ML19235A256
| ML19235A256 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 09/09/2019 |
| From: | Siva Lingam Plant Licensing Branch IV |
| To: | Bement R Arizona Public Service Co |
| Lingam S, 301-415-1564 | |
| References | |
| EPID L-2018-LLA-0194, EPID L-2018-LLE-0010 | |
| Download: ML19235A256 (21) | |
Text
September 9, 2019 Mr. Robert S. Bement Executive Vice President Nuclear/
Chief Nuclear Officer Mail Station 7602 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034
SUBJECT:
PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 - REGULATORY AUDIT
SUMMARY
FOR THE JUNE 17-20, 2019, AUDIT FOR THE LICENSE AMENDMENT AND EXEMPTION REQUESTS ASSOCIATED WITH FRAMATOME HIGH THERMAL PERFORMANCE FUEL (EPID L-2018-LLA-0194 AND EPID L-2018-LLE-0010)
Dear Mr. Bement:
By letter dated July 6, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18187A417), as supplemented by letters dated October 18, 2018, March 1, 2019, and May 17, 2019 (ADAMS Accession Nos. ML18296A466, ML19060A298, and ML19137A118, respectively), Arizona Public Service Company (the licensee) requested changes to the technical specifications (TSs) to support the implementation of Framatome Advanced Combustion Engineering 16x16 High Thermal Performance fuel design with M5 as a fuel rod cladding material and gadolinia as a burnable absorber for Palo Verde Nuclear Generating Station (Palo Verde), Units 1, 2, and 3. In addition to this license amendment request, the licensee is requesting an exemption from certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, Acceptance criteria for emergency core cooling systems [ECCS] for light-water nuclear power reactors, and 10 CFR Part 50, Appendix K, ECCS Evaluation Models, to allow the use of Framatome M5 alloy as a fuel cladding material. In addition, the proposed amendments would revise TS 2.1.1, Reactor Core SLs [Safety Limits]; TS 4.2.1, Fuel Assemblies; and TS 5.6.5, Core Operating Limits Report (COLR).
The proposed amendments would adapt the approved Palo Verde reload analysis methodology to address both Westinghouse and Framatome fuel, including the implementation of Framatome methodologies, parameters and correlations. The ability to use either Westinghouse or Framatome fuel will ensure security of the Palo Verde fuel supply by providing for multiple fuel vendors with reliable fuel designs and geographically diverse manufacturing facilities.
To discuss the outstanding issues from the first audit held on January 22-23, 2019, and to review the calculations and other supporting documentation supporting the license amendment and exemption submittals, the U.S. Nuclear Regulatory Commission staff conducted a second audit at the Palo Verde site in Wintersburg, Arizona, from June 17-20, 2019. The regulatory audit summary is enclosed with this letter.
R. Bement If you have any questions, please contact me at 301-415-1564 or via e-mail at Siva.Lingam@nrc.gov.
Sincerely,
/RA/
Siva P. Lingam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-528, STN 50-529, and STN 50-530
Enclosure:
Audit Summary cc: Listserv
REGULATORY AUDIT
SUMMARY
PERFORMED AT PALO VERDE SITE ON JUNE 17-20, 2019 IN SUPPORT OF THE FRAMATOME HIGH THERMAL PERFORMANCE FUEL LICENSE AMENDMENT AND EXEMPTION ARIZONA PUBLIC SERVICE COMPANY PALO VERDE NUCLEAR GENERATING STATION, UNITS 1, 2, AND 3 DOCKET NOS. 50-528, 50-529, AND 50-530
1.0 BACKGROUND
By letter dated July 6, 2018 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML18187A417), as supplemented by letters dated October 18, 2018, March 1, 2019, and May 17, 2019 (ADAMS Accession Nos. ML18296A466, ML19060A298, and ML19137A118, respectively), Arizona Public Service Company (APS, the licensee) requested changes to the technical specifications (TSs) to support the implementation of Framatome Advanced Combustion Engineering 16x16 High Thermal Performance (HTPTM) fuel design with M5 as a fuel rod cladding material and gadolinia as a burnable absorber for Palo Verde Nuclear Generating Station (Palo Verde or PVNGS), Units 1, 2, and 3. In addition to this license amendment request (LAR), APS is requesting an exemption from certain requirements of Title 10 of the Code of Federal Regulations (10 CFR) Section 50.46, Acceptance criteria for emergency core cooling systems [ECCS] for light-water nuclear power reactors, and 10 CFR Part 50, Appendix K, ECCS Evaluation Models, to allow the use of Framatome M5 alloy as a fuel cladding material. In addition, the proposed amendments would revise TS 2.1.1, Reactor Core SLs [Safety Limits]; TS 4.2.1, Fuel Assemblies; and TS 5.6.5, Core Operating Limits Report (COLR).
The proposed amendments would adapt the approved Palo Verde reload analysis methodology to address both Westinghouse and Framatome fuel, including the implementation of Framatome methodologies, parameters and correlations. The ability to use either Westinghouse or Framatome fuel will ensure security of the Palo Verde fuel supply by providing for multiple fuel vendors with reliable fuel designs and geographically diverse manufacturing facilities.
To discuss the outstanding issues from the first audit held on January 22-23, 2019, and to review the calculations and other supporting documentation supporting the license amendment and exemption submittals, the U.S. Nuclear Regulatory Commission (NRC) staff conducted a second audit at the Palo Verde site in Wintersburg, Arizona, from June 17-20, 2019. This document provides a report on the deliberations during the second regulatory audit.
