Latest revision as of 16:34, 4 February 2020

Text

Response to NRC Staff Request for Additional Information Regarding License Amendment and Exemption Requests Related to the Implementation of Next Generation Fuel
	ML17153A373
Person / Time
Site:	Palo Verde
Issue date:	06/02/2017
From:	Lacal M; Arizona Public Service Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
Shared Package
ML17153A372	List: ML17153A373;
References
	102-07509-MLL/TNW
	Download: ML17153A373 (88)
	v • d • e

Attachments 1, 3, and 4 of the Enclosure contain PROPRIETARY information to be withheid under 10 CFR 2.390 Qaps MARIA L. LACAL Senior Vice President, Nuclear Regulatory & Oversight Palo Verde 102-07509-MLL/TNW Nuclear Generating Station June 2, 2017 P.O. Box 52034 Phoenix, AZ 85072 Mail Station 7605 Tel 623.393.6491 U. S. Nuclear Regulatory Commission ATTN: Document Controi Desk Washington, DC 20555-0001 References; 1. Arizona Pubiic Service Company (APS) letter number 102-07277, License Amendment Request and Exemption Request to Support the Impiementation of Next Generation Fuei, dated July 1, 2016, [Agencywide Documents Access and Management System (ADAMS) Accession Number ML16188A332]

2. NRC correspondence to APS, Paio Verde 1, 2, and 3 - NGF LAR and Exemption RAIs (CAC Nos. MF8076 to MF8081), dated April 14, 2017,

[ADAMS Accession Number ML17107A005]

Dear Sirs:

Subject; Palo Verde Nuclear Generating Station (PVNGS)

Units 1, 2, and 3 Docket Nos. STN 50-528/529/530 Renewed Operating License Nos. NPF-41, NPF-51, NPF-74 Response to NRC Staff Request for Additional Information Regarding License Amendment and Exemption Requests Related to the Implementation of Next Generation Fuei (NGF)

By letter dated July 1, 2016 (Reference 1), Arizona Public Service (APS) submitted a license amendment request (LAR) pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR) and an exemption request pursuant to the provisions of 10 CFR 50.12 for PVNGS Units 1, 2, and 3 requesting approval of changes to the PVNGS Technical Specifications (TS) related to the implementation of NGF.

The proposed LAR would revise the TS requirements related to using Optimized ZIRLO as an approved fuel clad material and revise the TS referenced topical reports in the Core Operating Limits Report (COLR). These references include analytical methods that will be used to determine core operating limits following NGF, VIPRE-W Code, Critical Heat Flux (CHF) correlation, and zirconium diboride burnable absorber methodology implementation.

The U.S. Nuclear Regulatory Commission (NRC) staff provided requests for additional information (RAIs) by NRC correspondence, dated April 14, 2017 (Reference 2). The Enclosure to this letter provides the APS response to the RAIs. The RAIs do not affect the conclusions of the no significant hazards consideration determination [10 CFR 50.91(a)]

provided in the LAR.

A member of the STARS Alliance LLC Callaway

Diablo Canyon

Palo Verde

Wolf Creek Attachments 1, 3, and 4 transmitted herewith contain PROPRIETARY information.

When separated from Attachments 1, 3, and 4, this transmittai is decontrolled.

102-07509-MLI7TNW ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Response to NRC RAIs Regarding LAR to Impiement NGF Page 2 In the LAR, APS proposed a commitment to address thermal conductivity degradation (TCD) considerations of NGF. In accordance with the RAI 6, Subpart b response provided in the enclosure, APS proposes to replace the commitment with the foilowing iicense condition to account for TCD effects:

A fuel centerline temperature allowance at high burnup, as specified in Attachment 8 of the enclosure to APS letter 102-07277, dated July 1, 2016, will be set aside to account for the burnup dependent effects of Thermal Conductivity Degradation (TCD) when using the FATES3B code to determine input for non-LOCA and LOCA safety analyses.

In addition, the foilowing new commitment is being made in this submittai to support a transition to a new iong-term fuel evaluation model and associated methods upon approval by the NRC:

Upon NRC approval of a new long-term fuel evaluation model and associated methods that explicitly account for thermal conductivity degradation (TCD) that are applicable to PVNGS, APS will, within 6 months, provide and implement a schedule for reanalysis using the NRC approved new long term fuel evaluation model that is applicable to the PVNGS NGF design for affected licensing basis analysis.

A portion of the RAI responses involve proprietary information from Westinghouse Electric Company LLC. Attachment 6 of the enclosure is the Westinghouse affidavit signed by Westinghouse Electric Company LLC that sets forth the basis on which the proprietary information in Attachments 1, 3, and 4 of the enclosure 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 Attachments 1, 3, and 4 of the enclosure or the supporting Westinghouse affidavit should reference Westinghouse letter number CAW-17-4595 and be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066. Attachments 1, 3, and 4 of the enclosure contain proprietary information.

Following approval of the LAR in Reference 1, APS requests an implementation date 90 days from the date of NRC approval of the license amendment.

If you have any questions about this request, please contact Michael D. DiLorenzo, Licensing Section Leader, at (623) 393-3495.

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

Executed on: June 2. 2017 Sincerely, Lacal, Maria Digitally signed by Lacal, Maria L(Z06t49)

DN: cn=Lacat, Maria L(Z06149)

L(Z06149) Date: 2017.06.02 13:16:37 -OTOO*

MLL/TNW/MSC/sma

102-07509-MLLyTNW ATTN: Document Control Desk U. S. Nuclear Regulatory Commission Response to NRC RAIs Regarding LAR to Implement NGF Page 3

Enclosure:

Response to NRC Staff Requests for Additional Information Regarding License Amendment and Exemption Request to Implement Next Generation Fuel cc: K. M. Kennedy NRC Region IV Regional Administrator S. P. Lingam NRC NRR Project Manager for PVNGS M. M. Watford NRC NRR Project Manager C. A. Peabody NRC Senior Resident Inspector for PVNGS

Enclosure Response to NRC Staff Requests for Additional Information Regarding License Amendment and Exemption Request to Implement Next Generation Fuel

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR TABLE OF CONTENTS Enclosure - Response to NRC Staff Requests for Additional Information Regarding License Amendment and Exemption Request to Implement Next Generation Fuel - Westinghouse Responses to the NRC RAIs on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (Proprietary) - Westinghouse Responses to the NRC RAIs on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (Non-Proprietary) - Confirmation of FATES3B Thermal Conductivity Degradation Allowance for the Arizona Public Service Next Generation Fuel License Amendment Request (Proprietary) - SKBOR: A Computer Code for Calculating the Accumulation of Boric Acid in the Reactor Vessel (Proprietary) - SKBOR: A Computer Code for Calculating the Accumulation of Boric Acid In the Reactor Vessel (Non-Proprietary) - Affidavit from the Westinghouse Electric Company Submitted in Accordance with 10 CFR 2.390 to Consider Attachments 1, 3, and 4 as Proprietary

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Introduction By letter dated July 1, 2016 [Agencywide Documents Access and Management System (ADAMS) Accession No. ML16188A332] (Reference 1), Arizona Public Service (APS) submitted a license amendment request (LAR) pursuant to the provisions of Section 50.90 of Title 10 of the Code of Federal Regulations (10 CFR) and an exemption request pursuant to the provisions of 10 CFR 50.12 for PVNGS Units 1, 2, and 3 requesting approval of changes to the PVNGS Technical Specifications (TS) related to the implementation of NGF.

The proposed LAR would revise the TS requirements related to using Optimized ZIRLO as an approved fuel clad material and revise the TS referenced topical reports in the Core Operating Limits Report (COLR). These references include analytical methods that will be used to determine core operating limits following NGF, VIPRE-W Code, Critical Heat Flux (CHF) correlation, and zirconium diboride burnable absorber methodology implementation.

The U.S. Nuclear Regulatory Commission (NRC) staff provided requests for additional information (RAIs) by NRC document, dated April 14, 2017 (ADAMS Accession No. ML17107A005) (Reference 2). This enclosure provides the APS response to the RAIs. The NRC staff requests for information are provided first, followed by the APS response. The enclosure also includes RAI responses from Westinghouse (Reference 4), as indicated in the appropriate RAI response.

NRC Staff Request 1:

CENPD-178-P-A Methodology The CENPD-178-P-A methodology referenced in the submittal specifies that a hydraulic shaker attaches to the bottom of the assembly during the forced vibration fuel assembly testing. The testing facility contains an apparatus that uses an electro-mechanical shaker that attaches to the center of the assembly. Please clarify which shaker is used and justify its use as part of the methodology.

APS Response 1:

The non*proprietary version of the response to RAI 1 is provided in Attachment 2 of this enclosure and the proprietary version is provided in Attachment 1 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

NRC Staff Request 2:

Transition Core Control Element Assembly Drop Times Please provide Justification that the full core Next Generation Fuel (NGF) and full core Value Added Fuel (referred to as STD) control element assembly (CEA) drop time analyses bound the CEA drop times expected for transition cores containing both NGF and STD fuel assemblies.

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR APS Response 2:

The non-proprietary version of the response to RAI 2 is provided in Attachment 2 of this enclosure and the proprietary version is provided in Attachment 1 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

NRC Staff Request 3:

Bounded and Non-Imoacted Chapter 15 Events In Section 7 of Attachment 8, Non-Loss-of-Coolant-Accident (LOCA) Safety Analysis, Table 7-1 lists the impact of the use of NGF on Chapter 15 Non-LOCA events.

a. Transition to NGF fuel is determined to have "no impact" on a number of Chapter 15 events, listed below. For these events, please explain the process for determining that the inputs are unchanged and justify why they are unchanged. If any of the input has changed, please Justify that the event is not impacted.

Startup of an Inactive Reactor Coolant Pump

Inadvertent Deboration

Inadvertent Operation of the Emergency Core Cooling System (ECCS)

Steam Generator Tube Rupture

b. A number of Chapter 15 events, listed below, are determined to be "bounded" without specific Justification. Please Justify that these events are bounded. Please identify the bounding assumptions and Justify that they are appropriate. If the bounded event has been quantitatively analyzed, please provide the margin between the new NGF analysis and the bounding analysis. For comparison, please also provide the analogous margin associated with the current "bounded" STD analysis of record (AOR) for these events.

Decrease in Feedwater (FW) Temperature

Increase in Main FW Flow

Loss of External Load

Turbine Trip

Main Steam Isolation Valve Closure

Loss of Non-emergency AC Power to the Station Auxiliaries

Loss of Normal FW Flow

Chemical and Volume Control System Malfunction - Pressurizer Level Control System Malfunction with Loss of AC Power (LOP)

APS Response 3:

The non-proprietary version of the response to RAI 3 is provided in Attachment 2 of this enclosure and the proprietary version is provided in Attachment 1 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR NRC Staff Request 4:

Evaluated Chapter 15 Events

a. Please justify that the CENTS evaluations, completed as part of the Chapter 15 AORs, remain applicable for the transient system response, and do not require re analysis to support the transition to NGF.

b. Attachment 7/8, Section 7.1.3: Please demonstrate that the "system model changes" due to NGF are bounded by the AOR for the Increase in Main Steam Flow (IMSF),

Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (lOSGADV),

and lOSGADV with LOP events. Please demonstrate that the AOR flow coastdown remains conservative. Please identify what system model changes are necessary because of NGF, and justify why they are small. Please demonstrate that these small changes result in an insignificant impact to the overall transient system.

c. Attachment 7/8, Section 7.1.3: Please demonstrate that the current analysis for calculated fuel failure (i.e. AOR Departure from Nucleate Boiling Ratio (DNBR) vs.

NGF DNBR) bounds the IMSF+LOP and lOSGADV+LOP events.

d. Attachment 7/8, Section 7.3.4: Please identify the fuel specific failure analysis methodology and describe the fuel failure analysis that was completed. Please compare it with the fuel failure AOR and explain the relevance of less than 4 seconds of overall time in DNB.

APS Response 4:

The non-proprietary version of the response to RAI 4 is provided in Attachment 2 of this enclosure and the proprietary version is provided in Attachment 1 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

NRC Staff Request 5:

Additional Description of SKBOR Palo Verde has proposed to use the SKBOR code to determine the time available prior to precipitation of boric acid following a postulated LOCA affecting one of the reactor coolant system cold legs. The use of SKBOR represents a change to the original methodology for analyzing long-term core cooling that is described in topical report CENPD-254-P-A. A detailed description of the SKBOR methodology was not included in the license amendment request (LAR). Therefore, please submit documentation concerning the following:

a. A technical description of the SKBOR code.

b. A description of the post-processing steps (e.g., using NSAPLOT) to determine additional parameters such as the void distribution, loop differential pressure, and hot leg entrainment criteria.

c. A description of how the boric acid concentration of the sump fluid is determined.

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR APS Response 5:

The non-proprietary version of the response to RAI 5 is provided in Attachment 5 of this enclosure and the proprietary version is provided in Attachment 4 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

NRC Staff Request 6:

Thermal Conductivity Degradation and Radial Fall-Off Curve Penalty Palo Verde has proposed imposing a radial fall-off curve to offset the lack of explicit consideration of thermal conductivity degradation (TCD) in the fuel performance models in the FATES3B and STRIKIN II codes.

a. Please provide technical justification that the proposed allowance for TCD is adequate for the full set of analyzed events within Palo Verde's licensing basis (e.g., by comparing the results calculated by FATES3B and STRIKIN-II against those of a fuel performance code that explicitly models TCD and has been reviewed by the NRC staff).

b. Please provide confirmation that the actual radial fall-off curves implemented to ensure compliance with the TCD allowance will be included in the applicable Core Operating Limits Report submittals for PVNGS.

APS Response 6, Subpart a:

The non-proprietary version of the response to RAI 6, Subpart a, is provided in Attachment 2 of this enclosure and the proprietary version is provided in Attachment 3 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

APS Response 6, Subpart b:

The radial fall-off curve imposed to offset the lack of explicit consideration of thermal conductivity degradation (TCD) in the fuel performance models in the FATES3B and STRIKIN-II codes will be translated into a core operating limit on linear heat rate (LHR) versus cycle length. Bounding radial power fall-off (RFO) curves were generated as part of the fuel performance analysis supporting the implementation of NGF fuel at PVNGS. These bounding RFO curves have incorporated a fuel temperature allowance for TCD to maintain margin in the safety analyses. There is no direct control room indication of radial fall-off, and as such it is not a parameter that the operators can monitor. As part of the reload process, the actual cycle-specific RFO curves are assessed against the bounding RFO curves and the core operating limit on LHR is reduced in accordance with the reload process to ensure that the cycle specific RFO curves are enveloped by the bounding RFO curves at all burnups. As such, the RFO curves themselves will not be in the Core Operating Limits Report (COLR), instead the COLR section 3.2.1, Linear Heat Rate (LHR), will reflect the TCD allowance. The resulting COLR section 3.2.1 LHR limit details will be included in the COLR submittals for PVNGS.

In the LAR submittal, APS proposed a commitment to impose a fuel centerline temperature allowance discussed in the preceding paragraph to account for TCD effects. This commitment will be replaced with the proposed license condition as indicated below.

