constants into the plant computers.

Enclosure Attachment 1 NON-PROPRIETARY Setpoint Methodology Augmentation for Westinghouse NGF The NGF fuel design has multiple spacer grid types and CHF correlations. For the NGF design, Westinghouse proposed, and the NRC accepted, an augmented setpoint methodology

[Reference (6)] that includes additional process steps to incorporate separate temperature, pressure, flow, and Axial Shape Index (ASI) dependent biases into the DNBR uncertainty factors. Additionally, the augmented process incorporated the in the core thermal-hydraulic analyses.

Justification for Using Setpoint Methodology for Framatome Fuel The standard setpoints methodology applies to plants with COLSS/CPCS; it is not fuel type nor is it fuel vendor dependent. The Framatome fuel design is analyzed in the upstream core neutronics, core thermal-hydraulics, and fuel behavior analyses. The results of these analyses are input into the COLSS/CPCS setpoint methodology and OUA process to determine conservative DNBR and LHR/LPD uncertainty factors with at least a 95% probability and a 95%

confidence level for implementation into the COLSS and CPCS in the plant.

Application of Setpoint Methodology to Framatome Fuel As addressed immediately above, the COLSS/CPCS setpoints methodology is not fuel type dependent; requiring no changes to methodology, only changes to the OUA inputs for Framatome fuel.

The relevant Framatome fuel geometry parameters are the same as CE16STD fuel. The COLSS/CPCS OUA inputs that are fuel dependent (for example, rod bow factors, core neutronics design, thermal-hydraulic design, and fuel performance analyses) are updated to reflect the fuel type prior to use in the COLSS/CPCS OUA. These inputs to the OUA are updated for the fuel following the process allowed for those analyses.

With the original setpoints process, the CETOP adjustment factors generated within the thermal-hydraulic benchmark analysis account for the With the augmented setpoints process, the added setpoint steps would account for axial CHF correlation differences within the fuel assembly.

APS application of the COLSS/CPCS setpoints methodology to Framatome fuel is as follows:

• For full cores with Framatome fuel: Since Framatome fuel does not have multiple CHF correlations along the axial length of the fuel assembly (except for the bottom portion of the fuel assembly where DNB is not significant) the original setpoints methodology can be used. Use of the augmented process as described for NGF fuel may also be used.

• For mixed cores with CE16STD and Framatome fuel: Since both CE16STD and Framatome fuel do not have multiple CHF correlations along the axial length of the fuel assembly (except for the bottom portion of the fuel assembly where DNB is not

Enclosure Attachment 1 NON-PROPRIETARY significant) the original setpoints methodology can be used. Use of the augmented process as described for NGF fuel may also be used if desired.

• For mixed cores with NGF and Framatome fuel: Since NGF fuel does have multiple CHF correlations along the axial length of the fuel assembly, the augmented process as described for NGF fuel will be used for the entire core. This process will be used regardless of whether CE16STD fuel is also present in the core.

Use of the augmented setpoints process for Framatome fuel requires that the CETOP code have the Framatome BHTP CHF correlation implemented. The addition of the BHTP CHF correlation to the CETOP code would follow the process outlined in Section 5.4.3 of Attachment 10 of the APS application [Reference (3)] and discussed in the Response to NRC Question 5 in this enclosure.

Enclosure Attachment 1 NON-PROPRIETARY References

1. Letter from T. H. Essig (NRC) to H. Sepp (Westinghouse). Acceptance for Referencing of Licensing Topical Report WCAP-14565, VIPRE-01 Modeling and Qualification for Pressurized Water Reactor Non-LOCA Thermal-Hydraulic Safety Analysis" (TAC No.

M98666). January 19, 1999. [NRC ADAMS Accession No. ML993160096].

2. Letter from S. P. Lingam (NRC) to R. S. Bement (APS). Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments to Revise Technical Specifications to Support the Implementation of Next Generation Fuel (CAC Nos. MF8076, MF8077, and MF8078; EPID L-2016-LLA-0005). January 23, 2018. Amendment No. 205 to Renewed Facility Operating License No. NPF-41, Amendment No. 205 to Renewed Facility Operating License No. NPF-51, and Amendment No. 205 to Renewed Facility Operating License No. NPF-74. [NRC ADAMS Accession No. ML17319A107].

