ML25090A018
| ML25090A018 | |
| Person / Time | |
|---|---|
| Site: | Palo Verde  |
| Issue date: | 03/28/2025 |
| From: | Spina J Arizona Public Service Co |
| To: | Office of Nuclear Reactor Regulation, Document Control Desk |
| References | |
| 102-08927-JLS/KJG | |
| Download: ML25090A018 (1) | |
Text
10 CFR 50.90 A member of the STARS Alliance, LLC Callaway
- Diablo Canyon
- Palo Verde
- Wolf Creek JENNIFER L. SPINA Vice President Nuclear Regulatory and Oversight Palo Verde Nuclear Generating Station P.O. Box 52034 Phoenix, AZ 85072 Mail Station 7605 Tel 623.393.4621 102-08927-JLS/KJG March 28, 2025 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001
Subject:
Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos. STN 50-528,59-529, and 50-530 Renewed Operating License Number NPF-41, NPF-51, and NPF-74 Supplement to License Amendment Request to Revise the Technical Specifications 3.5.1 and 3.5.2 Safety Injection Tank Pressure Bands, and to Use GOTHIC Code Pursuant to 10 CFR 50.90, Arizona Public Service Company (APS) is submitting a supplement to our request for an amendment to the Technical Specifications (TS) for Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3. APS originally submitted an amendment in a letter dated August 28, 2024 [Agencywide Documents Access and Management System (ADAMS) Accession No. ML24241A278] requesting the approval of revising TS Section 3.5.1, Safety Injection Tanks (SITs) - Operating and TS Section 3.5.2, Safety Injection Tanks (SITs) - Shutdown and their bases.
Specifically, the proposed TS changes revise Surveillance Requirement (SR) 3.5.1.3 and SR 3.5.2.3 to increase the upper limit of their SIT pressure bands, and to list their pressure requirements in units of pounds per square inch absolute (psia), as reflected in the PVNGS safety analyses, with no instrument uncertainties included, instead of the SIT instrument units of pounds per square inch gauge (psig) with instrument uncertainties included. The proposed changes also include use of the Generation of Thermal Hydraulic Information for Containments (GOTHIC) code as part of the methodology to perform calculations of the containment pressure and temperature response to various postulated pipe breaks.
As a result of an ongoing audit, in which APS provided information to the NRC staff reviewers electronically, the NRC staff identified six (6) items that require more information to support their review. The enclosure to this letter provides the additional information.
No new commitments are being made to the NRC by this letter.
Should you need further information regarding this letter, please contact Michael D.
Dilorenzo, Licensing Department Leader, at (623) 393-3495.
102-08927-JLS/KJG ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Supplement to License Amendment Request to Revise the Technical Specifications 3.5.1 and 3.5.2 Safety Injection Tank Pressure Bands, and to Use GOTHIC Code Page 2 I declare under penalty of perjury that the foregoing is true and correct to the best of my knowledge.
Executed on __03/28/2025__________
(Date)
Sincerely, JLS/MDD/cr
Enclosure:
Supplemental Information cc:
J. D. Monninger NRC Region IV Regional Administrator W. T. Orders NRC NRR Project Manager for PVNGS N. Cuevas Resident Inspector for PVNGS Spina, Jennifer (Z08962)
Digitally signed by Spina, Jennifer (Z08962)
Date: 2025.03.28 17:02:41 -07'00'
1 ENCLOSURE Supplemental Information
ENCLOSURE Response to NRC Audit Questions 1
=
Background===
By letter dated August 28, 2024, [Reference 1] Arizona Public Service Company (APS) submitted a license amendment request (LAR) for Palo Verde Nuclear Generating Station (PVNGS), Units 1, 2, and 3. The amendments would modify PVNGS Technical Specification (TS) section 3.5.1, Safety Injection Tanks (SITs) - Operating and TS section 3.5.2, Safety Injection Tanks (SITs) - Shutdown, and their associated bases.