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Commitment as submitted in the NGF LAR (Reference 11 A fuel centerline temperature allowance at high burnup will be set aside to account for the burnup dependent effects of Thermal Conductivity Degradation (TCD) when using the FATES3B code to determine input for NGF non-LOCA and LOCA safety analyses.

Proposed License Condition A fuel centerline temperature allowance at high burnup, as specified in Attachment 8 of the enclosure to APS letter 102-07277, dated July 1, 2016, will be set aside to account for the burnup dependent effects of Thermal Conductivity Degradation (TCD) when using the FATES3B code to determine input for NGF non-LOCA and LOCA safety analyses.

The proposed license condition will remain in place until APS implements the NRC approved new long-term fuel evaluation model and associated methods that explicitly account for TCD that are applicable to PVNGS. APS is proposing the foilowing new commitment to support a transition to a new modei and methods:

Proposed Commitment to Transition to New Long Term Fuel Evaluation Model Upon NRC approval of a new long-term fuel evaluation model and associated methods that explicitly account for thermal conductivity degradation (TCD) that are applicable to PVNGS, APS will, within 6 months, provide and implement a schedule for reanalysis using the NRC approved new long term fuel evaluation model that is applicable to the PVNGS NGF design for affected licensing basis analysis.

NRC Staff Request 7:

Appendix K LOCA LOOP Considerations Palo Verde has assumed that the limiting results for evaluating the large-break LOCA event would occur when offsite power is unavailable. The availability of offsite power would result in earlier ECCS pump start times than considered in the LAR submittal. While it is possible that earlier ECCS pump start times may tend to refill the downcomer more rapidly (thereby promoting an earlier reflooding of the reactor core), it is also possible that the downcomer may continue to be refilled largely by the safety injection tanks, even if earlier ECCS pump start times are implemented. In the latter case, the earlier spilling of ECCS coolant into containment would tend to produce a more severe containment pressure reduction, and hence offer increased resistance to reflooding the core. The net impact of these countervailing tendencies on the results of the large-break LOCA analysis is not obvious; in particular, the NRC staff notes the counterintuitive observation that, according to the current, conservative Appendix K evaluation methodology, large-break LOCA scenarios with full availability of the ECCS are calculated to be more limiting than cases with a single failure that would reduce ECCS flow. Therefore, please provide the results of an additional analysis of the large-break LOCA event with offsite power available and realistic pump start times to confirm whether the results are bounded by the analysis presented in the LAR submittal.

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR APS Response 7:

The non-proprietary version of the response to RAI 7 is provided in Attachment 2 of this enclosure and the proprietary version is provided in Attachment 1 of this enclosure. of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

NRC Staff Request 8:

Statistical Treatment of the Inadvertent Fuel Misloadina Event Palo Verde has proposed to implement a statistical methodology for treating the event described in Section 15.4.7 of its Updated Final Safety Analysis Report (UFSAR),

"Inadvertent Loading of a Fuel Assembly into the Improper Position." The proposed methodology has not been previously reviewed and approved by the NRC staff and was not sufficiently described in the LAR.

a. Either please provide an adequate description and justification for the proposed approach or confirm that Palo Verde will continue to use current licensing-basis methods to analyze this event.

b. If fuel failure is predicted for the UFSAR Section 15.4.7 event "Inadvertent Loading of a Fuel Assembly into the Improper Position," please confirm that the consequences of the event are bounded by those of other analyzed events within the applicable event category, and hence, are a small fraction of Title 10 CFR 100 limits.

Further please clarify the failure mode of the affected fuel rods and explain whether the failure mechanism could propagate to neighboring rods.

APS Response 8, Subpart a:

APS is revising the approach in evaluating the inadvertent loading of a fuel assembly into the improper position (fuel mislead) and will not use the statistical methodology as presented in the NGF LAR (Reference 1). The departure from using the statistical methodology was discussed with the NRC staff during the audit held on March 8, 2017 (ADAMS Accession No. ML17102A400). APS will evaluate the fuel mislead event with a method that is consistent with the methodology described by the NRC in the Standard Review Plan (SRP), NUREG-0800 (Reference 3). Pursuant to 10 CFR 50.71(e), the PVNGS Updated Final Safety Analysis Report (UFSAR) Section 15.4.7, Inadvertent Loading of a Fuel Assembly into the Improper Position, will be updated to be consistent with the new analysis following NRC approval of the NGF LAR (Reference 1).

As a result of the change in methodology, APS is revising in its entirety section 7.4.7 of /8 to the enclosure of the NGF LAR (Reference 1) as Indicated below.

NGF LAR Attachment 7/8 to the Enclosure. Section 7.4.7 Event is reanalyzed.

UFSAR Section 15.4.7 describes the inadvertent loading of a fuel assembly into the improper position event (fuel mislead event). The fuel mislead infrequent incident event is defined by the interchanging of any two fuel assemblies in the core. The worst mislead is a mislead that results in the greatest number of assemblies

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR containing a radial peaking factor (Fr) above the Fr threshold (i.e., not necessarily the highest Fr) for fuel failure by Departure from Nucleate Boiling (DNB).

In determining the worst misload, multiple core designs and times-in-life during cycle depletion were considered. As a conservative simplification, all rods in an assembly were treated as failed if a single rod in the assembly exceeded the Fr threshold.

Per Section 15.4.7 of the SRP for the fuel misload event, fuel failure is permitted provided the offsite radiological dose consequences are limited to a small fraction, or 10%, of 10 CFR Part 100 guideline values. The PVNGS UFSAR contains two other infrequent incident events with the same offsite radiological dose consequence limit.

These are the Inadvertent Opening of a Steam Generator Atmospheric Dump Valve plus Loss of Power (UFSAR 15.1.4.5.3), and Anticipated Operational Occurrence (AOO) from the Specified Acceptable Fuel Design Limit (SAFDL) (UFSAR 15E.5.3).

The results of the limiting infrequent incident event (AOO from the SAFDL) currently bound the offsite radiological consequences of the worst fuel misload event.

In addition, RCS Activity is monitored per PVNGS Technical Specification (TS) 3.4.17, RCS Specific Activity. Any increase in activity indicative of fuel failure will be noticed.

The timing of the fuel failures In a fuel misload event Is such that there would be coolant activity indications prior to any significant offsite doses occurring. At the beginning of core life (BOL), any fuel failures associated with a fuel misload event would likely be in high reactivity fresh fuel; these fuel rods would not have large inventories of fission products. This activity would be detected by TS 3.4.17 monitoring before it reached significant levels. Towards middle of core life (MOL),

any fuel rod failures would occur gradually as power increased in the misload location (e.g., due to burnabie poison depletion). At MOL, RCS activity would build up gradually and be detected prior to reaching significant levels.

The core designs used for the fuel misload event analysis represented a wide range of PVNGS loading patterns to accommodate cycle-by-cycle core design variations and bound the event. In order to address DNB propagation, all rods in an assembly were conservatively treated as failed if a single rod in the assembly exceeded the Fr threshold.

APS Response 8, Subpart b:

The radiological consequences of the fuel misload event were evaluated in comparison to the limiting AOO from the SAFDL event. The evaluation determined that radiological inputs for a fuel misload event are either bounded by or the same as the radiological inputs previously reported to the NRC for the AOO from SAFDL event in UFSAR 15E.5. Therefore, the fuel misload event radiological consequences are within the SRP acceptance criteria of a small fraction of 10 CFR 100 limits. See APS Response 8, Subpart a, for further discussion of the radiological consequences of the fuel misload event.

8

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR NRC Staff Request 9:

Containment Analyses

a. In Section 9.1 of Attachment 8, Mass and Energy Release Analysis for Postulated LOCAs, it is stated that an evaluation of the impact of NGF on the LOCA Mass and Energy (M&E) AORs was performed. Additionally, a comparison of fuel parameters and operating conditions was performed. Please describe how the containment LOCA M&E release was determined for the NGF analyses. To confirm that the AOR LOCA M&E short term (UFSAR Table 6.2.1-4 and 6.2.1-5) and long term (from the end of post reflood) releases remain bounding, please provide quantitative results comparing to the AOR M&E releases for the following:

i. Short-term M&E release during blowdown. Also, please confirm that the AOR containment pressure response for peak pressure determination and containment temperature response for equipment environmental qualification remain bounding.

a. Long-term M&E release for the sump temperature response. Also, please confirm that the AOR sump temperature profile for the ECCS pumps net positive suction head analysis remain bounding.

b. In Section 9.2 of Attachment 8, M&E Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment, it is stated that the AOR FW temperature bounds the NGF temperature. What is the FW temperature used for AOR and the NGF analysis? Please explain how the AOR temperature produces bounding results.

c. In Section 9.2 of Attachment 8, it is stated that the M&E source energy based on NGF operating conditions will remain bounded by the AOR main steam line break (MSLB) source energy. Please Justify quantitatively that the parameters that determine the AOR MSLB containment M&E source energy bound those that determine the M&E source energy with NGF.

APS Response 9, Subparts a.i, b, and c:

The non-proprietary version of the response to RAI 9, Subparts a.i, b, and c is provided in of this enciosure and the proprietary version is provided in Attachment 1 of this enclosure. Attachment 6 of this enclosure provides the basis for the elements of the response to be withheld from public disclosure pursuant to the criteria of 10 CFR 2.390.

APS Response 9, Subpart a.ii:

The containment sump temperature profile for the ECCS pumps net positive suction head evaluation is dependent on the Loss of Coolant Accident (LOCA) mass and energy releases.

As documented in Attachments 1 and 2 of this enclosure, Westinghouse performed an evaluation comparing the analysis of record inputs to the changes required for NGF and concluded that the current analysis of record remained applicable. Therefore, the current analysis of record for the containment sump temperature profile for the ECCS pumps net positive suction head analysis is applicable to NGF and remains unchanged and bounded.

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR References

1. APS letter number 102-07277, License Amendment Request and Exemption Request to Support the Implementation of Next Generation Fuel, dated July 1, 2016 [ADAMS Accession No. ML16188A332]

2. NRC correspondence to APS, Palo Verde 1, 2, and 3 - NGF LAR and Exemption RAIs (CAC Nos. MF8076 to MF8081), dated April 14, 2017 [ADAMS Accession Number ML17107A005]

3. NUREG-0800, Revision 2, Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants: LWR Edition, dated June 1987

4. Westinghouse memo MT-17-56, Westinghouse responses to the NRC RAIs on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (ADAMS Accession No ML17107A005), dated May 15, 2017 p-

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Attachment 1 Westinghouse Responses to the NRC RAIs on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (Proprietary)

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Attachment 2 Westinghouse Responses to the NRC RAIs on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1 June 1,2017 Attachment 2 Westinghouse responses to the NRC RAIs on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (Accession Number ML17107A005)

(Non-Proprietary)

Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, PA 16066

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 2 of 34 RAI1. CENPD-178-P-A Methodology The CENPD-178-P-A methodology referenced in the submittal specifies that a hydraulic shaker attaches to the bottom of the assembly during the forced vibration fuel assembly testing. The testing facility contains an apparatus that uses an electro-mechanical shaker that attaches to the center of the assembly. Please clarify which shaker is used and justify its use as part of the methodology.

Response

The standard Westinghouse fuel assembly forced vibration test technique was used for CE16NGF. In the standard Westinghouse test, the fuel assembly is excited near its center rather than at its lower end, and the excitation is applied by an electro-mechanical shaker rather than by a hydraulic shaker. The purpose of the forced vibration test is to obtain fuel assembly natural frequencies, mode shapes, and damping, and the Westinghouse forced vibration test did provide those results for CE16NGF. Application of the test results in seismic/loss of coolant accident (LOCA) analyses was in the normal manner described in Reference 1.

References:

1. Combustion Engineering Mechanical Design Report, CENPD-178-P, Rev. 1-P, Structural Analysis of Fuel Assemblies for Seismic and Loss of Coolant Accident Loading, August 1981.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 3 of 34 RAI 2. Transition Core Control Element Assembly Drop Times Please provide justification that the full core Next Generation Fuel (NGF) and full core Value Added Fuel (referred to as STD) control element assembly (CEA) drop time analyses bound the CEA drop times expected for transition cores containing both NGF and STD fuel assemblies.

Response

A scram analysis was performed for a full core of STD fuel and a full core of NGF. The results were benchmarked using measured data from all three Palo Verde Plants. The benchmarked results were then plotted for comparison with the Safety Analysis CEA insertion curve. The curves for both fuel designs are [

]**' used for the safety analysis. The difference in computed scram time for the two fuel designs is

[

p' for any transition core containing both NGF and STD fuel assemblies. Therefore, the CEA drop time analyses performed bound the CEA drop times expected for transition cores.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1, 2017 Page 4 of 34 RAl 3. Bounded and Non-Impacted Chapter 15 Events In Section 7 of Attachment 8, Non-Loss-of-Coolant-Accident (LOCA) Safety Analysis, Table 7-1 lists the impact of the use of NGF on Chapter 15 Non-LOCA events.

a. Transition to NGF fuel is determined to have no impact on a number of Chapter 15 events, listed below. For these events, please explain the process for determining that the inputs are unchanged and justify why they are unchanged. If any of the input has changed, please Justify that the event is not impacted.

Startup of an Inactive Reactor Coolant Pump

Inadvertent Deboration

Inadvertent Operation of the Emergency Core Cooling System (ECCS)

Steam Generator Tube Rupture

b. A number of Chapter 15 events, listed below, are determined to be bounded without specific Justification. Please Justify that these events are bounded. Please identify the bounding assumptions and Justify that they are appropriate. If the bounded event has been quantitatively analyzed, please provide the margin between the new NGF analysis and the bounding analysis. For comparison, please also provide the analogous margin associated with the current bounded STD analysis of record (AOR) for these events.

Decrease in Feedwater (FW) Temperature

Increase in Main FW Flow

Loss of External Load

Turbine Trip

Main Steam Isolation Valve Closure

Loss of Non-emergency AC Power to the Station Auxiliaries

Loss of Normal FW Flow

Chemical and Volume Control System Malfunction - Pressurizer Level Control System Malfunction with Loss of AC Power (LOP)

Response

a. The key transient plant and physics data for each analysis was compared to the inputs considering the implementation of Next Generation Fuel (NGF). The following events showed no change in the inputs used within the analysis with the implementation of NGF.

Startup of Inactive Reactor Coolant Pump The startup of inactive reactor coolant pump (SIRCP) is examined in Modes 3 through 6, since plant operation with less than four RCPs running is only permitted in these modes. The introduction of NGF does not impact the maximum plant heatup and cooldown limits, the definition of the operating modes, or the reactivity condition (critical/subcritical). Thus, there are no changes to the maximum RCS heatup limit, the maximum cooldown limit, and the reactivity condition for Modes 3 through 6. Additionally, the introduction of NGF does not impact Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 5 of 34 the most positive isothermal temperature coefficient (ITC), the most negative ITC or the bounding minimum stuck rod.