2.0 SCOPE AND PURPOSE The audit was held on June 17-20, 2019, at the Palo Verde site in Wintersburg, Arizona, and was conducted in accordance with the audit plan provided to the licensee (ADAMS Accession No. ML19154A469). The purpose of the site audit was to continue discussion of outstanding issues and permit the NRC staff to perform an audit review of calculations, other documentation supporting the LAR/exemption submittals, and to help the staff prepare and finalize requests for Enclosure
additional information (RAIs) on those questions where docketed information is needed to complete the review. The details of the audit are discussed in Section 4.0 of this summary.
3.0 AUDIT TEAM The following NRC staff members participated in the site audit:
Shaun Anderson Paul Clifford Ravi Grover Joshua Kaizer John Lehning Mathew Panicker (Remote participation)
Diana Woodyatt The following APS personnel supported the audit:
Joseph Arias Walter Bodnar William Camolli Chris Cowdin Matthew Cox Michael Dilorenzo Shawn Gill John Gunn Robert Hicks Thomas Hook Philip Horspiegel Arjun Jiao Charles Karison Ryan Lane Luke McIntyre Calvin Meddings Bruce Rash Thomas Remick Dave Ricks Shawn Seegmiller Steven Sparks Carl Stephenson Bradley Sutton Patrick Trimble Thomas Weber Jenying Wu Amy Wullbrandt The following APS consultants also supported the audit:
Mark Drucker Sasan Elemadi
Dave Medek Vic Nazereth Hans Van de Berg The following Framatome personnel supported the audit:
Chris Allison Maec Dzuba Umar Faraz Lisa Gerken Nathan Hottle Greg Kessler Brett Matthews Wanda Roman Miao Sun Ryan Swanson Rick Willamson 4.0 AUDIT REPORT 4.1 Information Needs The licensee was requested to have the presentations and documents related to the areas of focus listed. The documentation was provided by presentations, documents, and calculation details. The following were the planned major areas of focus for detailed discussion and document review.
The deliberations during the audit, along with the original contents of the LAR, and the supplemental information will be used to generate RAIs to complete the comprehensive review of the license amendment and exemption requests. The licensees stated goal of the fuel transition is to increase security of fuel supply through reliable fuel designs and diverse manufacturing, to implement an improved fuel design, and to maintain existing methods to the extent practical.
4.2 Items Discussed during Audit Framatome Fuel Project Palo Verde Reload Design Process Fuel Assembly Mechanical Design Core Thermal Hydraulics Thermal-Hydraulics Code Modification Process Mixed Cores Loss-of-Coolant Accident (LOCA) Analysis Non-LOCA Analysis Setpoints Analysis Proposed TS changes
Framatome Fuel Project The licensee stated that it is planning to use Framatome Combustion Engineering (CE) 16x16 HTPTM fuel in the spring of 2020 refueling of Palo Verde, Unit 2 with 100 CE 16x16 HTPTM fuel assemblies. The LAR approval is needed for the implementation of the APS reload methodology to address the fuel transition. Two types of Westinghouse supplied fuel designs, CE-16 STD (standard or value-added) and CE-16 Next Generation Fuel (NGF) are resident in the Palo Verde core. The proposed change will support implementation of Framatome Advanced CE-16 HTPTM fuel design. With the implementation of the new Framatome fuel design, the Palo Verde core will be a mixed core which requires mechanical and thermal-hydraulic compatibility analyses and thermal-hydraulic stability analyses. Section 2 of Attachment 10 of the LAR dated July 6, 2018, provides a detailed summary of the fuel mechanical design analysis that includes mechanical compatibility, description of fuel assembly and its components, and fuel design evaluation results (Table 2-2 of the LAR dated July 6, 2018). In response to NRC staffs audit plan, the licensee led a discussion on fuel mechanical and nuclear design of the mixed core. The licensee presented documents on Mechanical Design Information for Safety and Neutronics Analyses, Fuel Design Mechanical Compatibility Evaluations (Upper Tie Plate, Lower Tie Plate, Guide Tube and Instrument Tube),
Fuel Assembly Normal Operating Analysis, Fuel Rod Cladding and Buckling Analyses for Normal Operating Condition, Copernic Based Centerline Fuel Melt and Transient Cladding Strain Analyses, Fuel Rod Cladding Stress Analysis, and Copernic Based Creep Collapse Analysis.
Palo Verde Reload Design Process The licensee intends to use NRC-approved APS reload methods to address the Framatome fuel; they are Framatome mechanical design, Framatome LOCA methods, COPERNIC code to describe fuel behavior to model thermal conductivity degradation (TCD) with exposure, NRC-approved Framatome BHTP (designation for Framatome) critical heat flux (CHF) correlation, and use of VIPRE-01 (Versatile Internals and Component Program for Reactors; Electric Power Research Institute (EPRI)) as an alternative to VIPRE-W (Versatile Internals and Component Program for Reactors; Westinghouse) code. The licensee proposed to add VIPRE-01 from EPRI to be used concurrently with VIPRE-01 from Westinghouse and proposed to implement a software quality assurance (QA) program in the implementation of different vendor codes for interchangeable licensing applications. The licensee proposed to use the two-stage version of the VIPRE code to be used in the same manner as the TORC (Thermal-hydraulics of Reactor Core) code to select the limiting assembly candidates for further evaluation. The licensee proposed to use Generic Letter 83-11, Licensee Qualification for Performing Safety Analyses in Support of Licensing Actions, and 10 CFR 50.59, Changes, tests and experiments, processes for implementing the change process for thermal-hydraulic codes. The licensee proposed to mix and match approved codes with approved correlations, such as, insert BHTP correlation and CE-1 correlations in CETOP (Combustion Engineering Thermal On-Line Program) code and BHTP code in VIPRE-01 and VIPRE-W codes.