Specifically, the proposed TS changes revise PVNGS Surveillance Requirements 3.5.1.3 and 3.5.2.3 to increase the upper limit of their SIT pressure bands and to list their pressure requirements in units of pounds per square inch absolute as reflected in the PVNGS safety analyses, with no instrument uncertainties included, instead of the SIT instrument units of pounds per square inch gauge with instrument uncertainties included. Using the safety analysis limit is consistent with how APS has treated the Safety Injection Tank volume limits (SR 3.5.1.2 and LCO 3.5.2) and the LCO 3.6.5 Containment Temperature limit. Conversion between the safety analysis limits and the operating procedures is done per the PVNGS 10CFR50 Appendix B design control program.
In addition, the proposed changes would include use of the Generation of Thermal Hydraulic Information for Containments (GOTHIC) code as part of the methodology to perform calculations of the containment pressure and temperature response to various postulated pipe breaks.
The NRC staff requested a regulatory audit to support their review of the request [Reference 2]. As part of the audit plan, the NRC staff identified six (6) items that required more information to support their review. APS is documenting the NRC requests, followed by the APS response in this Enclosure.
Audit Item 1 Document Request:
, Benchmark Evaluation for Use of the GOTHIC Code, of the LAR enclosure presents benchmarking of the proposed GOTHIC, Version 8.4 with the currently used COPATTA code. Since GOTHIC is not an NRC-approved code but accepted for containment response analysis, and the future possible changes in GOTHIC are not controlled by APS, provide documentation describing the method that will be used to justify the use of future GOTHIC versions that will not result in a numerical departure from the proposed containment response analysis.
APS Response to Audit Item 1:
Procedure 80DP-0CC02, Control of Non-Process Software and Data [Reference 3],
establishes a process with controls to ensure installation, design, testing, modification, and use of software and computer databases are commensurate with the importance to safety at PVNGS. In accordance with 80DP-0CC02, Sections 4.3.6H and 4.3.6J, new or revised software is required to go through a verification and validation process such as regression, functional, and user acceptance testing prior to being approved. Performance testing includes ensuring the software adequately and correctly performs all intended functions, properly handles abnormal conditions and events, as well as credible failures, does not perform adverse unintended functions, and does not degrade the system either by the computer program, or in combination with other functions or configuration items. In accordance with 80DP-0CC02, Section 4.3.6J, it is ensured that identification of affected software baselines is documented and further, that the verification and validation test plans
ENCLOSURE Response to NRC Audit Questions 2
document traceability, regression testing as applicable, testing of standards if applied to the software, required tests and test sequence, and overall criteria for acceptance. Procedure 80DP-0CC02 Section 4.1.5 also states that adding new software or revising software (using a revised version of software) is subject to 10 CFR 50.59 and/or 10 CFR 72.48 review per 93DP-0LC07 [Reference 4].
Procedure 93DP-0LC07 and associated Administrative Guideline (AG) 93DP-0LC07-01
[Reference 5] contain detailed requirements and guidance regarding determining if the proposed activity (software revision change) replace or adversely revise an evaluation of method of evaluation described in the Updated Final Safety Analysis Report (UFSAR). AG 93DP-0LC07-01, Section 5.2.5, states that, changes to software methodology or applications used to perform evaluations or that are a method of evaluation, shall be assessed by an individual that has knowledge and expertise in the application of the method or software application being reviewed. COPATTA, for the calculation of containment temperature and pressure, is listed in 93DP-0LC07-01 as being a method of evaluation as defined in Section 3.10 of NEI 96-07 [Reference 6], which drove APS to seek prior NRC approval before switching to GOTHIC as using GOTHIC instead of COPATTA evaluated in during the 10 CFR 50.59 process.