The introduction of NGF does not affect the major inputs to the event or affect the maximum temperature change during the SIRCP when used with the most conservative ITCs and therefore does not result in a loss of subcriticality. Therefore, the conclusions of the SIRCP event (UFSAR Section 15.4.4.4) remain valid and are not impacted by the implementation of NGF.

Inadvertent Deboration (IBP)

The introduction of NGF does not impact the following key plant input data:

The definition of operating mode for reactivity conditions and cold leg temperature limits

The limiting condition for operation temperatures and pressure limits

The shutdown margin

The number of operating charging pumps and the maximum flow per charging pump

The maximum allowed steam generator tube plugging

The maximum CEA withdrawal rate

The minimum refueling water boron concentration The introduction of NGF results in a smaller fuel rod diameter versus the standard fuel rod diameter. The smaller NGF fuel pin diameter results in a small increase in core volume, which results in a very small increase in total reactor coolant system (RCS) volume. Flowever, any increase in RCS volume results in an increase in the boron dilution time constant and a decrease in the rate of criticality. Therefore, the IBD event with the standard fuel RCS total volume bounds the NGF RCS total volume.

The introduction of NGF does not impact the key physics input parameters (e.g., critical boron concentration and Inverse boron worth). Inverse boron worth is a function of critical boron concentration, initial shutdown margin, RCS mass, charging flow, and the time interval to criticality. Since all of these inputs are not impacted by the introduction of NGF, the inverse boron worth is not impacted by the introduction of NGF.

The introduction of NGF does not affect the major inputs to the event. Therefore, the conclusions of the Inadvertent Deboration event (UFSAR Section 15.4.6.5) remain valid and are not impacted by the implementation of NGF.

Inadvertent Operation of the Emergency Core Cooling System fECCSJ The Inadvertent Operation of the ECCS event is assumed to actuate the two high pressure safety injection (HPSI) pumps and open the corresponding discharge valves. Inadvertent operation of the ECCS is only of consequence when the event occurs below the HPSI pump shutoff head. For the non-LOCA evaluation, the initial RCS pressure remains above the HPSI pump shutoff head.

The key non-LOCA transient plant parameter for the Inadvertent Operation of the ECCS event is the HPSI pump flow versus pressure curve. Implementation of NGF does not impact this HPSI pump flow versus pressure curve.

Therefore, there is no non-LOCA transient analysis impact due to the introduction of NGF.

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June I, 2017 Page 6 of 34 Therefore, the conclusions of the Inadvertent Operation of the Emergency Core Cooling System event (UFSAR Section 15.5.1.4) remain valid and are not impacted by the implementation ofNGF.

Steam Generator Tube Rupture (SGTRt The introduction ofNGF does not impact the following key SGTR event input; The initial transient input conditions for core power, reactor coolant pump (RCP) heat, cold leg temperature, pressurizer pressure, and mass flow The RPS setpoints and response times The engineered safety feature actuation system setpoints and response times for the safety injection actuation system, auxiliary feedwater actuation system, auxiliary feedwater (AFW)-lockout, and main steam isolation signal The AFW maximum enthalpy/temperature The MSSV setpoints and valve characteristics The radiological dose parameters for primary and secondary side coolant specific activities, primary-to-secondary leakage, and atmospheric dispersion/dilution factors; pressurizer level and heat minimum capacity The maximum refueling water temperature The shutdown cooling initiating temperature The minimum letdown flow The number of operating charging pumps, and the maximum flow per charging pump The emergency procedures for the SG tube (coverage strategy) procedure and functional recovery procedure The SGTR events are not impacted by the generic physics data because the event is a slow depressurization event and because the loss-of-power occurs three seconds after reactor trip. Thus, there is no impact on the SGTR due to the physics parameters. Hence, for the SGTR events, the introduction ofNGF has no impact.

As the introduction ofNGF does not affect the inputs to or the procedures utilized by the event as listed above, the conclusions of the Steam Generator Tube Rupture events (UFSAR Sections 15.6.3.1.2 and 15.6.3.2.6) remain valid and are not impacted by the implementation ofNGF.

b. The following events are not explicitly analyzed for either STD fuel or NGF as they are bounded by the results of other events as described below. As such, no comparison to quantitative margins is performed for these events.

Decrease in Feedwater (FWl Temperature The decrease in main FW temperature event would result in a smaller decrease in RCS temperature than an increase in main steam flow event involving the quick opening of eight Steam Bypass Control System (SBCS) valves or an inadvertent opening of a steam generator atmospheric dump valve (OSGADV) (See UFSAR Section 15.1.3 and UFSAR Section 15.1.4). The smaller RCS cooldown results in less of a power increase, and hence less of a decrease in the minimum hot channel departure from nucleate boiling ratio (DNBR) during the Methods, Technology. & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 7 of 34 transient. The minimum hot channel DNBR establishes whether a fuel design limit has been exceeded and therefore whether fuel cladding degradation might be anticipated.

For the decrease in main FW temperature event in combination with a single failure, the parameter of concern is likewise the minimum hot channel DNBR. Factors that would cause a decrease in DNBR include an increase in coolant temperature, a decrease in coolant pressure, an increase in local heat flux (including radial and axial power distributions effects), and a decrease in coolant flow rate. Evaluation of postulated single failures shows that the worse single failure for this event is a Loss of Offsite Power (LOP) following a turbine trip, which would cause the RCPs to coast down and rapidly reduce the coolant flow rate. This event, however, would result in an NSSS response that is similar to, but less severe than, that caused by the increase in main steam flow event involving the quick opening of eight SBCS valves or an inadvertent OSGADV in combination with LOP (see UFSAR Section 15.1.3 and 15.1.4). These events result in more severe RCS cooldown that in turn results in more of an increase in power, and hence more of a decrease in the minimum hot channel DNBR. Therefore, the DNBR at the moment the RCPs begin to coastdown would be bounded by those events. For this reason, the infrequent decrease in the FW temperature event (in combination with a single failure) is bounded by the infrequent event involving the quick opening of eight SBCS valves and the inadvertent OSGADV (in combination with single failure) with respect to the DNBR specified acceptable fuel design limit (SAFDL).

In addition, this event would result in a more benign minimum DNBR than the results from the limiting infrequent event that is described in the UFSAR Appendix I5.E. The event described in the UFSAR Appendix 15.E establishes a limiting infrequent event, including all incidents of moderate frequency in combination with a single failure, with respect to DNBR degradation, assuming that the DNBR is already at the SAFDL when the single failure (LOP) occurs.

With respect to the RCS pressure boundary performance, a decrease in the main FW temperature is characterized by an initial cooldown of the primary and secondary systems, and decreasing RCS and steam generator pressures. If the event results in a reactor trip and main steam isolation signal (MSIS), repressurization of the RCS and steam generators would occur due to decay heat from radionuclides in the core, the heat stored in the metal structures of the NSSS, and the heat from any operating RCPs. Additionally, if pressurizer pressure decreases below the safety injection actuation signal (SIAS) setpoints, safety injection flow may also result in repressurization of the RCS. Eventually, however, plant operators would take action to cooldown and depressurize the plant to shutdown cooling entry conditions. This may be accomplished by feeding the steam generators with auxiliary feedwater flow and by releasing steam through ADVs.

The subsequent heatup and repressurization of the NSSS would not challenge RCS pressure boundary peak pressure limits. Prior to the operators taking action to cool down the plant, the secondary system peak pressure would be limited by the main steam safety valves (MSSVs), which have sufficient capacities to relieve the steam that may be generated by NSSS heat sources. Furthermore, if the heat transfer rate from the RCS to the secondary system were degraded for any reason, as might occur when a LOP results in a loss of forced RCS coolant flow, the pressurizer safety valves (PSVs) may also open to limit the RCS peak pressure. Because the maximum allowable lift settings for the MSSVs and PSVs are well below the peak pressure regulatory limits for this event, a decrease in feedwater temperature, or a decrease in feedwater temperature in combination with a single failure, would not challenge the RCS pressure boundary through overpressurization of either the primary or secondary systems.

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 8 of 34 The offsite and control room radiological dose consequences associated with this infrequent event are bounded by those that may result from an Inadvertent OSGADV with a LOP event (see UFSAR Section 15.1.4) and/or the limiting infrequent event (see UFSAR Appendix 15.E), and comply with regulatory guidelines.

The introduction of NGF does not affect the basis of the decrease in main FW temperature event. It remains being bounded by the more severe consequences of: the increase in main steam flow event involving the quick opening of eight Steam Bypass Control System (SBCS) valves, or an inadvertent opening of a steam generator atmospheric dump valve (OSGADV), or the limiting infrequent event (See UFSAR Section 15.1.3, UFSAR Section 15.1.4, and UFSAR Appendix 15.E).

Increase in Main FW Flow The increase in main FW flow event would result in a smaller decrease in RCS temperature than an increase in main steam flow event involving the quick opening of eight Steam Bypass Control System (SBCS) valves or an inadvertent opening of a steam generator atmospheric dump valve (OSGADV) (See UFSAR Section 15.1.3 and UFSAR Section 15.1.4). The smaller RCS cooldown would result in less of a power increase, and hence less of a decrease in the minimum hot channel DNBR during the transient. The minimum hot channel DNBR establishes whether a fuel design limit has been exceeded and therefore whether fuel cladding degradation might be anticipated.

For the increase in main FW flow event in combination with a single failure, the parameter of concern is likewise the minimum hot channel DNBR. Factors that would cause a decrease in DNBR include an increase in coolant temperature, a decrease in coolant pressure, an increase in local heat flux (including radial and axial power distributions effects), and a decrease in coolant flow rate. Evaluation of postulated single failures shows that the worse single failure for this event is a Loss of Offsite Power (LOP) following a turbine trip, which would cause the reactor coolant pumps (RCPs) to coast down and rapidly reduce the coolant flow rate. This event, however, would result in an NSSS response that is similar to, but less severe than, that caused by the increase in main steam flow event involving the quick opening of eight SBCS valves or an inadvertent OSGADV in combination with LOP (see UFSAR Section 15.1.3 and 15.1.4). These events result in more severe RCS cooldown that in turn results in more of an increase in power, and hence more of a decrease in the minimum hot channel DNBR. Therefore, the DNBR at the moment RCPs begin to coastdown would be bounded by those events. For this reason, the infrequent increase in the FW flow event (in combination with a single failure) is bounded by the infrequent event involving the quick opening of eight SBCS valves and the inadvertent OSGADV (in combination with single failure) with respect to the DNBR SAFDL.

In addition, this event would result in a more benign minimum DNBR than the results from the limiting infrequent event that is described in the UFSAR Appendix 15.E. The event described in the UFSAR Appendix 15.E establishes a limiting infrequent event, including all incidents of moderate frequency in combination with a single failure, with respect to DNBR degradation, assuming that the DNBR is already at the SAFDL when the single failure (LOP) occurs.

With respect to the RCS pressure boundary performance, an increase in the main FW flow is characterized by an initial cooldown of the primary and secondary systems, and decreasing RCS and steam generator pressures. If the event results in a reactor trip and main steam isolation signal (MSIS), repressurization of the RCS and steam generators would occur due to decay heat from radionuclides in the core, the heat stored in the metal structures of the NSSS, and the heat from any operating RCPs. Additionally, if pressurizer pressure decreases below the Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 9 of 34 safety injection actuation signal (SIAS) setpoints, safety injection flow may also result in repressurization of the RCS. Eventually, however, plant operators would take action to cooldown and depressurize the plant to shutdown cooling entry conditions. This may be accomplished by feeding the steam generators with auxiliary feedwater flow and by releasing steam through ADVs.

The subsequent heatup and repressurization of the NSSS would not challenge RCS pressure boundary peak pressure limits. Prior to the operators taking action to cool down the plant, the secondary system peak pressure would be limited by the main steam safety valves (MSSVs), which have sufficient capacities to relieve the steam that may be generated by NSSS heat sources. Furthermore, if the heat transfer rate from the RCS to the secondary system were degraded for any reason, as might occur when a LOP results in a loss of forced RCS coolant flow, the pressurizer safety valves (PSVs) may also open to limit the RCS peak pressure. Because the maximum allowable lift settings for the MSSVs and PSVs are well below the peak pressure regulatory limits for this event, an increase in feedwater flow, or an increase in feedwater flow in combination with a single failure, would not challenge the RCS pressure boundary through overpressurization of either the primary or secondary systems.

The offsite and control room radiological dose consequences associated with this infrequent event are bounded by those that may result from an Inadvertent OSGADV with a LOP event (see UFSAR Section 15.1.4) and/or the limiting infrequent event (see UFSAR Appendix 15.E), and comply with regulatory guidelines.

The introduction of NGF does not affect the basis of the increase in main FW flow event. It remains being bounded by the more severe consequences of: the increase in main steam flow event involving the quick opening of eight Steam Bypass Control System (SBCS) valves, or an inadvertent opening of a steam generator atmospheric dump valve (OSGADV), or the limiting infrequent event (See UFSAR Section 15.1.3, UFSAR Section 15.1.4, and UFSAR Appendix 15.E).

Loss of External Load The results of the loss of load event are no more limiting with respect to RCS pressurization than those of the loss of condenser vacuum (LOCV) event presented in UFSAR Section 15.2.3. The LOCV also results in a turbine trip; however, feedwater flow is assumed to terminate following LOCV whereas it is assumed to ramp down to 5% following the loss of load. This larger reduction in heat removal capability results in a higher peak RCS pressure for the LOCV.

Like the LOCV, the DNBR increases during the loss of load due to the increasing pressure. Thus, the initial DNBR is also the minimum DNBR. For the loss of load, due to its similarity with the LOCV event, there are no concurrent single failures which, when combined with the loss of external load, result in consequences more severe than the LOCV event with respect to RCS pressurization. The limiting single failure with respect to fuel performance is the loss of offsite power following a turbine trip. This event with a loss of offsite power results in an event similar to the loss of flow (LOF) event discussed in UFSAR Section 15.3.1. Results of the LOF event are directly applicable to the loss of external load with loss of offsite power following a turbine trip.

The introduction of NGF does not affect the basis of the Loss of External Load event. The event remains bounded by the more severe consequences of the LOCV event (UFSAR Section 15.2.3) as the LOCV results in a turbine trip and also assumes feedwater flow is terminated following the LOCV.

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 10 of 34 Turbine Trip The results of the turbine trip event are no more limiting with respect to RCS pressurization than those of the loss of condenser vacuum (LOCV) event presented in UFSAR Section 15.2.3. The LOCV also results in a turbine trip; however, feedwater flow is assumed to terminate following LOCV whereas it is assumed to ramp down to 5% following the turbine trip. This larger reduction in heat removal capability results in a higher peak RCS pressure for the LOCV.

Like the LOCV, the DNBR increases during the turbine trip due to the increasing pressure. Thus, the initial DNBR is also the minimum DNBR. For the turbine trip, due to its similarity with the LOCV event, there are no concurrent single failures which, when combined with the turbine trip, result in consequences more severe than the LOCV event with respect to RCS pressurization. The limiting single failure with respect to fuel performance is the loss of offsite power following a turbine trip. This event with a loss of offsite power results in an event similar to the loss of flow (LOF) event discussed in UFSAR Section 15.3.1. Results of the LOF event are directly applicable to the turbine trip with loss of offsite power.