The NRC staff engaged in discussions with the licensee regarding the Palo Verde reload methodologies. One of the issues was the statement in the LAR dated July 6, 2018:
This LAR will adapt the approved PVNGS reload analysis methodology to address both Westinghouse and Framatome fuel, including the implementation of Framatome methodologies, parameters and correlations. The ability to use either Westinghouse or Framatome fuel will ensure security of the PVNGS fuel
supply by providing for multiple fuel vendors with reliable fuel designs and geographically diverse manufacturing facilities.
The NRC staff raised concern regarding part of the above section ...the approved PVNGS reload analysis methodology to address both Westinghouse and Framatome fuel, including the implementation of Framatome methodologies..., which implies that there is an approved Palo Verde methodology for Framatome fuel designs. The staff is expecting a response to an RAI regarding this issue.
Fuel Assembly Mechanical Design The licensee reported that the new Framatome CE 16x16 HTPTM fuel design is essentially the same design as the lead test assemblies at the Palo Verde core. The licensee specified that the fuel mechanical compatibility with the resident fuel with respect to upper and lower tie plates, guide tubes, and instrumentation tubes is maintained during all operating conditions.
For a faulted condition due to earthquakes and pipe breaks, the structural response will be analyzed by NRC-approved methodology, ANP-10337P-A, PWR Fuel Assembly Structural Response to Externally Applied Dynamic Excitations, April 2018. For structural response analysis the co-resident CE-16 STD and CE-16 NGF fuel is being reviewed for the mixed core.
Thermal-Hydraulics and Modification Process, Mixed Core Staff Concerns Mechanical/Thermal-hydraulic compatibility of the three different fuel types (CE 16x16 STD, CE 16x16 NGF and CE 16x16 HTPTM) with three different cladding materials and three different CHF correlations for departure from nucleate boiling (DNB) ratio calculations were discussed.
Discussed the details of the thermal-hydraulic characterization and thermal margin analysis for the mixed core at the Palo Verde units after the potential fuel transition to Framatome fuel.
The use of approved CHF correlations in approved codes that do not have specific approval for using such correlations, thereby changes to approved topical reports, which requires more rigorous review than normal fuel transition LARs was discussed. The licensees modification of approved codes by inserting new CHF correlations (approved for different codes) (e.g., ABB-NV (Westinghouse (ABB) Non-Vane CHF Correlation) or WSSV CHF correlations (Westinghouse Side Supported Vane CHF Correlation) with VIPRE-W and VIPRE-01 codes; CE-1 correlation with VIPRE-W, CETOP-D and TORC codes for use with VIPRE-01 code; and BHTP CHF correlation in VIPRE-01 and VIPRE-W codes with CETOP-D and TORC codes) is a concern to the NRC staff.
VIPRE-W modeling options such as 1-pass and 2-pass models as described in Section 5.1 of the LAR were discussed during the audit.
Mixed Cores During the audit, the NRC staff reviewed mechanical and thermal-mechanical compatibility analyses to determine compatibility of the new fuel with the co-resident fuel. The staff found
that the analysis was inadequate, and the licensee agreed to revise the draft response for compatibility analyses.
The thermal margin assessment for the core configuration involved inlet flow distribution (IFD) with a detailed full core model, with each assembly modeled to estimate the impact on the IFD of a mixed core.
The NRC staff is reviewing the CETOP-D/VIPRE benchmark calculations for CETOP correction factors for each fuel type quarter assembly that includes all three fuel types and different spacer grid loss coefficients.
Seismic Analysis for Mixed Core The NRC staff had discussions on how the faulted condition analysis that evaluates structural response of the fuel assembly to externally applied forces such as earthquakes and postulated pipe breaks based on the criteria established in the recently approved topical report ANP-10377P-A for the mixed core.
The NRC staff had discussions on how the seismic evaluation will be done for Framatome and Westinghouse/CE fuel designs in the Palo Verde core.
Based on the discussions with the licensee, the NRC staff has revised the RAIs for seismic-related analysis of the Framatome, Westinghouse and mixed cores.
Setpoints Analysis The NRC staff has concerns with the COLSS (core operating limits supervisory system)/CPCS (core protection calculator system) setpoints process briefly mentioned in the LAR, dated July 6, 2018, and with the further detail provided in response to NRC Question No. 7 in the licensees supplemental letter dated October 18, 2018. The LAR dated July 6, 2018, states:
Therefore, the MSCU [Modified Statistical Combination of Uncertainties]
methodology as described in CEN-356(V)-P-A, Revision 01-P-A, Modified Statistical Combination of Uncertainties, (Reference 11.1) is applicable to setpoint analysis of the CE16HTP fuel.
The LAR dated July 6, 2018, also states:
An available alternative for implementation of Framatome CE16 HTP fuel is to use the MSCU process steps as augmented by WCAP-16500-P-A, Supplement 1, Revision 1, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF), (Reference 11.2).