Future revisions to GOTHIC would enter the same 10 CFR 50.59 process. Section 6.2.3.4 of AG 93DP-0LC07-01 provides guidance on how to address software revisions associated with a method of evaluation (MOE). It states that, Use of a later version or release of a software program that is utilized as part of an MOE described in the UFSAR would typically be evaluated as a change to an element of an MOE. Per 93DP-0LC07-01 Section 4.3.6.4 elements of an MOE are part of the methods of evaluation and are subject to the use of the method. Section 6.2.3.6c of AG 93DP-0LC07-01 states, Changes to methods of evaluation included in the UFSAR are considered adverse and require evaluation under 10 CFR 50.59 or 72.48 if the changes are outside the constraints and limitations associated with use of the methodIf the changes are within the constraints and limitations associated with the use of the method, then the change is not considered adverse and may be screened out.
Methodology departures from a method of evaluation described in the UFSAR used in establishing the design bases or in the safety analyses would require NRC approval prior to implementation [Reference 4, Section 4.3.16 and Reference 5, Section 7.2.8].
Therefore, PVNGS procedural controls and the implementation of 10 CFR 50.59 ensure that future GOTHIC versions will not result in a numerical departure from the proposed containment response analysis. Further, analysis models and methods described in the benchmark evaluation are not restricted to the GOTHIC version 8.4 as PVNGS procedural controls and the implementation of the 10 CFR 50.59 process ensures methodology departures from GOTHIC version 8.4 would require NRC approval prior to implementation.
Audit Item 2 Document Request:
In section 3.0, Benchmarking of GOTHIC Results, of attachment 5 to the LAR enclosure, the licensee states, in part:
When the results between GOTHIC and COPATTA differed, GOTHIC parameters were subsequently adjusted and additional runs were performed, to ensure the reason(s) for the differences were understood.
ENCLOSURE Response to NRC Audit Questions 3
(a) Provide documentation describing which parameters in GOTHIC model were selected and adjusted in the benchmark process of GOTHIC with COPATTA code. Provide reasons and justification for selecting these parameters.
(b) For each parameter adjusted, provide documentation describing what reasons were understood from the differences in the benchmarking results.
APS Response to Audit Item 2:
(a) Parameters selected and adjusted are summarized from 13-NC-ZC-0232 Appendix F
[Reference 7] in the table below. The table also includes the results from the sensitivity runs for each parameter adjusted as summarized from 13-NC-ZC-0232 Appendix F. Parameters were selected based on software knowledge of what could be causing differences in results. Note that the sensitivity runs are only to understand the deltas and are not used to imply that the official GOTHIC run used for the analysis should make these modeling changes or additions.
Observed Differences between COPATTA and GOTHIC using identical inputs GOTHIC Parameters adjusted in a sensitivity run Results from the sensitivity run and understood reasons GOTHIC showed a slightly lower containment pressure profile and different shape over time (the GOTHIC pressure profile was absent of a momentary sharp reduction in pressure before reaching the peak pressure as shown in the COPATTA pressure profile).
A recirculation fan model is added to GOTHIC (with a new flow path configuration removing all liquid droplets and directing them to the liquid phase or sump). This is intended to replicate the COPATTA results by quickly removing suspended droplets using the high fan flow rate through the flow path and droplet removal capability of the flow path.
The resulting GOTHIC sensitivity run profiles (containment pressure and vapor temperature) nearly exactly match the COPATTA profiles, while the sump temperature profile showing much closer agreement, demonstrating that the initial discrepancy is due to the way COPATTA removes the suspended liquid from the containment atmosphere. At the end of each time step, COPATTA removes the liquid phase from the containment atmosphere and adds it to the sump. In GOTHIC, the liquid enters as droplets that remain suspended in the atmosphere and maintain the containment atmosphere at saturation, with no superheat. The GOTHIC modeling is more realistic.
GOTHIC showed a containment vapor temperature which is maintained at saturation compared to the COPATTA results which show a spike in temperature above the saturation temperature at the steam pressure.
GOTHIC showed a containment sump temperature profile which was noticeably lower than the COPATTA sump profile for approximately the first 500 seconds of the run.
(b) See response for (a) above.