The introduction of NGF does not affect the basis of the Turbine Trip event. The event remains bounded by the more severe consequences of the LOCV event (UFSAR Section 15.2.3) as the LOCV results in a turbine trip and also assumes feedwater flow is terminated following the LOCV rather being ramped down to 5% following the turbine trip.

Main Steam Isolation Valve (MSIV) Closure The results of the MSIV closure event are no more limiting with respect to RCS pressurization than those of the LOCV event presented in UFSAR Section 15.2.3. The LOCV also results in the termination of all main steam flow. However, main steam flow is terminated more rapidly during the LOCV since the closure time for the turbine stop valves is much shorter than that for the MSlVs. The faster reduction in heat removal results in a higher peak RCS pressure for the LOCV event.

Like the LOCV, the DNBR increases during the MSIV closure event due to the increasing pressure. Thus, the initial DNBR is also the minimum DNBR for the MSIV closure event.

Due to the similarity with the LOCV event, there are no concurrent single failures which when combined with the MSIV closure event result in consequences more severe than the LOCV event with respect to RCS pressurization. The limiting single failure with respect to fuel performance is the loss of offsite power following a turbine trip. This event with a loss of offsite power results in an event similar to the loss of ac power which initiates the LOF event discussed in Section 15.3.1. Results of the LOF event are directly applicable to the MSIV closure with loss of offsite power following a turbine trip.

The introduction of NGF does not affect the basis of the Main Steam Isolation Valve Closure event. The event remains bounded by the more severe consequences of the LOCV event (UFSAR Section 15.2.3) as the LOCV results in a more rapid termination of all steam flow than the MSIV event.

Loss of Non-emergency AC power (LOAC) to the Station Auxiliaries The results of the LOAC event are identical to those of the loss of reactor coolant flow event presented in UFSAR Section 15.3.1, and are no more limiting with respect to RCS pressurization than the LOCV event Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 11 of 34 discussed in UFSAR Section 15.2.3. During the LOCV event the plant experiences simultaneous losses of steam and feedwater flow and condenser availability. In addition, the plant experiences a complete loss of forced reactor coolant flow during the LOAC event. The loss of forced reactor coolant flow results in an earlier reactor trip for the LOAC event (on low RCP shaft speed) compared to the reactor trip for the LOCV event (on high pressurizer pressure). The earlier trip promotes a less severe primary-to-secondary heat imbalance and hence a lower peak RCS pressure for the LOAC event.

The fuel performance for the LOAC is no more limiting than that for the LOF event discussed in UFSAR Section 15.3.1. The LOAC is the initiating event for the LOF so the fuel performance results of the LOF event are directly applicable to the LOAC event.

The introduction of NGF does not affect the basis of the Loss of Non-emergency AC power to Station Auxiliaries event. The event remains bounded by the more severe consequences of the LOCV event (UFSAR Section 15.2.3) as the LOAC event results in an earlier reactor trip than the LOCV event resulting in a less severe primary to secondary heat imbalance.

Loss of Normal Feedwater Flow (LFW)

The maximum RCS pressure for the LFW event is less than that for the LOCV event discussed in UFSAR Section 15.2.3. The LOCV event results in the termination of main steam flow prior to reactor trip in addition to the total loss of normal feedwater flow. This additional condition aggravates RCS pressurization by further reducing the rate of primary-to-secondary heat transfer below that of the LFW event.

Like the LOCV, the DNBR increases during the LFW event due to the increasing RCS pressure. Thus the initial DNBR is also the minimum DNBR for the LFW event.

There are no concurrent single failures that when combined with LFW result in consequences more severe than the LOCV event with respect to RCS pressurization.

The limiting single failure with respect to fuel performance is the loss of offsite power following turbine trip.

For the LFW event, prior to turbine trip the DNBR increases due to the RCS pressure increase. DNBR then briefly decreases after turbine trip due to the reactor coolant flow coastdown on loss of offsite power. The DNBR decreases similar to the DNBR transient associated with the total loss of reactor coolant flow event shown in UFSAR Section 15.3.1; however, the DNBR decrease for LFW is not as severe due to the earlier reactor trip relative to the initiation of the coolant flow coastdown. Therefore, the minimum DNBR remains above the limit.

The introduction of NGF does not affect the basis of the Loss of Normal Feedwater event. The event remains bounded by the more severe consequences of the LOCV event (UFSAR Section 15.2.3) as the LOCV event results in a greater primary to secondary heat imbalance than the loss of normal feedwater event.

Chemical and Volume Control System Malfunction - Pressurizer Level Control System Malfunction with Loss of ACPowerfLOPl The pressurizer level control system (PLCS) malfunction event produces an increasing RCS pressure that compensates for the elevated RCS temperatures, such that the available thermal margin does not degrade before the onset of the LOP. With respect to the DNBR acceptance criteria, this event is no more adverse than the total Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 12 of 34 loss of reactor coolant flow event (UFSAR Chapter 15.3.1). Thus, the overall DNBR degradation experienced during a PLCS malfunction with LOP would be bounded by the total loss of reactor coolant flow event (UFSAR Chapter 15.3.1).

Additionally, the introduction of NGF does not affect the key PLCS malfunction event with LOP input data.

There are no changes to;

The initial transient input conditions for core power, RCP heat, cold leg temperature, pressurizer pressure, and mass flow

The reactor protection system setpoints and response times

The pressurizer safety valves and main steam safety valve setpoints and valve characteristics

The pressurizer level and uncertainty

The minimum letdown flow

The number of operating charging pumps and the maximum flow per charging pump The introduction of NGF does not affect the key physics parameters (e.g., the most positive moderator temperature coefficient, beginning-of-cycle kinetics, or scram worth).

As no key event inputs are changing, the introduction of NGF does not affect the PLCS malfunction with LOP event.

In conclusion, as the events described above remain bounded by the results of other events, no comparison to quantitative margins was performed and, therefore, the margins for these events listed in the UFSAR are unchanged by the introduction of NGF.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 13 of 34 RAI 4. Evaluated Chapter 15 Events Please justify that the CENTS evaluations, completed as part of the Chapter 15 AORs, remain applicable for the transient system response, and do not require re-analysis to support the transition to NGF.

Attachment 7/8, Section 7.1.3: Please demonstrate that the system model changes due to NGF are bounded by the AOR for the Increase in Main Steam Flow (IMSF), Inadvertent Opening of a Steam Generator Atmospheric Dump Valve (lOSGADV), and lOSGADV with LOP events. Please demonstrate that the AOR flow coastdown remains conservative. Please identify what system model changes are necessary because of NGF, and Justify why they are small. Please demonstrate that these small changes result in an insignificant impact to the overall transient system.

Attachment 7/8, Section 7.1.3: Please demonstrate that the current analysis for calculated fuel failure (i.e.

AOR Departure from Nucleate Boiling Ratio (DNBR) vs. NGF DNBR) bounds the IMSF+LOP and lOSGADV+LOP events.

Attachment 7/8, Section 7.3.4: Please identify the fuel specific failure analysis methodology and describe the fuel failure analysis that was completed. Please compare it with the fuel failure AOR and explain the relevance of less than 4 seconds of overall time in DNB.

Response

a. The system response impact of the CENTS (WCAP-15996-P-A) input model due to NGF is isolated to the inputs to the thermal hydraulics of the core model. CENTS uses a simple core model that does not model individual fuel rods but instead models the overall Impact of the core in CENTS node-flowpath system representation. The major input that impacts these calculations is the core pressure drop which will modify the loss factors and therefore the calculated change in flowrates in CENTS.

The other CENTS inputs that would be impacted by the implementation of NGF would be the core flow area increase due to NGFs smaller fuel rod diameter. This would impact the nominal flow conditions in the RCS. However, the non-LOCA safety analyses are initiated at the technical specification RCS flow limits that did not change for NGF. Additionally, due to the lumped nature of the CENTS node-flowpath system representation, the small change in core flow area has an insignificant impact on the CENTS system response.

With regards to the core pressure drop changes in CENTS, as most of the Non-LOCA transients consider a constant flowrate these changes in loss factors would have no impact on the system response. Therefore, the impact of the change in core pressure drop would be isolated to those transients that have a large change in reactor coolant system flowrate (i.e., transients that consider a loss-of-power and subsequent coastdown of the RCPs). However, the CENTS RCP inputs for the pre-NGF configuration were selected to generate a conservatively faster RCP coastdown than what was generated with using the COAST (CENPD-98-A) code.

When this coastdown was compared to the NGF specific RCP calculated coastdown, it was concluded that the conservative selection of inputs for the pre-NGF configuration yielded a more conservative (faster) coastdown and would thereby bound the results if the NGF configuration was directly modeled. See Figure 1 for a comparison of the CENTS calculated flow coastdowns between the pre-NGF and NGF configurations.

Methods, Technology, & Licensing Memo Template Version I -0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 14 of 34 Figure 1: CENTS Calculated Normalized Flow Coastdown Comparison

b. The impact of NGF on the CENTS system model changes are detailed in the response to RAI 4.a and are applicable to the increase in main steam flow (IMSF), inadvertent opening of a steam generator atmospheric dump valve (lOSGADV), and ISOGADV with Loss of Offsite Power (LOP) events. The system response impact of the CENTS input model due to NGF is isolated to the inputs affecting the thermal hydraulics of the core model. CENTS uses a simple core model that does not model individual fuel rods but instead models the overall impact of the core in CENTS node-flowpath system representation. The major input that affects these calculations is the core pressure drop that will modify the loss factors and, therefore, the calculated change in flowrates in CENTS. However, it was shown that the conservative selection of reactor coolant pump inputs of the CENTS coastdown for the pre-NGF input model yielded a conservative response when compared to the NGF specific input model (see Figure 1).

The other CENTS inputs that would be impacted by the implementation of NGF would be the core flow area increase due to NGFs smaller fuel rod diameter. This would affect the nominal flow conditions in the RCS.

However, the Non-LOCA safety analyses are initiated at the technical specification RCS flow limits that did not change for NGF. Additionally, due to the lumped nature of the CENTS node-flowpath system representation, the small change in core flow area has an insignificant impact on the CENTS system response.

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 15 of 34 Per UFSAR Section 15.1.4.5.2, Bullet 15, a bounding fuel failure fraction of 5.5% is assumed in the radiological dose analysis for the lOSGADV+LOP event. Per UFSAR Section 15.1.3.5, the IMSF+LOP radiological dose consequences are bounded by those that may result from the lOSGADV+LOP event. This conclusion is not impacted by the introduction of NGF. The NGF calculated fuel failure levels for the IMSF+LOP and lOSGADV+LOP events are [ 1*' respectively. These results remain under the conservative fuel failure assumption of 5.5% used in the lOSGADV+LOP radiological dose analysis; therefore, the NGF results are bounded by the current analysis.

Per Section 5.7 of the PVNGS Licensing Amendment Request (LAR), the seized rotor and sheared shaft ,

accidents are classified as Condition IV events. DNBR calculations are performed to quantify the inventory of rods that would undergo DNB and conservatively be presumed to fail. The [

I*"* approach was applied to execute a steady state core thermal hydraulic VIPRE-W model where the boundary conditions provided by the system transient analyses [

1** The rods in DNB calculation of the subject events indicate that the current UFSAR Dose analyses are satisfied. This process is consistent with the current NRC approved licensed method to calculate fuel failure for loss of flow accidents (Section 3.2.2.1 of CENPD-183-A).

The less than 4 seconds of overall time in DNB refers to the time in DNB that ensures the bounding fuel clad strain evaluation remains applicable and thereby precludes DNB propagation. Per UFSAR Section 15.3.4, Bullet D, DNB propagation is evaluated by verifying that the bounding fuel clad strain evaluation is still applicable. It was determined that the minimum time in DNB required to reach the NRC imposed strain limit of 29.3% (CEN-372-P-A) is 4.5 seconds over a range of conditions. An analysis was performed to determine the minimum time to reach the strain limit of 29.3% for the PVNGS NGF design under the following conditions:

Heat Flux: 2.0-7.00 MBtu/hr-ft2

Mass Flux: 1.4-3.5 Mlbm/hr-fl2

Quality:-0.2-0.1

RCS Pressure: 1800-2300 psia

Fuel Rod Pressure: 2350-3000 psia It was determined that under these conditions the time to reach the NRC imposed strain limit of 29.3% was greater than 4.5 seconds and, therefore, the conclusion of overall time in DNB less than 4.5 seconds to preclude DNB propagation remains applicable to the PVNGS NGF design. The time in DNB for the limiting sheared shaft event was found to be less than 4 seconds, which remains under the minimum time (4.5 seconds) to reach the strain limit of 29.3%. Therefore, DNB propagation is precluded.

Methods, Technology, & Licensing Memo Template Version 1 -0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1, 2017 Page 16 of 34 RAIS. Additional Description of SKBOR Palo Verde has proposed to use the SKBOR code to determine the time available prior to precipitation of boric acid following a postulated LOCA affecting one of the reactor coolant system cold legs. The use of SKBOR represents a change to the original methodology for analyzing long-term core cooling that is described in topical report CENPD-254-P-A. A detailed description of the SKBOR methodology was not included in the license amendment request (LAR). Therefore, please submit documentation concerning the following:

a. A technical description of the SKBOR code.

b. A description of the post-processing steps (e.g., using NSAPLOT) to determine additional parameters such as the void distribution, loop differential pressure, and hot leg entrainment criteria.

c. A description of how the boric acid concentration of the sump fluid is determined.

Response

The proposed Westinghouse response to the RAl is as follows.

a. A technical description ofthe SKBOR code.

A technical description of the SKBOR code is enclosed in Attachments 4 (Proprietary) and 5 (Non-Proprietary).

b. A description ofthe post-processing steps (e.g., using NSAPLOT) to determine additional parameters such as the void distribution, loop differential pressure, and hot leg entrainment criteria.

The post-processing steps (e.g., using NSAPLOT) to determine additional parameters such as the void distribution, loop differential pressure, and hot leg entrainment criteria are described.

Void Distribution The Yeh void fraction model (References 1 through 3), as coded, in SKBOR has been simplified such that the core exit void fraction is used in the upper plenum. However, there is a significant area change above the core such that the superficial gas velocity will be reduced; hence, the void fraction will be reduced in regions above the heated core. The margin lost due to this simplification is recaptured by using the adjustment factor:

Equation 5-1 Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1, 2017 Page 17 of 34 Where; UcEx or OP ~ Adjusted void fraction in the core exit region and outlet plenum region calculated per Equation 5-1 Acex or OP = Total flow area in the core exit region or outlet plenum region (ft^)

acore = Core exit void fraction at the top of active fuel exit Acore Total flow area at the core exit at the top of active fuel (ft^)

exit Regime-specific exponent from Table 5-1 Table 5-1: Yeh Void Fraction Model Constants and Exponents Regime b Small 0.67 bubble Large 1.0 < -^ < 4.31 0.47 bubble

bcr

>4.31 0.393

'bcr The superficial gas velocity, jg, is obtained from the time-dependent output from SKBOR. The critical bubble rise velocity, V^cr, 's calculated as follows:

U \ Uf L/rt lU ^^

= 1-53 Pf Where:

<7 = Surface tension (Ibf/ft)

Pf.Pg - Saturation densities for the liquid and gas phases (Ibm/ft^), respectively g = Gravitational acceleration, 32.174 ft/s^

Pc = Gravitational constant, 32.174 ft-lbm/lbf-s^

Using the time-dependent void fraction from one SKBOR case, the inner vessel mixing volume can be recalculated using the appropriate adjustment factor for regions above the core. SKBOR is then re-executed with a user supplied input array for the time-dependent mixing volume.