The supplemental information in the letter dated October 18, 2018, includes the following text:
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 CE16 STD and Framatome fuel: Since both CE16 STD 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 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.
The NRC staff concerns are that (1) neither the original MSCU process nor the augmented MSCU process is directly applicable to CE16 HTPTM cores and (2) it has not yet been demonstrated that applying either of these methods will produce acceptable results.
During the audit, the NRC staff reviewed APS calculation RA-13-C00-2019-016, Revision 00, Framatome VQP [Vendor Qualification Program]: Mixed Core Master Setpoint Overall Uncertainty Analysis Demonstration. This calculation provides an example COLSS and CPCS setpoints analysis for a mixed NGF and HTPTM core. The stated goal of this calculation is to show that the existing NRC-approved COLSS and CPCS overall uncertainty analysis (OUA) methodology including the WCAP-16500 Supplement 1 Revision 1, eight-step process can be applied to both the NGF and HTPTM fuel types for the purpose of generating COLSS and CPC OUA Uncertainty Factors. Note that the staff confirmed that the APS controlled version of WCAP-16500 (N001-0205-00063) included Addendum 1, which justified removal of a 3.0 percent penalty from the S1R1 version of this topical report.
The NRC staffs detailed review of the demonstration setpoints analysis yielded the following observations:
- 1. The original MSCU methodology is not applicable to CE-16 HTPTM fuel and will likely produce non-conservative setpoints.
- 2. Strict application of the modified MSCU methodology may produce non-conservative setpoints. These methods are sensitive to axial shape index breakpoints, which are fuel design-specific.
- 3. Application of the modified MSCU methodology to mixed core configurations results in a significant loss of DNB thermal margin (relative to a full core of either STD, NGF, or HTPTM) due to worse case statistics and/or application of limiting ranges and adjustments.
Because of the NRC staffs review, the demonstration analysis will be revised. Furthermore, the LAR and RAI response defining the setpoints methods and their applicability to CE-16 HTPTM needs to be modified. APS will need to clearly define the setpoints process (i.e., a revised MSCU methodology) to be used for reload core containing CE16 HTPTM fuel assemblies. The staff will need to review this new methodology.
Limiting Fuel Types in Mixed Core During the audit, NRC and APS staff discussed the process for assessing fuel rod failures during postulated accidents with mixed core configurations. It was not clear to the NRC staff that the proposed method would capture fuel previously labelled as non-limiting. For example, the core thermal-hydraulic analyses may define the NGF fuel bundles as non-limiting in a core containing most of lower exposure HTP fuel bundles. Because of this designation, DNB fuel failure probability distribution functions would not be created for the NGF bundle. The NRC staff concerns were that under accident conditions, any fuel bundle has the potential to experience fuel rod failure. As such, predicted fuel rod failures, which ignore populations of fuel bundles may be non-conservative. Because of these discussions, APS agreed to modify its methods and assess all bundle types.
Control Rod Ejection During the audit, the NRC staff reviewed the control element assembly (CEA) ejection transient calculation TA-13-C00-2018-207, Revision 2. The limiting ejected rod worth and peaking are dominated by centerline melt restriction. The Framatome fuel melting temperature of 4,843 degrees Fahrenheit was used to back-calculated maximum allowable ejected worth. The corresponding maximum radial average fuel enthalpy was always less than 160 calories/gram (cal/g), well below 230 interim limit or earlier 280 cal/g limit. In response to an earlier RAI, APS committed to adhere to the interim RAI criteria and guidance. However, STRIKIN-II does not appear to be capable of predicting pellet radial power and temperature distributions during the transient. Based upon discussions with APS, STRIKIN-II may assume a uniform power profile radially across the pellet. STRIKIN-II appears to calculate a parabolic temperature profile.
Hence, it is not clear how STRIKIN-II can be employed to satisfy local fuel melting restrictions.
During the audit, APS informed the NRC staff that it had contracted Framatome to complete a more detailed CEA analysis using the recently approved AREA 3D methodology. The evaluation included a mixed CE-16 STD and CE-16 HTPTM core configuration. During the audit, the staff reviewed ANP-3785P, Revision 0. It is unclear if this analysis was reviewed in accordance with Framatome QA procedures or had been acceptance reviewed in accordance with Palo Verde procedures.
While this detailed 3D analysis is not being relied upon for reaching a safety finding, the results suggest that existing legacy APS rod ejection methods are overly conservative. This comparison is consistent with the NRC staffs assessment of 3D Purdue Advanced Reactor Core Simulator control rod ejection predictions and the overall conclusions within Research Information Letter 0401. The Framatome detailed assessment evaluated multiple power levels and time-in-life. The maximum radial average fuel enthalpy rise remained well below cladding failure thresholds defined in the interim guidance. In addition, the maximum fuel temperature experienced during the transient remained well below melting conditions.
DNB Propagation APS methodology employs old CE methods, involving bounding INTEG calculations and geometric considerations. Analysis of record results support maximum allowed time-in-DNB of less than 4.5 seconds and a maximum calculated strain of 29.3 percent. During the audit, the NRC staff asked APS if they wanted NRC approval to (1) expand the applicability of the 4.5 second time-in-DNB (29.3 percent strain) screening criteria to M5 cladding (referred to as little m) or (2) expand the applicability of the DNB propagation INTEG code and methods for
M5 cladding (referred to as big M). The scope of the NRC review effort is significantly higher for the big M approach. APS was unsure and will discuss with staff at a later date.