ENCLOSURE Response to NRC Audit Questions 4
Audit Item 3 Document Request:
In section 3.0, Benchmarking of GOTHIC Results, of attachment 5 to the LAR enclosure, the licensee states, in part:
The conversion of the COPATTA model of the DEDLS [double-ended discharge leg slot]
break LOCA [loss-of-coolant accident] with maximum SI [safety injection] flow to a corresponding GOTHIC model is performed using GOTHIC Version 8.4 [GOTHIC, Generation of Thermal Hydraulic Information for Containments, Version 8.4(QA),
Electric Power Research Institute, Palo Alto, CA, 2022]. However, the analysis models and methods described in this benchmark evaluation are not intended to be restricted to a specific GOTHIC code version.
Provide documentation describing your justification that the analysis models and methods described in the benchmark evaluation are not restricted to the GOTHIC 8.4 version.
APS Response to Audit Item 3:
See Audit Item 1 Response.
Audit Item 4 Document Request:
In LAR enclosure, attachment 6, Technical Analysis of Changes to the SIT Pressure Bands, section 5.0, Post-LOCA Containment Pressure and Temperature Response - Impact of Changes to The SR 3.5.1.3 SIT Pressure Band, the licensee does not provide the impact of the changes on the containment sump temperature response and on the net positive suction head (NPSH) of the pumps that draw water from the sump during the post-LOCA recirculation phase.
The licensee is requested to provide documentation of the following information:
(a) Post-LOCA sump temperature response based on the Westinghouse and Framatome fuels.
(b) Transient available NPSH, required NPSH, and minimum NPSH margin for the pumps that draw water from the sump during post-LOCA recirculation phase.
(c) Any containment accident pressure (CAP) above the vapor pressure at the transient sump temperature used in the calculation of transient available NPSH in item (b).
APS Response to Audit Item 4:
(a) Reported post-LOCA sump maximum temperature for the bounding case that bound all licensed fuel types has not changed as a result of the LAR. Using the GOTHIC code (with the increased pressure band) results in lower post-LOCA sump temperature in the bounding case and thus, APS has conservatively selected to report out the previously calculated COPATTA maximum temperature to downstream analyses for the bounding LOCA case. This does not imply that COPATTA produces the output of the calculation; instead, the raw GOTHIC results have discretionary margin added to them that results in the reported output of the bounding calculation being the same as that previously reported. Graphs (including numerical call outs) showing the previous COPATTA results and the raw GOTHIC data, plus the bounding profile (equal to the maximums from the previous COPATTA results) are shown below from Calculation 13-NC-ZC-0232
[Reference 8, Figures 7.2 and 7.5].
ENCLOSURE Response to NRC Audit Questions 5
ENCLOSURE Response to NRC Audit Questions 6
ENCLOSURE Response to NRC Audit Questions 7
(b) Per (a) the reported post-LOCA sump temperatures used in downstream NPSH analyses have not changed as a result of the LAR and as such, there is no change to NPSH. What is currently stated in the PVNGS UFSAR [Reference 9] regarding NPSH remains accurate.
UFSAR Section 6.2.2 details the NPSH methodology and minimum NPSH margins for all pumps taking suction from the sump during recirculation.
Per UFSAR Section 6.2.2.2.1.G Item Pump NPSH:
The NPSH calculation was performed from sump temperatures ranging from 77°F to 252°F and is based on the following:
- The minimum containment flood level of El. 84-6 is used when determining available NPSH.
- Credit is not taken for containment accident pressure above saturation in determining available NPSH; however, minimum partial pressure of air in containment prior to the accident is considered. The minimum initial partial pressure of air is only credited for sump water temperatures below the temperature at which the vapor pressure is equal to the initial partial pressure of air (193.8°F).