The time dependent mixing volume, VCORET, is calculated as follows; VCORET = LPFRAC xVLP + (l- acore] x VAC + {1- Ucex) x VCEX -I- (1 - Uop) x VOP

\ avg /

where:

LPFRAC = Fraction of lower plenum volume credited (50%)

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1, 2017 Page 18 of 34 VLP Lower plenum volume (ft^)

VAC Core free volume (ft^)

VCEX Core exit region free volume (ft^)

VOP Outlet plenum free volume (ft^)

acore Time dependent core average void fraction calculated by SKBOR avg Loop Differential Pressure Presently, the SKBOR computer code does not have the capability to evaluate the loop differential pressure to confirm that there is sufficient margin in the calculated mixing volume to account for loop pressure drop effects. The evaluation is performed as a post-processing step.

The time-dependent static head of the liquid in the downcomer and the time-dependent collapsed liquid level in the inner vessel are calculated using the time-dependent steaming rate calculated by SKBOR. This establishes the supportable loop pressure drop, ^Psupportabie the difference between the static head of the liquid in the downcomer, APp^, and the static head of the liquid in the inner vessel, APjy:

^^Supportable ~ (Eq. 5-2)

The mixing volume is justified if the following condition is met;

^Supportable >AP,Loop therefore.

Loop ^ APnr AP (Eq. 5-3)

Static Head of the Downcomer Liquid With no boiling, hence no void in the downcomer, the downcomer collapsed liquid level (CLLoc) >s equal to the bottom of the cold leg elevation referenced to the bottom of the active fuel (excluding the fuel alignment plate thickness):

CLLnr AZr Static Head of the Inner Vessel Liquid The inner vessel collapsed liquid level {CLLiy) is calculated, which will then be used with the downcomer collapsed liquid level (CLLpc) to calculate a supportable loop pressure drop. The equation shown below is valid as long as the CLL/y falls within the active fuel region, e.g.;

CLLjy < 12.5 ft The inner vessel CLL is then calculated using the following equation:

VCORET - (LPFRAC x VLP)

CLLjy ACC Where; VCORET = Time-dependent varying mixing volume, Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 19 of 34 LPFRAC = Fraction of lower plenum credited VLP = Volume of lower plenum, ACC = Core flow area, Loop Pressure Drop The loop pressure drop is calculated as follows;

^Pioop = KLOOP X ^ X (rhboij)^

Pg LPioop = KLOOP X (CONV X x where:

KLOOP = Loop loss coefficient Pg = Saturated vapor density, Ibm

^boii ~ Boil-off mass flow rate. Ibm Ibm - irP CONV = Conversion factor = 9266 Ibf- ft-s^

ACL = Reference area for KLOOP, ft^

The loop loss coefficient, KLOOP, from the outlet plenum of the RV to the inlet of the downcomer is calculated for the resistance network is depicted in Figure 5-1.

Cold Leg Hot Leg Cold Leg RV Outlet RV Downcomer Plenum Cold Leg Hot Leg Cold Leg Figure 5-1: Loop Pressure Drop Resistance Network Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 20 of 34 Assessment of Loop Pressure Drop Effects The loop pressure drop that can be supported is:

^^Supportable ~ ~

and the mixing volume used in SKBOR is justified when:

Loop ^ APnr AP;i where:

APoc - CLLdc X P/X (J) X Ibm Pf 59A j^@Psat = 20.0 psia AP/v However, as the boric acid concentration in the core increases so does the density of solution. Therefore, AP/v, is adjusted as follows:

- AP,7 X ^1 + Apso(ute(Ccore ~ Qump))

where:

^Core = core region boric acid in wt%

^Sump = sump boric acid concentration in wt%

APsoiute = density change due to solute = Vwt%

When accounting for the density change due to solute in the mixing volume, the supportable loop pressure drop, ^Psupportable becomes.

^Supportable = APnr-AP, And the loop pressure drop margin, AP^'^p^, is:

ApTnargin _ a n _ AP

'Loop ^Supportable 'Loop The mixing volume is justified when AP^^p^ > 0.

Hot Leg Entrainment Criteria The liquid film entrainment threshold in the hot leg is evaluated by applying both the Wallis-Steen liquid entrainment onset criterion (Reference 4) and the Ishii-Grolmes inception criteria (Reference 5). These entrainment correlations are valid for flow conditions where the liquid phase does not take up a significant volume of the pipe (such as in the hot legs post-LOCA) and viscous effects in the liquid are not dominant, i.e., the liquid phase is in the turbulent regime.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 21 of 34 Wallis-Steen Liquid Entrainment Onset Criterion The liquid entrainment onset correlation (Eq. 12.43 of Reference 4) can be rearranged and expressed as follows:

V2 Jg>n2 (ei]

\Pg) where U2 is the dimensionless gas velocity for onset of entrainment. Steen suggested a value of 2.46 X 10'^ for 7T2, however, a more conservative value of 2.0 X 10"* is used in the evaluation. This value of TI2 is consistent with the value used in the calculation (Reference 6) to respond to a NRC request for additional information regarding the Beaver Valley extended power uprate.

Where, at a saturation pressure of 20 psia:

- surface tension of liquid = 3.919x10 6 Ibm

= viscosity of gas = 1-1 ^ ^

Pg Ibm Pf = density of liquid = 59.4 Ibm Pg = density of gas = 0.04978

= gravitational constant = 32.174^^'^

9c Using the above properties as input, the following results are obtained for the liquid entrainment threshold in terms of superficial gas velocity in the hot leg:

/ 59.4l?

= 2.0 X 10-^

^0.04978^y ^2.618 X 10"^ j

= 103.42 The total gas mass flow rate at the entrainment threshold is calculated:

~ jg ^ ^ ^hot leg) ^ Pg ~ 99.05 where for a single hot leg, Afiot leg = 9.62 ft^

The decay heat fraction can be related to the core steam mass flow rate as follows, where PZERO is the licensed power of 4070 MWt including calorimetric uncertainty.

PZERO x(^) P\ x(___ Btu (948;-T^)

T^gas =

hfg +

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 22 of 34 P_ ^gas ^ ^^sub)

Po PZERO X (948j^)

Btu With no subcooling and hfg = 960.1 @ P^at = 20.0 psia, the decay heat fraction is calculated:

= 0.02465 Po This decay heat fraction corresponds to approximately 1100 seconds after shutdown for Appendix K decay heat. Therefore, gas flow in the hot legs should drop below the entrainment threshold at about 20 minutes based upon Appendix K decay heat.

Ishii-Grolmes Liquid Entrainment Onset Criterion The Ishii-Grolmes entrainment inception criterion has three separate regimes based on the liquid film Reynolds number, Re^, in the channel. Two of the regimes are further subdivided based on the magnitude of the liquid viscosity number, N^. Based on the hot leg injection flow rate of -415 gpm, the liquid film will not be in the low Reynolds number regime {Ref < 160). Of the two remaining regimes, transition and rough turbulent, the rough turbulent regime requires the lowest gas velocity to entrain droplets from the film as shown in Figures 1 and 3 of the paper (Reference 5). In short, in the rough turbulent regime it is easier to strip droplets off an already unstable interface; whereas, in the laminar and transition regimes, a higher gas velocity is needed to create the instability at the interface. The entrainment onset criterion for the rough turbulent regime can be applied irrespective of the liquid flow direction and is, therefore, applicable for countercurrent flow that occurs in the hot leg during simultaneous hot and cold side injection.

The liquid entrainment onset correlation per Reference 5 can be expressed as follows:

for W/i < ^ where = and Lp = Pf- Pg Where, Ng is the liquid viscosity number and ]g is the superficial velocity of the gas phase.

The equation is re-written as follows with conversion factors included to be compatible with the steam tables typically used by Westinghouse safety analysis groups:

,0.5 forlV<i fe) g = 32.174g where Ng = and /lp = pf- Pg where ft-lbm gc = 32.174 Ibf-s^

The following properties of saturated liquid and gas phases of water at 20 psia are used in the correlation:

a = surface tension of liquid = 0.003919 ^

1.711X10-*^ 5 Pf = viscosity of liquid =---------Ti-TBm ft-lbm

= 5-318 X 10 ^

32.174 Ibf-s^

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-I7-56, Rev. 1, Attachment 2 June 1,2017 Page 23 of 34 Ibm Pf density of liquid = 59.40 W

Ibm Pa density of gas = 0.04978 Using the above properties as input, the following results are obtained for the liquid entrainment threshold in terms of superficial gas velocity in the hot leg:

j = 75.62 ^ with = 6.936 x IQ-'^ < ^

Applying the value of 75.6 fl/s with comparable gas flow in each hot leg, the total gas mass flow rate at the entrainment threshold becomes:

Ibm

^gas ~ /g ^ (2 X ^hot leg') ^ Pg ~ 72.43 where for a single hot leg, igg = 9.62 ft^

The decay heat fraction can be related to the core steam mass flow rate as follows, where PZERO is the licensed power of 4070 MWt including calorimetric uncertainty.

P\ .

PZ£fiOx(^)x(948;-^) Btu \

mgas = hfg +

P ^ T^gas ^ O^fa ^^sub)

X (948 At the saturation pressure of 20 psia with no subcooling, the decay heat fraction is calculated:

^=0.01802 Po This decay heat fraction corresponds to approximately 3450 seconds after shutdown for Appendix K decay heat. Therefore, gas flow in the hot legs should drop below the entrainment threshold at about 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> based upon Appendix K decay heat.

c. A description of how the boric acid concentration of the sump fluid is determined.

The boric acid concentration of the sump fluid is determined as follows:

The sump fluid is a general term used to refer the make-up coolant in the SKBOR calculation. During the injection mode of emergency core cooling system (ECCS) operation, the make-up coolant boric acid concentration is modeled as the refueling water tank (RWT) technical specification maximum concentration (4400 ppm). During the sump recirculation mode of ECCS operation that begins at 6100 seconds, the make-up coolant boric acid concentration is modeled as the mass weighted average boric acid concentration of the reactor coolant system (RCS) initial concentration (2100 ppm), the safety injection tank (SIT) concentration (4400 ppm), and the RWT concentration (4400 ppm) minus the mass of boric acid that accumulates in the reactor vessel during the injection phase. Also subtracted from the sump mass is an amount of pure water needed to saturate the containment atmosphere which effectively raises the boric acid Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 24 of 34 concentration in the sump by a small amount. Figure 5-2 provides the make-up coolant boric acid concentration for the limiting Standard fuel analysis.

Make-Up Coolant Concentration

+---- -+-

+----

Time after LOCA (hr)

NSM>U)T Scsakw CX^. 824753700 Figure 5-2: Standard Fuel Boric Make-Up Coolant Acid Concentration Methods, Technology. & Licensing Memo Template Version I-O

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 25 of 34

References:

1. J.P. Cunningham and H.C. Yeh, Experiments and Void Correlation for PWR Small-Break LOCA Conditions, Trans. ANS 17, p. 369-370 (1973).

2. L.E. Hochreiter and H.C. Yeh, Mass Effluence During FLECHT Forced Reflood Experiments, Nuclear Engineering and Design, 60, p. 413-429 (1980).

3. H.C. Yeh, Modification of Void Fraction Calculation, Proceedings of the Fourth International Topical Meeting on Nuclear Thermal-Hydraulics, Operations and Safety, Volume 1, Taipei, Taiwan, June 6, 1988.

4. One-dimensional Two-phase Flow, G.B. Wallis, McGraw-Hill Book Company, 1969.

5. Ishii, M., Grolmes, M. A., Inception Criteria for Droplet Entrainment in Two-Phase Concurrent Film Flow, AlChE Journal, Vol. 21, No. 2, pp. 308-318, March 1975.

6. ML051940575, Beaver Valley Power Station, Unit Nos. 1 and 2, Response to a Request for Additional Information in Support of License Amendment Request Nos. 302 and 173, July 2005.

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 26 of 34 RAI 6. Thermal Conductivity Degradation and Radial Fall-Off Curve Penalty Palo Verde has proposed imposing a radial fall-off curve to offset the lack of explicit consideration of thermal conductivity degradation (TCD) in the fuel performance models in the FATES3B and STRlKfN-II codes.

a. Please provide technical justification that the proposed allowance for TCD is adequate for the full set of analyzed events within Palo Verdes licensing basis (e.g., by comparing the results calculated by FATES3B and STRIKfN-Il against those of a fuel performance code that explicitly models TCD and has been reviewed by the NRC staff).

Response

a. A benchmark study between FATES3B and PADS was performed to ensure that the Thermal Conductivity Degradation (TCD) allowance submitted with the Palo Verde Next Generation Fuel (NGF) Licensing Amendment Request (LAR) (Reference 1) remained valid in light of the most recent industry data. The benchmark study is a code-to-code comparison, which shows the temperature differences between the FATES3B and PADS code. The PADS code has considered a wide array of thermal data from the Halden reactor, including data at a wide range of powers and burnups. The benchmark study indicates the differences in the thermal model behavior between the two codes including any models that contribute to fuel temperature predictions. The differences in fuel temperatures and rod internal pressures were reviewed with respect to how they are used in the relevant Palo Verde NGF LAR analyses. This comparison showed that in all cases the conclusions in the NGF LAR are unchanged. The TCD allowance included in the NGF LAR remains valid and no additional penalty is required to account for TCD.

Attachment 3 contains supplemental information in support of the above response.

References:

1. Maria L. Lacal (APS) to U.S. Regulatory Commission, Attachment 7, Description and Assessment of Proposed License Amendment, July 1, 2016, Accession No. ML16188A333.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 27 of 34 RAI 7. Appendix K LOCA LOOP Considerations Palo Verde has assumed that the limiting results for evaluating the large-break LOCA event would occur when offsite power is unavailable. The availability of offsite power would result in earlier ECCS pump start times than considered in the LAR submittal. While it is possible that earlier ECCS pump start times may tend to refill the downcomer more rapidly (thereby promoting an earlier reflooding of the reactor core), it is also possible that the downcomer may continue to be refilled largely by the safety injection tanks, even if earlier ECCS pump start times are implemented. In the latter case, the earlier spilling of ECCS coolant into containment would tend to produce a more severe containment pressure reduction, and hence offer increased resistance to reflooding the core. The net impact of these countervailing tendencies on the results of the large-break LOCA analysis is not obvious; in particular, the NRC staff notes the counterintuitive observation that, according to the current, conservative Appendix K evaluation methodology, large-break LOCA scenarios with full availability of the ECCS are calculated to be more limiting than cases with a single failure that would reduce ECCS flow. Therefore, please provide the results of an additional analysis of the large-break LOCA event with offsite power available and realistic pump start times to confirm whether the results are bounded by the analysis presented in the LAR submittal.