3. Letter No. 102-07727-MLL/SMM from M. L. Lacal (APS) to Document Control Desk (NRC). Palo Verde Nuclear Generating Station, Units 1, 2, and 3, Docket Nos. STN 50-528, 50-529, and 50-530, License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel. July 6, 2018.

[NRC ADAMS Accession No. ML18187A417].

4. Electric Power Research Institute. NP-2511-CCM-A. VIPRE-01: A Thermal-Hydraulic Code for Reactor Cores. February 2017. Volumes 1 through 5.

5. Combustion Engineering, Inc. CEN-356(V)-P-A, Revision 01-P-A. Modified Statistical Combination of Uncertainties. May 1988.

6. Letter from T. B. Blount (NRC) to J. A. Gresham (Westinghouse). Final Safety Evaluation for Westinghouse Electric Company Addendum 1 to Topical Report WCAP-16500-P, Supplement 1, Revision 1, "Application ofCE Setpoint Methodology for CE 16x16 Next Generation Fuel" (TAC No. ME3583). July 1, 2010. [NRC ADAMS Accession No. ML101720184].

7. Letter from H. N. Berkow (NRC) to J. F. Mallay (Framatome). Final Safety Evaluation for Topical Report BAW-10241-P, Revision 0, "BHTP Correlation Applied in LYNXT" (TAC No. MB7033). September 29, 2004. [NRC ADAMS Accession No. ML042730472].

8. Letter from H. N. Berkow (NRC) to R. L. Gardner (Framatome). Final Safety Evaluation for Framatome ANP (FANP), Appendix A to Topical Report (TR) BAW-10241(P), Revision 1, "Extension of the BHTP CHF [Critical Heat Flux] Correlation Ranges" (TAC No. MC6374).

July 25, 2005. [NRC ADAMS Accession No. ML052070383].

9. Letter from B. Pham (NRC) to G. R. Overbeck (APS). Paio Verde Nuclear Generating Station, Unit 2 (PVNGS-2) - Issuance of Amendment on Replacement of Steam Generators and Uprated Power Operations (TAC No. MB3696). September 29, 2003.

Amendment No. 149 to Facility Operating License No. NPF-51. [NRC ADAMS Accession No. ML032720538].

10. Letter from M. B. Fields (NRC) to J. M. Levine (APS). Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Issuance of Amendments Re: Replacement of Steam Generators and Uprated Power Operations and Associated Administrative Changes (TAC Nos. MC3777, MC3778, and MC3779). November 16, 2005. Amendment No. 157 to

Enclosure Attachment 1 NON-PROPRIETARY Facility Operating License No. NPF-41, Amendment No. 157 to Facility Operating License No. NPF-51, and Amendment No. 157 to Facility Operating License No. NPF-74. [NRC ADAMS Accession No. ML053130275].

11. NUREG-0800. Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition. Section 4.2, "Fuel System Design," Revision 3.

March 2007. [NRC ADAMS Accession No. ML070740002].

12. Letter No. 102-04847-CDM/TNW/RAB from C. D. Mauldin (APS) to Document Control Desk (NRC). Palo Verde Nuclear Generating Station (PVNGS), Unit 2, Docket No. STN 50-529, Response to Request for Additional Information Regarding Steam Generator Replacement and Power Uprate License Amendment Request. October 11, 2002. [NRC ADAMS Accession No. ML022940385].

13. Letter from S. Dembek (NRC) to J. Mallay (Framatome). Framatome ANP Topical Report BAW-10231, "COPERNIC Fuel Rod Design Computer Code" - Correction ofErrorin Safety Evaluation (TAC No. MA6792). June 14, 2002. [NRC ADAMS Accession No. ML021360461].