For sump water temperatures greater than 193.8°F, the saturated sump model is used Minimum NPSH margins are as follows (copied from UFSAR Section 6.2.2.2.1.G Item Pump NPSH):
Further details regarding the minimum containment water level are provided in UFSAR Section 6.2.1.1.2.4. UFSAR Table 6.3.2-1, Section 6.3.1.4.M.3, Section 6.3.1.4.M.4, and Section 6.2.2.2 provide numerical values and details regarding HPSI pump NPSH, LPSI pump NPSH, and CS pump NPSH; required NPSH is copied from the UFSAR and included in the table below. Note that a transient NPSH is not part of the PVNGS design and licensing basis.
Pump NPSH Required (copied from UFSAR Section 6.2.2.2 and Table 6.3.2-1)
Pump NPSHR [ft]
HPSI 25 (recirc, runout, and injection)
LPSI 20 (recirc); 26 (runout and injection)
CS 20 (recirc, runout, and injection)
(c) Please refer to (b) above.
ENCLOSURE Response to NRC Audit Questions 8
Audit Item 5 Document Request:
In LAR enclosure, attachment 6, section 3.1.1, CE16STD Large Break LOCA, the licensee states, in part:
The core wide cladding oxidation (CWO) cases are not limiting with respect to maximum SIT pressure.
Provide documentation explaining why the CWO cases are not limiting with respect to the maximum SIT pressure.
APS Response to Audit Item 5:
A safety injection tank (SIT) study was performed by the vendor (Westinghouse Electric Company (WEC)) in accordance with the appropriate requirements of the 1999 Evaluation Method (EM) for Large Break Loss of Coolant Accident (LBLOCA) Emergency Core Cooling System (ECCS) Performance analysis of Combustion Engineering (CE) designed plants.
The 1999 EM, CENPD-132 - Calculative Methods for the CE Large Break LOCA Evaluation Model, is approved by the Nuclear Regulatory Commission (NRC) and complies with the requirements of Appendix K of 10 CFR 50 and is the PVNGS licensing basis as listed in Technical Specification 5.6.5.b.5 and UFSAR section 6.3.6 [Reference 10].
As part of the LBLOCA analysis of record for CE16STD fuel, sensitivity studies were performed on Time-in-Life, Safety Injection Tank (SIT), Break Spectrum and Failure Modes to show bounding results for Peak Cladding Temperature (PCT), Peak Local Oxidation (PLO) and Core Wide Cladding Oxidation (CWO). The results of the study determined that the SIT conditions of minimum pressure, maximum liquid volume, minimum temperature, and maximum discharge friction factor were the limiting configuration for CWO. Thus, based on the results of the sensitivity study in the analysis of record, the sentence in Section 3.1.1 of and Section 4.3 of Reference 11 that states the CWO cases are not limiting with respect to maximum SIT Pressure was made.
Audit Item 6 Document Request:
In LAR enclosure, attachment 6, sections 3.1.1, CE16STD Large Break LOCA, and 3.1.2, CE16NGF Large Break LOCA, the licensee states CE16STD and CE16NGF large break LOCA analysis of record (AOR) limiting peak cladding temperature (PCT) cases has been reevaluated for an increase in its modeled maximum SIT pressure to 675 psia. As stated in these sections, the impact on the PCT for both cases is 2°F resulting in a PCT of 2108°F for CE16STD and PCT of 2132°F for CE16NGF. The licensee is requested to provide documentation of the following:
(a) The large break LOCA re-analysis methodology used to re-evaluate the impact on PCT, peak local oxidation (PLO), and CWO.
(b) Confirm the methodology in response to (a) is the same as in the AOR. Provide justification if a different methodology is used.
(c) Confirm the inputs and assumptions in the re-analysis using the methodology in response to (a) were kept the same as in the AOR with the exception of the change in the SIT pressure. In case the conservatism in any input or assumption was reduced, provide justification.
ENCLOSURE Response to NRC Audit Questions 9
APS Response to Audit Item 6:
The methodology used in the evaluation to determine the change in PCT is as follows:
- 1. The limiting PCT and PLO case as applicable is identified in the respective analysis of record (AOR), CE16STD or CE16NGF.