Response

This request for additional information (RAI) asks Arizona Public Service (APS) to provide the results of an additional analysis of the large break loss-of-coolant accident (LBLOCA) event with offsite power available (OPA) and realistic pump start times to confirm whether the results are bounded by the analysis presented in the license amendment request (LAR), Reference 1. APS asserts that no additional analysis of the LBLOCA event is necessary because the analyses presented in the LAR submittal fully conform to NRC-approved methodologies associated with the 10 CFR Part 50 Appendix K evaluation model used, and because treatment of offsite power was properly addressed in the application of the evaluation model to PVNGS.

As required per the NRC approved next generation fuel (NGF) topical report WCAP-16500-P-A (Reference 5),

the LBLOCA emergency core cooling system (ECCS) performance analysis documented in the analysis of record (AOR) is conducted within a computational framework described in the LBLOCA evaluation methodology (EM) topical report (TR), Reference 2. This computational framework consists of documented case studies, as well as NRC imposed constraints and limitations. This computational framework includes modeling a LBLOCA with some conditions consistent with a loss of offsite power (LOOP) and others with OPA to assure overall conservatism in the analysis results.

As background. Combustion Engineering (C-E) and Westinghouse informed the Atomic Energy Commission (AEC) in 1973 that LOCA analyses were prepared assuming a loss of offsite power, but the Commissioners required nonetheless that analyses assume the availability of offsite power when necessary to ensure conservatism, such as in calculating containment back pressure due to the operation of pressure-reducing systems [CLI-73-39, 6 AEC 1085, Reference 3, page 1122]. Subsequently, the NRC staff noted in Section 15.3.8 of the Safety Evaluation Report (SER), NUREG-0852, Reference 4) for the Combustion Engineering Standard Safety Analysis Report (CESSAR), that ...During the LOCA calculation, offsite power is assumed to be lost... However, this pertained only to establishing the timing of safety injection flow delivery for use in evaluating primary system flow parameters and system behavior during the accident. As noted in Section Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 28 of 34 6.3.3.2.4 of the CESSAR, however, offsite power was assumed to be available for the purpose of minimizing containment pressure-reducing equipment startup times (for example, containment spray). Thus the evaluation model used a hybrid approach in which offsite power was assumed to be both available and unavailable at the same time, to assure conservatism throughout the analysis results.

The NRC-approved evaluation model used for the PVNGS application is derived from the earlier CESSAR evaluation model, and thus retains many of the same elements of methodology that have been previously accepted by the NRC. For example, with respect to the calculation of containment back pressure in the COMPERC-ll/LB computer code, the analysis presented in the EAR utilized input data for the containment (e.g., containment initial conditions, containment volume, containment passive heat sinks, and operation of containment heat removal systems) that were specifically selected to minimize the transient containment pressure. An input value of [ gpm was used to model the maximum safety injection flow per injection point (and thus maximize spillage to containment) in the PRESS module of COMPERC-ll/LB. This value conservatively bounds the actual PVNGS plant configuration, which would yield about ( % less flow and thus less spillage.

The approach to conservatism initially adopted in the CESSAR evaluation model continued into the development of the LBLOCA EM, Reference 2 applied in the AOR and reported in the LAR (Reference 1) for the implementation of NGF at PVNGS. For example:

Section 1 V.D. I .b( 1) of Volume 11 of the TR (Reference 2b) states that LOOP is assumed upon pipe rupture.

Section 4.2.1 of Supplement 3-P-A (Reference 2e) describes studies demonstrating that the most damaging single failure of ECCS equipment following a LBLOCA is the failure of one low pressure safety injection pump. The NRC stated a concern that assuming no single failure of ECCS equipment may result in higher calculated peak cladding temperatures than assuming the most damaging single failure. This is because the increased spillage of injection fluid would produce a lower containment back pressure. A sensitivity study was performed to determine which assumption (single failure or no single failure) is more limiting for the analysis of the Westinghouse nuclear steam supply system (NSSS) with the 1985 EM (Reference 2e). The results show that the assumption of no single failure of ECCS equipment is the more limiting assumption. The TR for the 1985 EM (Reference 2e) concluded that when analyzing Westinghouse NSSS with the 1985 EM, no single failure of ECCS equipment will be assumed. This result was validated again with the 1999 EM (Reference 2f)-

Section 4.2.2 of Supplement 3-P-A of the TR (Reference 2e) describes sensitivity studies which show that tripping the RCPs at the start of the LBLOCA. consistent with the assumption of LOOP, yields higher peak cladding temperature (PCT).

Table 3.3-1 of Supplement 4-P-A of the TR (Reference 2f) describes general guidelines for LBLOCA design input parameters for the 1999 EM (Reference 2f). These guidelines include modeling the (a)

^(b)[ ' and (c) [

r.

Section 3.4.3 of Supplement 4-P-A (Reference 2f) presents the results of three worst single failure analyses using the 1999 EM. These cases are (a) no failure of an ECCS component, (b) the loss of a Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 29 of 34 LPSl pump, and (c) the loss of a diesel generator. The analyses concluded that no failure of an ECCS component produces the highest PCT for the 1999 EM. Other plant configuration combinations of containment size and ECCS delivery rates, may lead to a different conclusion, therefore, the TR stated that the worst single failure analysis will be performed for each application of the 1999 EM. The worst single failure of an ECCS component includes consideration of the most limiting value of the refueling water storage tank (RWST) temperature as described in Section 3.3.1 of (Reference 2f).

Section 3.3.3 of Supplement 4-P-A (Reference 2f) describes a study of three different times for safety injection pump (SIP) actuation time using the 1999 EM. The three cases analyzed are (1) safety Injection actuated during early reflood (based on safety injection actuation signal [SIAS] and delay time), (2) safety injection actuated at the end of blowdown; i.e., at the time of annulus downflow (TAD),

and (3) safety injection actuated after the safety injection tanks (SlTs) empty. Table 3.3-4 of (Reference 2f) provides the results of the three cases analyzed in this parameter study. Earlier actuation of safety injection pump delivery. Cases (1) and (2), is [

pc

Section 4.2.2 of Supplement 3-P-A of the TR (Reference 2e) describes a sensitivity study performed to determine the effect of reactor coolant pump (RCP) operation during blowdown on the results of a LBLOCA analysis. The cases analyzed were the (a) RCPs tripped co-incident with the start of the LBLOCA and (b) RCPs operating throughout the blowdown period of the LBLOCA. The results of the study show that tripping the RCPs at the start of the LBLOCA yields the higher PCT. As a result, the application of the 1999 EM to LBLOCA analyses assumes the RCPs will be tripped at the start of the LOCA.

In addition to the aforementioned conservatisms that are part of the approved methodology, the application of the 1999 EM is also associated with discretionary conservatisms applied by the LBLOCA analyst. Section A.l of Revision 4 to Appendix A of Supplement 4-P-A (Reference 2h) describes user-controlled input parameters for introducing discretionary conservatism into the 1999 EM analysis by (a) increasing the core two-phase mixture level used in the steam cooling calculation, and (b) reducing the reflood rate thereby slowing the core reflooding and increasing cladding temperature. The application of discretionary conservatism in the AOR represents an increase of [ °F on PCT, [ % on peak local oxidation (PLO), and [ p*" % on core-wide oxidation (CWO).

Section 3.3 of CENPD-132-P-A, Supplement 4, Revision 1, Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model, acknowledged that ECCS performance sensitivities and plant parameter values may vary from one plant design to another. The topical report provided the results of a parametric study with respect to SI pump actuation time for a reference CE PWR plant design, and concluded that [

I** The parametric study also concluded that selection of SI pump actuation time was of low impact to ECCS performance analyses for CE plants, affecting the calculated PCT by [ p' °F. The Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 30 of 34 associated NRC Safety Evaluation (SE) dated December 15, 2000, concluded that this method for selecting the SI pump actuation time was acceptable because it also best represented actual SI actuation design logic.

The LBLOCA ECCS performance analysis for PVNGS utilized the guidelines of CENPD-132-P-A with respect to selecting the SI pump actuation time. Further evaluation by Westinghouse in response to the NRC RAl concludes that this selection method remains applicable to the PVNGS ECCS design (that is, the methodology applies to more than the reference CE PWR plant design in the topical report). Specifically, a PVNGS specific sensitivity evaluation was performed to assess the effects of having offsite AC power available, which would permit both an earlier SI pump actuation than the CENPD-132-P-A methodology, as well as short-term operation of the Reactor Coolant Pump (RCPs) during blowdown which is not credited in licensing basis loss of offsite AC power cases. A case was run without discretionary conservatism by reducing the SI pump actuation time to 0 seconds and securing the RCPs at [ p' seconds, which is during blowdown prior to the time of annulus downflow. This case resulted in a reduction in PCX by approximately [ °F. A case was also run without discretionary conservatism by reducing only the SI pump actuation time to 0 seconds (time of break). This case resulted in a PCX increase of approximately [ °F, which is consistent with the level of sensitivity demonstrated in the CENPD-132-P-A parametric study ([ p' °F) with respect to SI pump actuation time. These results demonstrate that, with offsite power available, the beneficial effect of short-term RCP operation during blowdown would overwhelm any adverse effect of increased spillage to containment due to an earlier SI pump actuation time. Therefore the PVNGS sensitivity evaluation supports the conclusion that the generic CENPD-132-P-A sensitivity evaluation is applicable to the PVNGS ECCS design. Thus, the approved CENPD-132-P-A methodology framework, with a loss of offsite AC power and a maximum SI pump delay, provides conservative results that bound LBLOCA scenarios in which offsite AC power remains available.

Regardless, the limiting PVNGS ECCS performance analysis for NGF, reported in Table 8-2 of the LAR (ML16188A333), yielded a PCX of 2129.6 °F that includes approximately [ ]*' °F of discretionary conservatism. This discretionary conservatism is significantly larger than the level of sensitivity demonstrated in the sensitivity studies with respect to SI pump actuation time, and provides further assurance that the PVNGS ECCS performance analysis documented in the LAR remains the analysis of record for NGF.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment!

June 1,2017 Page 31 of 34

References:

1. APS Letter, 102-07277-MLL/GWA, Palo Verde Nuclear Generating Station (PVNGS) Units 1, 2, and 3, Docket Nos. STN 50-528,59-529, and 50-530, License Amendment Request and Exemption Request to Support the Implementation of Next Generation Fuel, July 1, 2016 (ADAMS Accession No. ML16188A332); Attachment 8, WCAP-18076-P, Revision 1, Reload Transition Safety Report for Palo Verde Nuclear Generating Station Units 1, 2 and 3 with Combustion Engineering 16x16 Next Generation Fuel, June 30,2016.

2. CENPD-132

a. CENPD-132 P, Volume I, Calculative Methods for the C-E Large Break LOCA Evaluation Model, August 1974.

b. CENPD-132 P, Volume II, Calculative Methods for the C-E Large Break LOCA Evaluation Model, August 1974.

c. CENPD-132P, Supplement 1, Calculational Methods for the C-E Large Break LOCA Evaluation Model, February 1975.

d. CENPD-132-P, Supplement 2-P, Calculational Methods for the C-E Large Break LOCA Evaluation Model, July 1975.

e. CENPD-132, Supplement 3-P-A, Calculative Methods for the C-E Large Break LOCA Evaluation Model for the Analysis of C-E and W Designed NSSS, June 1985.

f. CENPD-132, Supplement 4-P-A, Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model, March 2001.

g. CENPD-132-P-A Supplement 4-P-A Addendum 1-P-A, Calculative Methods for the CE Nuclear Power Large Break LOCA Evaluation Model, Improvement to 1999 Large Break LOCA EM Steam Cooling Model for Less Than 1 in/sec Core Reflood, August 2007.

h. CENPD-132-SUPP 4-P-A, APP A-REV004, Revision 4 to the Supplement to Appendix A of CENPD-132 Supplement 4-P-A, December 2008.

3. CLI-73-39, 6 AEC 1085, December 28, 1973.

4. NUREG-0852, Revision 000, Safety Evaluation Report related to the final design of the Standard Nuclear Steam Supply Reference System CESSAR System 80, Docket No. STN 50-470, November 1981.

5. WCAP-16500-P-A, Revision 0, CE 16x16 Next Generation Fuel Core Reference Report, August 2007; Supplement 1-P, Application of CE Setpoint Methodology for CE 16x16 Next Generation Fuel (NGF), March 2007.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1, 2017 Page 32 of 34 RAI9. Containment Analyses

a. In Section 9.1 of Attachment 8, Mass and Energy Release Analysis for Postulated LOCAs, it is stated that an evaluation of the impact of NGF on the LOC A Mass and Energy (M&E) AORs was performed.

Additionally, a comparison of fuel parameters and operating conditions was performed. Please describe how the containment LOCA M&E release was determined for the NGF analyses. To confirm that the AOR LOCA M&E short term (UFSAR Table 6.2.1-4 and 6.2.1-5) and long term (from the end of post reflood) releases remain bounding, please provide quantitative results comparing to the AOR M&E releases for the following:

i. Short-term M&E release during blowdown. Also, please confirm that the AOR containment pressure response for peak pressure determination and containment temperature response for equipment environmental qualification remain bounding.

ii. Long-term M&E release for the sump temperature response. Also, please confirm that the AOR sump temperature profile for the ECCS pumps net positive suction head analysis remain bounding.

b. In Section 9.2 of Attachment 8, M&E Release Analysis for Postulated Secondary System Pipe Ruptures Inside Containment, it is stated that the AOR FW temperature bounds the NGF temperature. What is the FW temperature used for AOR and the NGF analysis? Please explain how the AOR temperature produces bounding results.

c. In Section 9.2 of Attachment 8, it is stated that the M&E source energy based on NGF operating conditions will remain bounded by the AOR main steam line break (MSLB) source energy. Please justify quantitatively that the parameters that determine the AOR MSLB containment M&E source energy bound those that determine the M&E source energy with NGF.

Response

a. The short-term and long-term mass and energy releases were not explicitly performed for the NGF fuel transition. Instead, an evaluation (Reference 1) was performed that compared the core pressure drops and primary fuel parameters that impact the M&E releases. This approach was discussed with the NRC on a conference call April 12, 2017, where it was requested that the details of the evaluation be presented in lieu of the mass and energy comparison.

Long-Term Mass and Energy Releases The NGF pressure losses were scaled to the AOR conditions. Flow paths which used the core stations were evaluated and compared. The evaluation showed an increase in pressure loss through active core flow paths.