14. Combustion Engineering, Inc. C-E Methods for Loss of Flow Analysis. June 1984.

15. NUREG-0857. Safety Evaluation Report Related to the Operation of PVNGS Units 1, 2, and 3, Docket Nos. STN 50-528, STN 50-529, and STN 50-530. Initial issue (November 1981) through Supplement No. 12 (November 1987).

16. NUREG-0852. Safety Evaluation Report Related to the Final Design of the Standard Nuclear Steam Supply Reference System CESSAR System 80. Initial issue (November 1981) through Supplement No. 3 (December 1987).

17. Letter from L. R. Wharton (NRC) to G. R. Overbeck (APS). Palo Verde Nuclear Generating Station, Units 1, 2, and 3- Issuance of Amendments Re: Various Administrative Controls (TAC Nos. MB1668, MB1669, andMB1670). October 15, 2001.

Amendment No. 137 to Facility Operating License No. NPF-41, Amendment No. 137 to Facility Operating License No. NPF-51, and Amendment No. 137 to Facility Operating License No. NPF-74. [NRC ADAMS Accession No. ML012880473].

18. Letter No. 102-04639-CDM/TNW/JAP from C. D. Mauldin (APS) to Document Control Desk (NRC). Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3, Docket Nos. STN 50-528/529/530, Notification of Intent to Use CENTS Code. December 20, 2001. [NRC ADAMS Accession No. ML020240434].

19. Westinghouse Electric Company LLC. WCAP-15996-P-A, Revision 1 (CENPD-282-P-A, Revision 2). Technical Description Manual for the CENTS Code. November 2005.

Volumes 1 through 4. [NRC ADAMS Accession No. ML053290344].

20. Letter from H. N. Berkow (NRC) to G. Bishoff (Westinghouse). Final Safety Evaluation for Topical Report WCAP-15996-P, "Technical Description Manual for the CENTS Code" (TAC No. MB6982). December 1, 2003. [NRC ADAMS Accession No. ML032790634].

21. Letter from H. N. Berkow (NRC) to G. Bischoff (Westinghouse). Final Safety Evaluation for Topical Report WCAP-15996-P, "Technical Description Manual for CENTS Code" (TAC No. MB6982). November 24, 2004. [NRC ADAMS Accession No. ML043270382].

Enclosure Attachment 1 NON-PROPRIETARY

22. Letter from S. A. Richards (NRC) to P. W. Richardson (Westinghouse). Safety Evaluation of Topical Report CENPD-404-P, Revision 0, "Implementation ofZIRLO' Material Cladding in CE Nuclear Power Fuel Assembly Designs" (TAC No. MB1035). September 12, 2001. [NRC ADAMS Accession No. ML012670041].

23. Letter from H. N. Berkow (NRC) to J. A. Gresham (Westinghouse). Final Safety Evaluation for Addendum 1 to Topical Report WCAP-12610-P-A and CENPD-404-P-A, "Optimized ZIRLO^**," (TAC No. MB8041). June 10, 2005. [NRC ADAMS Accession No. ML051670403].

24. Combustion Engineering, Inc. CENPD-188-A. HERMITE: A Multi-Dimensional Space-Time Kinetics Code forPWR Transients. March 1976.

25. Letter from S. P. Lingam (NRC) to R. K. Edington (APS). Palo Verde Nuclear Generating Station, Units 1, 2, and 3 - Request to Change the Quality Assurance Program Description (CAC Nos. MF6537, MF6538, and MF6539). July 22, 2016. [NRC ADAMS Accession No. ML16194A323].

26. NUREG-1475. Applying Statistics. March 2011. Revision 1. [NRC ADAMS Accession No. ML11102A076].

27. Mary Gibbons Natrella. Experimental Statistics. 1963. Reprinted 1966. National Bureau of Standards Handbook 91.

28. Sommerville, P. N. Tables for Obtaining Non-Parametric Tolerance Limits. Annals of Mathematical Statistics, Vol. 29, No. 2, pp. 599-601. June 1958.