To remove this decision from the critical path, the NRC staff suggested that APS could demonstrate that rod internal pressure remains below transient system pressure for all rods predicted to experience DNB during all postulated accidents. Hence, DNB propagation would not be a concern.
ACTION: Awaiting APS decision on path forward.
Non-LOCA Transients The NRC staffs audit in this area focused upon the following topics, also addressed within a set of draft RAIs provided to the licensee prior to the site audit:
- 1. Calculational details of Palo Verde Loss of Flow Analysis.
- 2. Demonstration that fuel centerline melt temperature will not be exceeded, considering inadequacy of 21 kilowatts/foot limit at certain times in cycle life and simultaneous separate burnup dependent limits for Westinghouse and Framatome fuel.
- 3. Understanding of which transients utilize input from the COPERNIC computer code.
- 4. Basis for use of the convolution method for Framatome fuel.
- 5. Understanding of DNB probability distribution function for Framatome fuel and how it impacts applicable transient analysis.
- 6. Basis for continued use of DNB propagation analysis including evidence of strain behavior for Framatome fuel.
- 7. Evidence of comparisons for M5 cladding to Zircaloy-4 alloy discussed in Section 6.4 of Attachment 8 of the LAR dated July 6, 2018.
- 8. Understanding of how Framatome fuel parameters (Hgap (fuel-to-clad gap coefficient of conductance), Gadolinia effects) are accounted for in CENTS and HERMITE codes.
- 9. Understanding of how updated guidance on reactivity-initiated accidents will be addressed in the Updated Final Safety Analysis Report (UFSAR) following the fuel transition.
Non-LOCA-related documents audited by the NRC staff are listed Section 4.3 of this document.
Discussions of the above listed topics provided the licensee a deeper understanding of the information which the staff is requesting to complete the safety evaluation. Additionally, the NRC staff was able to gain insight into the licensees intended approach to respond to RAIs in the relevant topic areas. Discussion of these topics in greater detail is anticipated in the NRC staffs RAIs, the licensees RAI responses, and the NRC staffs safety evaluation.
Loss-of-Coolant Accident The NRC staffs audit in this area focused upon the following topics:
- 1. Calculation reports supporting results and conclusions presented in large- and small-break LOCA licensing reports (ANP-3639P and ANP-3640P), including loss-of-offsite power and offsite power available cases, reactor coolant pump (RCP) trip cases, break locations, other sensitivity studies, etc.
- 2. For realistic large-break LOCA (LBLOCA) analysis, view documentation of statistical fidelity during implementation, selection of number of cases, initial seed, etc.
- 3. Representative S-RELAP5 input decks for limiting large- and small-break LOCA cases.
- 4. Capability to plot S-RELAP5 parameters from limiting large- and small-break LOCA cases upon request during audit (may be done remotely if necessary).
- 5. Review of system parameters and initial conditions for large- and small-break LOCA analysis.
- 6. Comparison of sampled parameter ranges with actual plant limits and data (e.g., axial shape index sample range versus current COLR values, linear heat generation rate sampling range basis (CE only), safety injection tank liquid volume maximum range for sampling versus UFSAR Table 6.3.2-2, Revision 19, (ADAMS Accession No. ML17234A028), upper range of reactor coolant system flow.
- 7. Calculation results showing impact of predicted strain/rupture/relocation behavior.
LOCA-related analyses audited by the NRC staff are listed in the Section 4.3 of this document.
In addition to these topics from the audit plan, the NRC staff also discussed with the licensee a set of draft RAIs transmitted to the licensee before the site audit.
These RAIs covered topics including:
Impact of Framatome HTPTM fuel on post-LOCA debris accumulation in reactor vessel.
Outlying data in the prediction of certain LBLOCA figures of merit.
Reactor coolant pump trip sensitivity studies in small-break LOCA (SBLOCA) analysis.
Impact of TCD on larger SBLOCA scenarios.
Timing and impact of switchover to sump recirculation for SBLOCA analysis.
Treatment of mixed core configurations in LOCA analysis.
Reasonability of predicted output parameters in SBLOCA analysis.
Modifications to generically approved SBLOCA evaluation model.
Modeling of non-Framatome fuel in LBLOCA analysis.
Discussions concerning these topics during the audit provided the licensee a deeper understanding of the specific information being requested by the NRC staff; furthermore, the NRC staff gained insight into the licensees intended approach to address remaining concerns in each of these areas. Discussion of these topics in greater detail is anticipated in the NRC staffs RAIs, the licensees RAI responses, and the NRC staffs safety evaluation.
Technical Specifications The NRC staffs audit in this area focused upon a set of draft RAIs transmitted to the licensee prior to the site audit.
These RAIs covered topics including:
Adequacy of proposed wording for SL 2.1.1.2, regarding fuel centerline melt limits.
Addition of undefined term zirconium-alloy clad to TS 4.2.1.
Interpretation of proposed TS 4.2.1 with respect to lead test assemblies.
Applicability of certain methods defined in TS 5.6.5 (Core Operating Limits Report) only to specific fuel vendors; Applicability of certain methods defined in TS 5.6.5 to reactor cores incorporating batch-scale quantities of fresh fuel of two or more different designs (i.e., mixed fresh batches).