Conversely, the evaluation also showed that the flow paths upstream and downstream to the active core saw a decrease in pressure losses. The overall pressure loss changes through the reactor vessel from the inlet nozzles to the outlet nozzles were insignificant based on Reference 2, which states that changes in geometry losses of +/-10% have shown no significant change in the transient results. Therefore, these differences in pressure losses are negligible for the blowdown M&E release calculation and will have no significant effect on the transient results.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 33 of 34 The fuel parameters evaluated were:

Core average linear heat rate Scaling to a core power of 4070 MWt, the maximum value for core average linear heat, results in core average linear heat rate of 5.735 kW/ft. The AOR blowdown analysis calculated a core average linear heat rate value of 5.925 kW/fl. The AOR blowdown value is larger and results in more energy being transferred to the coolant, which is conservative for containment mass and energy releases.

Pellet and cladding geometry Three primary dimensions changed due to NGF, which were pellet outside radius, cladding inside radius and cladding outside radius. All three of these dimensions were reduced for NGF, which will result in less surface area for heat transfer from fuel to coolant. Higher heat transfer from fuel to coolant is conservative for containment M&E release analyses, therefore, the AOR geometry is bounding and conservative. Its also noted that these geometry changes resulted in increasing the flow area through the core, however, the effect of flow areas is already included in the core pressure drops discussed previously.

Centerline Temperature The power at each axial node is slightly higher for NGF than the AOR; however, the centerline temperature is what drives the heat transfer and, therefore, the slight power difference is considered to be negligible for LBLOCA M&E release analysis. The centerline temperatures used in the AOR are based on the Erbia fuel composition plus a 50°F addition for conservatism. With this conservatism, the AOR centerline temperatures bound all the NGF centerline temperatures.

Decay Heat The AOR assumed a 102% thermal rating modifier to set the initial core power level which is bounding for NGF decay heat.

Metal/Water Reaction The Zirc-water reaction option is enabled for the LBLOCA M&E release analysis. However, the M&E release analysis biases the reactor core parameters to extract as much energy from the fuel and components as possible in order to generate steam. This results in lower fuel clad temperatures such that the metal-water reaction is negligible.

Short-Term Mass and Energy Releases The short-term mass and energy releases generated for the tributary line break transients is a short duration event (approximately 1 second). There is not sufficient time for the reactor core, and the primary and secondary sides to interact to significantly affect the mass and energy releases. The initial conditions that Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 2 June 1,2017 Page 34 of 34 have an impact on the analysis such as maximum reactor coolant system pressure and temperature have not changed as a result of the NGF transition.

b. The full-power feedwater temperature used for the AOR is 450°F, and the full-power feedwater temperature for NGF is 448°F. A higher feedwater temperature results in more energy being transferred to the steam generator secondary side until the feedwater system is isolated. This results in higher energy steam being released during the blowdown of the steam generator and therefore a higher pressure and temperature containment response.

c. The main steam line break M&E evaluation (Reference 3) calculated the total energy in the fuel region in hand calculations based on average temperature, UO2 density, specific heat and volume (Table 1).

The AOR total fuel region energy is greater than the NGF total fuel region energy (34.67 MBtu vs.

34.53 MBtu). A higher fuel region total energy results in more energy being transferred to the reactor coolant system which will then transfer more energy to the steam generator secondary side. This results in higher energy steam being released during the blowdown of the steam generator and therefore a higher pressure and temperature containment response.

Table 1: Total Energy in the Fuel Region Comparison AOR Parameter NGF (Fuel Region 1/2)

Average Temp (°F) 1967/1217 1628.3 U02 Density (Ib/ft^) 684.21 684.21 Specific Heat (Btu/lb-°F) 0,0783/0.0750 0.076854 Volume (ft) 209.61/201.23 403.3 Total Fuel Region Energy (MBtu) 34.67 34.53

References:

1. LTR-SCC-15-002, Rev. 1, Palo Verde NGF/MUR LBLOCA M&E AOR Evaluation.

2. CENPD-132 Volume 1, Rev. 01, Calculative Methods for the C-E Large Break LOCA Evaluation Model.

3. LTR-SCC-15-003, Rev. 0, Palo Verde NGF /MUR MSLB M&E AOR Evaluation.

Methods, Technology, & Licensing Memo Template Version I-O

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Attachment 3 Confirmation of FATES3B Thermal Conductivity Degradation Allowance for Arizona Public Service Next Generation Fuel License Amendment Request (Proprietary)

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Attachment 4 SKBOR: A Computer Code for Calculating the Accumulation of Boric Acid in the Reactor Vessel (Proprietary)

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Attachment 5 SKBOR: A Computer Code for Calculating the Accumulation of Boric Acid in the Reactor Vessel (Non-Proprietary)

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1 June 1,2017 Attachment 5 SKBOR: A Computer Code for Calculating the Accumulation of Boric Acid in the Reactor Vessel (Non-Proprietary)

Westinghouse Electric Company 1000 Westinghouse Drive Cranberry Township, PA 16066

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 2 of 27 TABLE OF CONTENTS 1.0 Introduction..................................................................................................................................3 2.0 Limits of Applicability..................................................................................................................3 3.0 Problem Formulation, Initial Conditions, andGoverning Equations......................................... 4 3.1 Initial Conditions................................................................................................................. 4 3.2 Governing Equations............................................................................................................5 3.3 Void Fraction Model.............................................................................................................6 3.4 Core-to-Lower Plenum Boric AcidTransport Model............................................................ 10 3.5 Liquid Carryover Model..................................................................................................... 12 4.0 Input Description........................................................................................................................16 4.1 Timing and Problem Control................................................................................................16 4.2 Component Boron Concentrations...................................................................................... 18 4.3 Component Fluid Masses....................................................................................................18 4.4 Effective Vessel Mixing Volume.........................................................................................19 4.5 Core Power and Decay Heat Model Options....................................................................... 21 4.6 Lower Plenum Subcooling Model.......................................................................................22 4.7 Upper Plenum Condensation Model...................................................................................22 4.8 Core Pressure..................................................................................................................... 22 4.9 Core-to-Lower Plenum Boric AcidTransport Model........................................................... 23 4.10 Liquid Carryover Model..................................................................................................... 23 5.0 Output Description.................................................................................................................... 24 5.1 Graphics Output Binary File............................................................................................... 24 6.0 Summary of SKBOR Releases................................................................................................... 26 7.0 References...................................................................................................................................27 Methods, technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 3 of 27 1.0 Introduction This report describes the SKBOR computer code, which is used to determine: (1) the time at which emergency core cooling system (ECCS) recirculation should be realigned to the reactor coolant system (RCS) hot legs (or cold legs for upper plenum injection (UPI) plants) to prevent boron precipitation in the long term post-LOCA; (2) the interval at which cycling between hot and cold leg injection should be completed, for plants without sufficient simultaneous hot and cold leg injection; and, (3) the amount of sump dilution at hot leg switchover time.

Section 2.0 identifies the limits of applicability regarding the calculations performed by the code.

Section 3.0 describes the SKBOR problem formulation and initial conditions and presents the governing equations that are solved by the code. Section 4.0 describes the input data used by the program. Section 5.0 describes the contents of the graphics output binary file that is generated by SKBOR. Finally, Section 6.0 summarizes the releases of SKBOR on UNIX and Linux platforms.

2.0 Limits of Applicability The void correlation used in SKBOR has only been validated for pool boiling scenarios in rod bundle geometries such as for Westinghouse 3-/4-loop nuclear steam supply system (NSSS) and Combustion Engineering NSSS during the cold leg recirculation phase. Presently, different methods are used to calculate the void fraction for Westinghouse 2-loop NSSS with upper plenum injection.

The transport model is limited to [

The liquid carryover model option is intended for advanced plant or non-standard applications.

Linux is the registered trademark of Linus Torvalds in the U.S. and/or other countries.

UNIX is a registered trademark of The Open Group.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 4 of 27 3.0 Problem Formulation, Initial Conditions, and Governing Equations A typical SK.BOR calculation considers two volumes: one representing the effective vessel mixing volume (referred to as the CORE) and one representing the remaining system inventory (referred to as the SUMP). In some cases, a third volume representing steam/water mixing in the reactor vessel upper plenum (referred to as the UP) is also considered. All mass storage is assumed to occur in either the CORE or SUMP, with any mass entering the UP in a given timestep assumed to return to either the CORE or SUMP at the end of the timestep. Figures 3-1 and 3-2 illustrate the mass and boron calculations in SKBOR.

I 3.1 Initial Conditions The initial conditions assumed in SKBOR can be summarized as follows:

The calculation is initiated at the user-specified start time TSTART (s) and a system pressure of PCORE (psid) that is assumed to remain constant throughout the calculation.

The CORE is initially assumed to be full of borated liquid, with a concentration (WTFcore'^ weight fraction) equal to the initial system-average boron concentration or a user-specified boron concentration with a density (Pcore^ Ibm/ft^) given by the following equation:

PcoRE ~ X (1 -F 0.1629 X WTFcore^

where Pf {Ibm/ft^) represents the saturated liquid density of water at a system pressure of PCORE (psia).

Denoting the effective vessel mixing volume as Vqore the initial CORE mass, Mt.core ilbm), is:

^T,CORE ~ ^CORE ^ PcORE and the CORE boron. Mg,core Qbm), and water, M^xore Obm), masses are:

^B.CORE ^ ^T.CORE ^ ^TFcqRE

^W.CORE ~ ^T.CORE ^ ~ ^TFcore)

Methods. Technology. & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1, 2017 Page 5 of 27 The SUMP is initially assumed to contain borated liquid with a concentration (WTFsump^ weight fraction) equal to the initial system-average boron concentration. Denoting the total system mass as Mtxot Qbm), the initial SUMP mass {Mjsump Q-bm) is:

^T,SUMP ^ ^T.TOT ~ ^T.CORE Then, the SUMP boron, Mq^ump Obm), and water, ^ump Obm), masses are:

^BSUMP ~ ^T,SUMP ^ ^TFsumP

^W,SUMP ~ ^T,SUMP X (1 ~ WTFsump) 3.2 Governing Equations The mass and boron calculations in standard SKBOR application are illustrated in Figures 3-1 and 3-2, respectively, and can be summarized as follows for a given timestep. At (s) (note that SI and CONDENSATION terms only apply when WSI # 0, see Section 4.7):

The core fluid density, Pcorei is calculated at the beginning of the timestep.

The decay heat mass boil-off over the timestep At is computed as:

_Qcore X (P/Pq) X At

^T.BOIL

^fg (V where: ^T.BOiL ~ decay heat mass boil-off over timestep ilbm)

Qcore ~ initial core power level (fiTf//s)

P/Pq = normalized core power fraction at beginning of timestep Y5 = enthalpy of formation (BTU/lbrn)

{hf hip) = lower plenum subcooling {BTU/lbm)

The CORE and SUMP boron, water, and total masses at the end of the timestep are computed from the values at the beginning of the timestep using the following information:

The mass exiting the CORE over the timestep At (i.e., Mj gQn) is assumed to consist of unborated vapor at enthalpy hg.

Boric acid is added from the sump at density Pf and concentration WTFsump, as required to keep VcoRE full.

The vapor exiting the core is assumed to condense fully in containment and return to the sump as unborated liquid.

The total system mass remains constant, with all mass storage assumed to occur in either the CORE or SUMP.

Denoting boron mass as Mg and total (i.e., boron + water) mass as Mt, the CORE and SUMP boron concentrations are computed using the following general equations:

Cg (weight fraction) = Mg/Mj

The hot leg switchover time is defined as the time at which the core boron concentration reaches the assumed boric acid solubility limit (weight percent, w/o).

Methods. Technology. & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1, 2017 Page 6 of 27 Void Fraction Model The void fraction model used in SKBOR is the Yeh correlation (References 1 through 3). The model is summarized in this section.

Nomenclature JgJfJ Superficial velocity, of the gas, liquid, and mixture Qcore Core power, (^)

Heat flux, (^)

Ph Fuel rod heated perimeter, (/t)

% Decay heat power fraction hf,hg Saturation enthalpies for the liquid and gas phases, Heat of vaporization, ~

Mass nux, (^)

G m Mass flow rate,

^core Core flow area, (/t^)

A fuel Surface area of the fuel, (/t^)

surface Z Core axial elevation, relative to bottom of active fuel, (ft)

^core Active fuel length, (ft) t Time, (s)

Gravitational acceleration, ^32.174 9

Gravitational constant, ^32.174^^^^)

9c Vbcr Critical bubble velocity, b Regime-specific exponent from Table 3.3-1 C Regime-specific coefficient Table 3.3-1 a Void fraction Density, (^)

P a Surface tension, Methods, Technology, & Licensing Memo Template Version I-O

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 7 of 27 Conservation of Energy For the case of pool boiling with no liquid subcooling, steam superheating, or liquid carry-over, the energy balance is:

For uniformly distributed core power, a,c Conservation of Mass Over a sufficiently short time interval, the system can be considered to be in a quasi-steady state such that the net change in the mass inventory is approximately zero. Therefore, for the purpose of calculating the void fraction, conservation of mass for the active (heated) core region control volume gives; (for small At)

Since the flow area through the core (>lcore) does not vary axially over the heated length:

(for small At)

The mass flux at a given elevation is:

(for small At) and therefore; where:

Methods, Technology, & Licensing Memo Template Version I-O

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 8 of 27 The following expressions are obtained for yy(z) and jg{z):

-,a.c The time-dependent heat flux is calculated as follows:

where:

CcoreCO ~ QaareO- = 0) X (t) and decay heat power fraction at time, t, following the shutdown of the reactor core.

Then, the core average void fraction is calculated as follows.

acore avg j^^ore and acore = a(Zcore) exit Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 9 of 27 Yeh Correlation Equations The full form of the Yeh void fraction correlation (References 1 through 3) is:

a(z) = C <Pf) \ ^bcr )

where:

Hpf - Pg)99c V.rr = 1.53 Pf The variables b and C vary by regime and are defined as shown in Table 3.3-1. The superficial velocities, jf and jg, are calculated as derived earlier. In regions of the inner vessel above the active ftiel that are included in the mixing volume, changes in the flow area of the region relative to the active fuel region will change the superficial gas velocity (jg) which, in turn, changes the void fraction in the region relative to the core exit void fraction, acore. The core exit void fraction is adjusted according to:

exit The subscript, UP, is used to represent any region above the active core. If multiple regions above the active core with differing flow areas are modeled, then the adjustment must be done for each region.

The exponent b is determined using Table 3.3-1 based on superficial gas velocity at the core exit, jg.core. Currently, SKBOR does not adjust the upper plenum void fraction. This must be done by the exit user.