- a. All the input files for each of the identified cases are retrieved from data storage.
- b. The AOR case(s) are rerun on the current system platform
- 2. Each unique case file for the sensitivity study is created from the AOR files with only the maximum SIT pressure being changed.
- 3. The large break LOCA sensitivity cases are run using the AOR input files and the unique case file via NRC approved large break LOCA 1999 Evaluation Model (EM),
CENPD-132 - Calculative Methods for the CE Large Break LOCA Evaluation Model, which is the PVNGS licensing basis as listed in Technical Specification 5.6.5.b.5 and UFSAR section 6.3.6 [Reference 10].
- 4. The results of the case(s) are extracted from the case output and collated determine the final result.
There is no change in the methodology between the previous AORs and the new sensitivity analysis, the same NRC approved large break LOCA 1999 EM is used. With the exception of the SIT pressure, all of the inputs and assumptions used in the sensitivity analysis are the same as in the previous AORs.
Additionally, in support of this response, the vendor (Westinghouse Electric Company) proprietary LOCA evaluation [Reference 11] was made available for the NRC staff member(s) to review as part of the ongoing audit.
==
Conclusions:==
The APS responses answer the document requests posed by the NRC, in response to the NRC review of APS license amendment request, to revise PVNGS Technical Specifications (TS) 3.5.1 and TS 3.5.2 and to use GOTHIC code.
References (Background):
- 1. ML24241A278, Palo Verde Nuclear Generating Station Units 1, 2, and 3 Docket Nos.
STN 50-528,59-529, and 50-530 Renewed Operating License Number NPF-41, NPF-51, and NPF-74 License Amendment Request to Revise the Technical Specifications 3.5.1 and 3.5.2 Safety Injection Tank Pressure Bands, and to Use GOTHIC Code, From Todd Horton (APS), dated August 28, 2024, (SWMS ID 102-08841).
- 2. ML25016A345, Regulatory Audit in Support of License Amendment Request to Revise Technical Specifications 3.5.1 and 3.5.2 and to use GOTHIC Code for Palo Verde Nuclear Generating Station, Units 1, 2, AND 3 Docket NOS: 50-528, 50-529, and 50-530 EPID: L-2024-LLA-0116, from William Orders (NRC) to Adam Heflin (APS), Dated January 21, 2025.
References (Audit Item 1):
- 3. PVNGS PROC 80DP-0CC02, Revision 23, Control of Non-Process Software and Data
- 4. PVNGS PROC 93DP-0LC07, Revision 36, 10 CFR 50.59 and 72.48 Screenings and Evaluations
ENCLOSURE Response to NRC Audit Questions 10
- 5. PVNGS PROC 93DP-0LC07-01, Revision 12, 10 CFR 50.59 and 72.48 Administrative Guideline
- 6. NEI 96-07, Revision 1, Guidelines for 10 CFR 50.59 Implementation References (Audit Item 2):
- 7. PVNGS CALC 13-NC-ZC-0232, Revision 13, Loss of Coolant Accident Pressure and Temperature Containment Analysis for Limiting Case References (Audit Item 3):
(Same as Audit Item 1 References)
References (Audit Item 4):
- 8. PVNGS CALC 13-NC-ZC-0232, Revision 13, Loss of Coolant Accident Pressure and Temperature Containment Analysis for Limiting Case
- 9. PVNGS Updated Final Safety Analysis Report, Revision 22A, dated February 2024 References (Audit Item 5):
- 10. PVNGS Updated Final Safety Analysis Report, Revision 22A, dated February 2024
- 11. N001-0205-00317, Rev 00, Evaluation of a SIT Maximum Pressure Increase on the Palo Verde Nuclear Generating Station Units 1, 2, and 3 LBLOCA Analysis of Record,
[SDOC CVERM-LOCA-CN-AA-000001, Rev 0].
References (Audit Item 6):
(Same as Audit Item 5 References)