Applicability of radial fall off curve penalty imposed on Westinghouse Next Generation Fuel in Amendment No. 205 to future Westinghouse fuel designs to which the FATES3B code would be applied.
Discussions concerning these topics during the audit provided the licensee a deeper understanding of the specific information being requested by the NRC staff. In several cases, the licensee discussed potential revisions to the proposed TS changes included in its submittal as a means to address concerns discussed by the NRC staff during the audit. The NRC staff gained insight into the licensees intended approach to address remaining concerns in each of these areas. Discussion of these topics in greater detail is anticipated in the NRC staffs RAIs, the licensees RAI responses, and the NRC staffs safety evaluation.
Containment Analysis The NRC staffs audit in this area focused upon the following topics:
Discussion of the need for a quantitative versus qualitative evaluation to confirm any stated analyses of record remain bounding.
Expected differences in stored energy and decay heat between currently licensed fuel and requested amendment fuel.
Mass and Energy release for short- and long-term containment response.
Containment analysis related documents audited by the NRC staff are listed in Section 4.3 of this document.
Discussions of the above listed topics provided the licensee a deeper understanding of the information which the NRC staff is requesting to complete the safety evaluation. Additionally, the NRC staff was able to gain insight into the licensees intended approach to respond to
containment analysis related RAIs. Discussion of these topics in greater detail is anticipated in the NRC staffs RAIs, the licensees RAI responses, and the NRC staffs safety evaluation.
4.3 Supporting Information from the Licensee The licensee was requested to make the appropriate personnel or contractors, who are familiar with the proposed LAR, available for the audit (either in person or on the phone).
The NRC staff also requested the licensee to have the supporting documents related to the above topics available and be prepared to discuss them with the staff during the audit.
List of Documents Reviewed during the Audit
- 1. Document No. APS-CP-001, Implementation and Testing of the BHTP CHF Correlation into VIPRE-01 Thermal-Hydraulic Computer Code via DLL, Revision 1, March 28, 2018.
- 2. Engineering Evaluation, PV-E0897, V7, ENG WO #: 17-14728-009, January 2018.
- 3. Analysis No. RA-13-C00-2018-014, Framatome Fuel LAR; VIPRE-W to VIPRE-01 Benchmarking Comparison with BHTP CHF Correlation, PV-E1729P. Ver. 5a, October 26, 2018.
- 4. Letter from R. A. Clark (NRC) to A. E. Lundvall, Jr. (Baltimore Gas and Electric)
Forwards Safety Evaluation Accepting HERMITE/MacBeth Critical Heat Flux, dated July 15, 1983.
- 5. VDP Number A-15835, APS Log Number 13-N001-1301-01228-0, System 80 tm Inlet Flow Distribution Supplement 1-P to Enclosure 1-P to LD-82-054, February 1993.
- 6. VDP Number A12908, APS Log Number 13-M001-1301-1251-2, Statistical Combination of Uncertainties, October 26, 1995.
- 7. 80DP-0CC03, Revision 20, Non-Process Software Quality Assurance Program Implementation.
- 8. 80DP-0CC02, Revision 19, Non-Process Software Quality Assurance Program.
- 9. 32-2500462-000, Palo Verde VQP SBLOCA Base Deck Input Development, Revision 0.
- 10. ALION-REP-APS-9472-02, Framatome CE16 HTPTM Fuel Assessment Report, Revision 0.
- 11. 32-2500463-000, Palo Verde VQP SBLOCA Steady State Initialization and RODEX2-2a Input Development, Revision 0.
- 12. FS1-0043902, Palo Verde SBLOCA Switchover Inputs, Revision 1.
- 13. 32-2500465-000, Palo Verde VQP RLBLOCA Base Deck Input Development, Revision 0.
- 15. 32-2500466-001, Palo Verde VQP RLBLOCA Steady State Initialization, Revision 1.
- 16. 32-9259942-000, Palo Verde ECCS Flow Splits Analysis, Revision 0.
- 18. 32-9268125-000, Palo Verde VQP Realistic Large Break LOCA (RLBLOCA)
Uncertainty Analysis, Revision 0.
- 19. 32-9280432-001, Palo Verde VQP Inadvertent Opening of a Pressurizer Safety Relief Valve, Revision 1.
- 20. 51-2500534-000, Palo Verde Long-Term Cooling Evaluation for AREVA Fuel, Revision 0.
- 21. 51-9264109-002, Analytical Input Summary for Palo Verde VQP SBLOCA and RLBLOCA Analyses, Revision 2.
- 22. TA-13-C00-2019-002: CE16 HTPTM DNB ROPM Analysis.
- 23. TA-13-C00-2018-001: Framatome VQP: Summary of Non-LOCA Transients.
- 24. UFSAR Chapter15_events.
- 25. FS1-0036442, Revision 1: Palo Verde Main Steam Line Break (MSLB) Mass and Energy (M&E) Evaluation for Framatome Fuel.
- 26. FS1-0034571, Revision 1: Palo Verde Vendor Qualification Program (VQP) Information for Co-resident Fuel Assembly.
- 27. 162-15387-DWR/DAM: Transmittal of Information for Vendor Qualification Program (VQP).
- 28. TA-13-C00-2017-002, Revision 1: FRAMATOME VQP: Limiting Infrequent Event (UFSAR Appendix 15E), Appendix F - Fuel Failure Calculation Demonstration for a Mixed Core.