ATTACHMENT 2 Affidavit from Arizona Public Service Company Submitted in Accordance with 10 CFR 2.390 to Consider Attachment 3 as a Proprietary Document

AFFIDAVIT STATE OF ARIZONA )

)ss.

CITY OF PHOENIX )

1. My name is Bruce Rash. I am employed by Arizona Public Service Company ("APS"). My present capacity is Vice President, Nuclear Engineering, for the Palo Verde Nuclear Generating Station ("PVNGS"), and in that capacity I am authorized to execute this Affidavit.

2. APS is the operating agent for PVNGS. I am familiar with the policies established by APS to determine whether certain APS information is proprietary and confidential, and to ensure the proper application of these policies.

3. I am familiar with APS information in the following document: Attachment 3 to the enclosure for APS Correspondence 102-07807, Supplemental Information Regarding License Amendment Request and Exemption Request to Support the Implementation of Framatome High Thermal Performance Fuel, referred to herein as "Document." Information contained in this Document has been classified by APS as proprietary in accordance with the policies established by APS for the control and protection of proprietary and confidential information.

4. The information contained in this Document is proprietary and confidential in natures and of the type customarily held in confidence by Framatome (formerly Areva, Inc.),

Westinghouse, and APS, and not made available to the public. Based on my experience in the nuclear industry, I am aware that other companies also regard the type of information contained in the Document as proprietary and confidential.

5. This Document has been made available to the U.S. Nuclear Regulatory Commission in confidence with the request that the information contained in this Document be withheld from public disclosure. The request for withholding proprietary information from public

disclosure is made in accordance with 10 CFR 2,390. The information qualifies for withholding from public disclosure under 10 CFR 2.390(a)(4) "Trade secrets and commercial or financial information."

6. APS applied the following criteria to determine that the information contained in the Document should be classified as proprietary and confidential; (a) APS has a non-disclosure agreement with Westinghouse Electric Company LLC

("Westinghouse"), Framatome, and Structural Integrity Associates. Inc. (SI) (the NDA is referred to as the "Westinghouse-AREVA-SI-APS NDA"), under which Westinghouse and Framatome have provided to APS certain proprietary and confidential information contained in the Document.

(b) The information reveals details of Westinghouse's, APS's, and/or Framatome's research and development plans and programs, or the results of these plans and programs.

(c) The information includes test data or analytical techniques concerning a process, methodology, or component, the application of which results in a competitive commercial advantage for Westinghouse, APS, and/or Framatome.

(d) The information reveals certain distinguishing aspects of a process, methodology, or component, the exclusive use of which provides a competitive commercial advantage for Westinghouse, APS, and/or Framatome on product optimization or marketability.

(e) The unauthorized use of the information by one of Westinghouse's, APS's, and/or Framatome's competitors would permit the offending party to significantly reduce its expenditures, in time or resources, to design, produce, or market a similar product or service.

(f) The information contained in the Document is vital to a competitive commercial advantage held by Westinghouse, APS. and/or Framatome, would be helpful to

their competitors, and would likely cause substantial harm to the competitive position of Westinghouse, APS, and/or Framatome.

(g) It reveals aspects of past, present, or future Westinghouse, Framatome, or APS funded development plans and programs of potential commercial value.

7. In accordance with APS's policies governing the protection and control of proprietary and confidential information, the information contained in this Document has been made available, on a limited basis, to others outside APS only as required and under suitable agreement providing for nondisclosure and limited use of the information.

9. APS's policies require that proprietary and confidential information be kept in a secured file or area and distributed on a need-to-know basis. The information contained in the Document has been kept in accordance with these policies.

10. The foregoing statements are true and correct to the best of my knowledge, information, and belief, and if called as a witness I would competently testify thereto. I declare under penalty of perjury under the laws of the State of Arizona that the above is true and correct.

Bruce Rash SUBSCRIBED before me this 1 day of O _ 2018.

.a NOTARY PUBLIC, STATE OF ARIZONA MY COMMISSION EXPIRES:

Reg. #: donna NORMAN MyOaamMonbM AuguN<<.a<<1