Documents Placed in Certrec CEA Ejection
- 1. 103-3785NP-000, Rod Ejection Accident (AREA) analysis for Palo Verde.
- 2. 103-3785P-000, Rod Ejection Accident (AREA) Analysis for Palo Verde.
- 3. FS1-0043389, Palo Verde Bounding AREA Analysis.
CETOP/BHTP Software QA Documents
- 5. Software QA Documents; (CETOP-D-DQAP, CETOP-D-SDD, CETOP-DSII, CETOP-D-SITR, CETOP-D SRS, CETOP-D-USMAN, CETOP-D-VVP, CETOP-D-VVTR).
COLSS-CPCS Setpoints - Mixed Core
- 6. COLSS-CPC Uncertainty Methods NRC 6-2019#2 - PowerPoint presentation at audit.
- 7. RA-13-C00-2019-016 Mixed Core COLSS-CPCS Setpoints.
Several Draft RAI Response
- 8. Most of the RAI responses.
Fuel failure - DNB Statistical Convolution
- 9. TA-13-C00-2017-002 - Anticipated Operational Occurrence (AOO) from Specified Acceptable Fuel Design Limit (SAFDL).
Mechanical Design Information
- 10. FS1-0023928 2.0 Palo Verde (PVE) Vendor Qualification Program (VQP) Mechanical Design Information for Safety and Neutronics Analyses.
- 11. FS1-0029226 1.0 Pale Verde (CE16x16) Fuel Design Mechanical Compatibility Evaluations (Upper Tie Plate).
- 12. FS1-0029792 1.0 Palo Verde (CE16x16) Fuel Design Mechanical Compatibility Evaluations (Lower Tie Plate).
- 13. FS1-0030336 1.0 Palo Verde (CE16x16) Fuel Design Mechanical Compatibility Evaluations (Guide Tube and Instrument Tube).
- 14. FS1-0031655 1.0 Palo Verde CE16x16 VQP Lift off Analysis.
- 15. FS1-0031752 1.0 PVE2-23 VQP Fuel Assembly Normal Operating Analysis.
- 16. FS1-0032029 1.0 PVE2-23 VQP CE16x16 HTPTM Fuel Assembly Shipping and Handling.
- 17. FS1-0032120 2.0 Palo Verde VQP Fuel Rod Cladding Stress and Buckling Analyses for Normal Operating Condition.
- 18. FS1-00322129 2.0 Calculation - Palo Verde VQPCOPPERNIC Based Creep Collapse Analysis.
LOCA Analyses
- 19. 32-2500462-000 Palo Verde VQP SBLOCA Base Deck Input Development.
- 20. 32-2500463-000 Palo Verde VQP SBLOCA Steady State Linearization and RODEX2 Input Development Plan.
- 21. 32-2500465-000 Palo Verde VQP RLBLOCA Base Deck Input Development.
- 22. 32-2500466-001 Palo Verde RLBLOCA Steady State Initialization.
- 23. 32-9259942-000 Palo Verde ECCS Flow Splits Analysis.
- 25. 32-9268125-000 Palo Verde VQP RLBLOCA Uncertainty Analysis.
- 26. 32-9280432-001 Palo Verde VQP Inadvertent Opening of A Pressurizer Safety Relief Valve.
- 27. 51-2500534-000 Palo Verde Long Term Cooling Evaluation for AREVA Fuel.
- 28. 51-9264100-002 Analytical Input Summary for Palo Verde VQP SBLOCA and RLBLOCA Analyses.
Containment Mass and Energy Releases
- 29. 162-15387-DWR.DAM.
- 30. FS1-0034571 1.0 Palo Verde VQP Information for Co-resident Fuel Assessment.
- 32. FS1-0037324 1.0 Palo Verde Units 1, 2 and 3 AREVA VQP LBLOCA M&E Analysis of Record (AOR) Evaluation (1).
- 34. LTR-CRA-17-149 Palo Verde MSLB M&E for AREVA Fuel (1).
Non-LOCA
- 35. UFSAR Chapter 15 events.
- 36. NA-13-C00-2017-010-R02-ILFA, Inadvertent Loading of a Fuel Assembly.
- 38. TA-13-C00-2018-207_r02-CEAE.
- 39. TA-13-C00-2019-002_r00-DNB ROPM.
- 40. TA-13-C00-2019-005 R00- Hot Zero Power (HZP) Control Element Assembly Withdrawal (CEAW).
- 41. TA-13-C00-2019-007_r00-CEA Drop.
Nuclear Analysis X-Sections and Depletions
- 42. NA-13-C00-2017-201r0 VQP N0 N1 N2 Depletion.
Thermal-Hydraulic Compatibility Analyses (Mixed Core Section 5.7 of Attachment 10
- 43. CN-APS-GEN-004 Palo Verde Mixed Core Thermal-Hydraulic Impact of the AREVA VQP Assemblies.
- 44. CN-NFPE-18-12 Assessment of Compatibility of AREVA and Westinghouse Fuel Assemblies in Palo Verde Cores - Selected Non-seismic LOCA Fuel and CEA Mechanical Aspects.
- 45. FS1-029626 1.0 Palo Verde VQP Thermal-Hydraulic Compatibility.
- 46. FS1-0030976 1.0 Palo Verde (APS) Guide Tube Heating Rate Calculations Using APOLLO-2A.
- 47. FS-0032658 2.0 Palo Verde VQP Flow-Induced Vibration (FIV) Analysis.
- 48. FS1-00034571 1.0 Palo Verde VQP Information for Co-resident Fuel Assessment.
- 49. FS1-0042640 Framatome Response to PVE Information Request 162-15666.
Thermal Hydraulics
- 50. APS-CP-001-R1-ZachryBHTP.
- 51. CHF-RAI-4 Response draft.
- 52. RA-13-C00-2017-201 Revision 0 Mixed Core IFD.
- 54. RA-13-C002018-013 Revision 0 Statistical Combination of Uncertainties (SCU)4FramatomeFuel.
- 55. RA-13-C00-2018-014 Revision 0 VIPRE-W to V1 Comparison.
- 56. RA-13-C00-2018-020-Revision 1-Baseline VIPRE Models.
- 57. RA-15-C00-2018-031-CETOP-Basedeck.
- 58. RA-13-C00-2019-005-R0-V1 SenAnalysis
- 59. RA-13-C00-2019-009 R0-CETOP Sensitivity.
- 60. RA-13-C00-2019-010-VIPRE-Mixed Core Demo.
- 61. RA-13-C00-2019-014 R0- V-W Sen Analysis.
- 62. RA-13-C00-2019-015-Boundary-Cond-Study.
- 63. VIPRE-1 APS_TwoPass-USMAN.
- 64. VIPRE-W_Users Manual for VIPRE-W External CHF Correlations.
Westinghouse Compatibility Evaluations
- 65. READ-ME-First (Descriptions of 15 Documents Listed below).
- 66. CN-APS-026 Revision 2 Fuel Mechanical Performance Analysis for NGF and Standard Fuel in Palo Verde Units 1-3 at Measurement Uncertainty Recapture (MUR) Conditions.
- 67. CN-APS-GEN-004 Palo Verde Mixed Core Thermal-Hydraulic Impact of the AREVA VQP Assemblies.
- 68. CN-NFPE-18-9 Fuel Assembly Seismic LOCA Analysis for the Implementation of AREVA Vendor Qualification Program (VQP) Assemblies in APS.
- 69. CN-NFPE-18-12 Assessment of Compatibility of AREVA and Westinghouse Fuel Assemblies in Palo Verde Cores - Selected Non-Seismic LOCA Fuel and CEA Mechanical Aspects.
- 70. CN-NFPE-18-24 Fuel Mechanical Design Assessment of Impact of Framatome VQP Fuel Assemblies on Westinghouse Fuel Assemblies for the Palo Verde Units.
- 71. CN-NFPE-18-25 Seismic Grid Strength Margins for Westinghouse Standard Fuel and CE16NGF' Assemblies in Palo Verde Cores with Eight Co-resident AREVA Fuel Assemblies (Beginning-of-Life (BOL) Fuel).
- 72. CN-SATH-17-019 Evaluation and Analysis for Palo Verde Units 1, 2 and 3 Loss of Coolant Analyses and Hydraulic Blowdown Loads for Framatome VQP Test Assemblies.
- 73. CN-SDA-17-8 Detailed Core Analysis for Palo Verde with 8 Framatome VQP Fuel Assemblies.
- 75. LTR-CRA-17-149 Palo Verde MSLB M&E Evaluation for AREVA Fuel.
- 76. LTR-SDA-18-002 Evaluation of the Effect of Eight VQP Fuel Assemblies on the Reactor Vessel Internals and the Reactor Coolant System at Palo Verde Nuclear Generating Station Uni.
- 77. LTR-SEE-15-109 Evaluation of Palo Verde Generating Station RSB 5-1 Cooldown (Shutdown Cooling System Entry to Cold Shutdown) with Fuel Transition to AREVA VQP.
- 78. LTR-SEE-17-245 RCS Hydraulic Changes to Palo Verde Generating Station Units 1, 2 and 3 Due to Insertion of AREVA VQP Assemblies and Transition to a Full Core of AREVA VQP Fuel.
- 79. LTR-TA-17-172 PVNGS 1, 2, and 3 System Performance (CENTS) Evaluation for the Vendor Qualification Program.
- 80. LTR-TA-18-3 PVNGS 1, 2 and 3 Operating and Design Transient and Control System Evaluations for the Vendor Qualification Program.
5.0 CONCLUSION
Through the audit, the NRC staff obtained an enhanced understanding of the licensees submittals and the details of the included safety analyses and their results. There was open communication throughout the audit, and this helped the NRC staff to communicate concerns about the submittals and have them answered by APS and Framatome. The NRC staff provided draft RAIs much before the site audit took place so that the licensee can provide the draft responses for the NRC review before the site audit. Because of the discussions that were conducted at this audit, the number of potential draft RAIs has been reduced and the scope of the remaining questions has been focused directly on the topic of concern.
ML19235A256 *Audit Summary by memo OFFICE NRR/DORL/LPL4/PM NRR/DORL/LPL4/LA NRR/DSS/SNPB/BC*
NAME SLingam PBlechman RLukes DATE 9/4/19 9/4/19 8/12/19 OFFICE NRR/DSS/SNPB/BC (A)* NRR/DORL/LPL4/BC NRR/DORL/LPL4/PM NAME JBorromeo RPascarelli SLingam DATE 8/12/19 9/6/19 9/9/19