Table 3.3-1: Constants and Exponents for the Yeh Void Fraction Model Regime c b Small bubble ^<1.0 0.925 0.67

'hrr Large bubble 1.0 <^< 4.31 0.925 0.47

hrr 1.035 0.393 Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 10 of 27 3.4 Core-to-Lower Plenum Boric Acid Transport Model A two-region boric acid transport model is available in SKBOR. This model predicts the boric acid concentrations in the core and lower plenum by assuming liquid-density-gradient-gravity-driven exchange flow through the lower core plate. Each region is assumed to be well mixed and therefore, the boric acid concentration and temperature in each region is assumed to be uniformly distributed throughout. The core is assumed to be at saturation temperature while the lower plenum can be either saturated or subcooled, depending on the user-specified initial condition. In this scenario, the flow required to make-up for boil-off, Qbon (i e-, an externally supplied flow) and bidirectional flow is present in all of the lower core plate holes such that the total exchange flow rate, Q, is determined using Eq. 3-3. The problem is depicted schematically in Figure 3-3.

Inception of boric acid transport is determined by two factors. First, the density gradient between the core and lower plenum due to solute concentration differences must overcome the density gradient caused by the temperature difference between the core and lower plenum if subcooling exists in the lower plenum region. Second, since there is upflow through the reactor vessel due to the makeup of liquid boil-off, the buoyancy-driven exchange flow in the downward direction must be larger than the boil-off flow rate in the upward direction such that the downward flow can penetrate through the lower core plate and into the lower plenum. By modeling the inception in this fashion, both the effects of subcooling in the lower plenum and upward liquid kinetic energy due to the makeup of boil-off are accounted for.

The volumetric flow of make-up water through the lower plenum and into the core is Qbou, which is equal to the boil-off rate, and the source concentration (weight fraction) of boric acid into the lower plenum from the sump is denoted by the symbol Mi. The quantity of interest is the concentration of boric acid in the core, M2, as a function of time, t, within the core region.

The boil-off flow, Qboih through the vessel is essentially an externally supplied flow that passes through the lower core plate and carries boric acid into the core region. Initially, the boric acid concentration in the core increases with time at a rate directly proportional to QboU- However, when the boric acid in the core becomes sufficiently concentrated, the density of the core solution exceeds that of the solution in the lower plenum. This density difference induces a buoyancy-driven, countercurrent downflow of the heavier core liquid and consequential upflow of the lighter liquid through the openings in the lower core plate.

Model Equations The following linear expression for the density difference, Ap2i between the liquid in the lower plenum (region 2) and the reactor core (region 1) can be written as:

(Eq. 3-1) where T and M refer to the temperature and boric acid weight fraction, respectively, p is the volumetric coefficient of thermal expansion of water, k is the boric acid expansion coefficient.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 11 of 27 p is the effective constant density of the liquid solution in the vessel and is defined as the average between the lower plenum and core densities:

(a+a)

^ 2 (Eq. 3-2)

The term Q^oii represents the net upward flow through the lower core plate required for make-up due to boil-off of liquid in the core region. It is the difference between the actual upward flow through the plate, Qi2 and the buoyancy-driven downward flow from the core to the lower plenum, Qgp. A volumetric flow balance across the lower core plate requires that:

Qboil ~ Qu QbF (Eq. 3-3)

The countercurrent flow occurs within each opening (hole) in the lower core plate. Denoting N as the number of holes in the plate and assuming that each hole in the plate has the same diameter, Qbou/N may be regarded as an externally imposed, upward forced flow opposite to the downward buoyant flow, Qbf/^ in each opening.

The time histories of the solute concentrations. Mi, and the temperatures, Tj, in each region are given transient solute mass and liquid energy balances. Applying the correlations for Qbf along with Eq. 3-3, the transient mass and liquid energy balances can be simplified to form a set of nonlinear equations that are functions of Tj, Ml, Qgf, and Q^ou which can be solved numerically for a set of given initial and boundary conditions.

Inception Criteria In view of Eq. 3-1, when Api2 ^ 0 the system is stably stratified and the buoyancy-driven back flow, Qbf is zero, that is:

Qbf=0 when Api2 ^ 0 (Eq. 3-4)

Also, when the destabilizing density difference, Api2 is positive but small, the buoyancy-driven back flow is not large enough to penetrate the upward makeup flow, Qbou through the core plate and the net downward transport rate is again zero:

Qbf ~ ^

when ^12^0-boil (Eq. 3-5)

Eqs. 3-4 and 3-5 define the ineeption criteria for the transport of higher concentration boric acid from the core to lower plenum. If the liquid in the lower plenum is not subcooled, then Api2 ^ 0 and Eq. 3-5 is the only criterion that has to be met, and transport between the core and lower plenum will occur sooner in the transient. If subcooling exists in the lower plenum (Api2 ^ 0)f the concentration gradient between the core and lower plenum will have to overcome the oppositely opposing temperature gradient in addition to Eq. 3-5, and inception will occur later in the transient.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. I, Attachment 5 June 1,2017 Page 12 of 27 3.5 Liquid Carryover Model This model option allows the user to specify a liquid carryover fraction by means of a two-phase mixture mass quality. The liquid carryover from the core is returned to the sump and the liquid mass lost from the core for a given timestep is replenished by the source injection.

The two-phase mixture mass quality exiting the core region is expressed as follows:

x=

M+M, where M is the mass of the steam (g) and liquid (1) constituents. The steam mass is generated due to boiling in the core region and the liquid mass is the carryover from the core. Rearranging the above equation, an expression for the mass carryover, or entrainment, as a function of mass quality is written as:

{\-x)

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1, 2017 Page 13 of 27 UPPER PLENUM (UP)

Upper Plenum Exit (Vapor)

(Liquid)

Boil-ofT Condensation (Vapor) (Liquid)

CORE Vapor Fully Condenses in Containment SUMP Core Makeup (Liquid)

Sump Inlet (Liquid)

Eigure3-1: Mass Calculations in SKBOR Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 14 of 27 UPPER PLENUM (UP)

Upper Plenum Exit Condensation (Ce.core)

CORE Vapor Fully Condenses in Containment SUMP Core Makeup

(^B.sump)

Figure 3-2: Boron Calculations in SKBOR Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietaiy Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 15 of 27 Figure 3-3: Mass Calculations in SKBOR with the Two-Region Transport Model Active Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 16 of 27 4.0 Input Description This section describes the input variables that are available in SKBOR. Default values are provided where available, and recommended input values are suggested where possible. (Note that an entry of

{N/D} in the Default column means that the variable has no default value.)

4.1 Timing and Problem Control The following inputs are used in the various timing and problem control functions in SKBOR:

Name Default Units Description TSTART {100} s Transient start time. Default value is standard for hot leg switchover time calculations.

TEND {100,000} s Transient end time. Default value is generally sufficient for hot leg switchover time calculations.

DT {1} s Timestep size. Default value is recommended for hot leg switchover time calculations.

DTPLOT {100} s Interval at which plot points will be written to the binary graphics output file.

DTPRINT {1,000} s Interval at which timestep printouts will be written to the ASCII output file.

TITLE {N/D} Title for program output.

VOID {0} No units This is a flag to instruct the code as to whether or not the internally coded void fraction calculations are to be used. The values are:

0 No void fraction calculations are to be performed by the code.

Note that voiding can still be accounted for by using the VCORET array.

1 Void fraction calculations are to be performed by the code. This option requires additional inputs LCORE, NCELLS, NFA, NRODA, CDO, LPFRAC, LPVOL7, LPVOL8, COREVOL9, UPVOLIO, and UP VOL 11. This option overrides user-specified inputs for HLP, HSl, VCORE, NVCORET, TVCORET, VCORET, and WSI.

WTPLIMIT {23.53} w/o This variable allows the analyst to specify an alternate boric acid solubility limit. The default value is 23.53 w/o.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 17 of 27 Name Default Units Description TRANCF {0} No units The flag for the user-specified source concentration option. The values are:

0 No user input source concentration used and the make-up coolant source is fed from the sump.

1 The user-specified source concentration option is used which requires the following additional input: TRANCA, TRANCT, and TRANC.

TRANCA {0} No units The array size to be input for the variables TRANCT and TRANC. The maximum array size is 100.

TRANCT {-1.0} The transient time array used for the source concentration input times. The first value corresponds to the value TSTART and the last value corresponds to the switch to sump recirculation or the value TEND.

MIXM {0} No units The flag for the two-region transport model. The values are:

0 The two-region transport model is not used.

1 The two-region transport model is used.

ENTF {0} No units The flag for the liquid carryover option. The values are:

0 No liquid carryover from the core region.

1 Liquid carryover from the core region is considered which requires additional input for the variable ENTX.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 18 of 27 4.2 Component Boron Concentrations In SKBOR, the following inputs are used to define the initial component boron concentrations.

Name Default Units Description CACC {0} ppm Cold leg accumulator initial boron concentration.

CBIT {0} ppm Boron injection tank (BIT) initial boron concentration.

CICE {0} ppm Ice condenser initial boron concentration.

CPIPE {0} ppm ECCS/BIT piping initial boron concentration.

CRCS {0} ppm Reactor coolant system initial boron concentration.

CRWST {0} ppm Refueling water storage tank initial boron concentration.

TRANC {0} ppm The transient concentration array used for the source concentration input values. The number of entries corresponds to the number of TRANCT entries.

4.3 Component Fluid Masses In SKBOR, the following inputs are used to define the initial component fluid masses.

Name Default Units Description MTACC {0} Ibm Cold leg accumulator initial fluid mass.

MTBIT {0} Ibm Boron injection tank initial fluid mass.

MTICE {0} Ibm Ice condenser initial fluid mass.

MTPIPE {0} Ibm ECCS/BIT piping initial fluid mass.

MTRCS {0} Ibm Reactor coolant system initial fluid mass.

MTRWST {0} Ibm Refueling water storage tank initial fluid mass.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 19 of 27 Effective Vessel Mixing Volume The effective vessel mixing volume in SKBOR can be specified several ways: (1) a time dependent volume calculated by SKBOR that accounts for voiding, (2) a user input time dependent volume (VCORET), or (3) a user input constant value (VCORE). The following inputs are used by SKBOR to calculate the effective mixing volume.

Option I - Internally Calculated by SKBOR Name Default Units Description LCORE {0.0} ft This variable is defined as the length of the active fuel region.

NCELLS {0} count This variable is defined as the number of cells to be used in the core voiding calculation.

NFA {0} count This variable is the number of fuel assemblies.

NRODA {0} count This variable is defined as the number of active fuel rods per fuel assembly.

CDO {0.0} in This variable is defined as the fuel rod cladding outside diameter.

LPFRAC {0.0} N/D This variable is defined as the fraction of the lower plenum to be included in the liquid mixing volume.

LPVOL7 {0.0} ft^ This variable is defined as the volume of the reactor vessel lower head region.

LPVOL8 {0.0} ft^ This variable is defined as the volume of the reactor vessel lower plenum (core support) region.

COREVOL9 {0.0} ft^ This variable is defined as the total core region volume.

UP VOL 10 {0.0} ft^ This variable is defined as the volume of the core outlet region.

UPVOLll {0.0} ft^ This variable is defined as the volume of the reactor vessel upper plenum.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 20 of 27 Option 2 - Variable Mixing Volume vs. Time Name Default Units Descrintion NVCORET {0} count Number of points in table of effective vessel mixing volume vs. time (maximum 100).

TVCORET {100*0.0} s Time values in table of effective vessel mixing volume vs. time. Must be in ascending order and bound the problem duration TSTART to TEND.

VCORET {100*0.0} ft' Mixing volume values in table of effective vessel mixing volume vs. time. Must be greater than 0 for I = 1 to NVCORET.

Ootion 3 - Constant Mixing Volume vs. Time Name Default Units Descrintion VCORE {0.0} ft' Effective vessel mixing volume.

Methods. Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 21 of 27 4.5 Core Power and Decay Heat Model Options In SKBOR, the initial core power level is denoted as PZERO, and the core power uncertainty factor is denoted as DPZERO. IDKMOD and AMARG are used to select the decay heat model and multipliers, respectively. Alternate decay heat models can be incorporated using NDK, TDK, and DK.

Name Default Units Description DPZERO {1.02} No units Core power uncertainty factor.

PZERO {0} MWt Core power level without uncertainty.

IDKMOD {1} No units Decay heat model option:

0 User input table 1 1971 ANS infinite without residual fissions 2 1971 ANS finite without residual fissions The default value corresponds to the model prescribed in Section I.A.4 of 10 CFR 50, Appendix K.

AMARG(I) {3*1.2, 1.0} No units Array of decay heat multipliers:

1 Multiplier on fission product decay, 0-1,000 s 2 Multiplier on fission product decay, 1,000-10,000,000 s 3 Multiplier on fission product decay, 10,000,000 s and beyond 4 Multiplier used with U-239 and Np-239 terms The default values correspond to the model prescribed in Section 1.A.4 of 10 CFR 50, Appendix K.

NDK {0} count Number of points in table of normalized core power fraction vs. time (maximum 100), used with the IDKMOD = 0 option.

TDK {100*0.0} s Time values in table of normalized core power fraction vs. time, used with the IDKMOD = 0 option. Must be in ascending order and bound the problem duration TSTART to TEND.

DK {100*0.0} No units Power values in table of normalized core power fraction vs. time, used with the IDKMOD = 0 option. Must be greater than 0 and less than or equal to 1 for 1 = 1 to NDK.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 22 of 27 4.6 Lower Plenum Subcooling Model In SKBOR, credit for lower plenum subcooling can be modeled to reduce the effective core boil-off rate and extend the hot leg switchover time. This model is not active when the internally coded void fraction model is used.

Name Default Units Description HLP {-1.0} Btu/lbm Lower plenum enthalpy (e.g., HLP = 148.016 Btu/lbm for P = 14.7 psia and T = 180 °F). If no credit for lower plenum subcooling is taken, then the default value should be used.

4.7 Upper Plenum Condensation Model In SKBOR, credit for upper plenum condensation can be used to reduce the effective core boil-off rate and extend the hot leg switchover time. This model is not active when the internally coded void fraction model is used.

Name Default Units Description WSI {0} Ibm/s Mass flow rate of pumped injection entering the upper plenum that interacts with steam. The default value is typically used for hot leg switchover calculations and cycling calculations during cold leg injection; refer to Section 8.0 for guidance on performing cycling calculations during hot leg injection.

HSI (-1.0} Btu/lbm Enthalpy of pumped injection entering the upper plenum that interacts with steam (e.g. HSI = 148 Btu/lbm for P = 14.7 psia and T = 180 °F). If no credit for upper plenum subcooling is taken, then the default value should be used.

4.8 Core Pressure PCORE allows the user to specify a constant core pressure between 14.7 and 200 psia (inclusive).

Name Default Units Description PCORE {14.7} psia Core pressure, assumed to remain constant with time. Must be between 14.7 and 200 psia (inclusive).

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 23 of 27 4.9 Core-to-Lower Plenum Boric Acid Transport Model The transport model is limited to applications in which the pressure is 14.7 psia. Also, if subcooling is provided as input to the transport model, the subcooling is only considered in the transport model, but not in other calculations such as void fraction.

Name Default Units Description 4.10 Liquid Carryover Model This model is intended for advanced plant or non-standard applications.

Name Default Units Description ENTX {1.0} No units The value for the two-phase mixture mass quality exiting the core region during the transient. ENTX

= 1.0, only dry steam exiting the core. ENTX < 1.0, wet steam exiting the core.

Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1, 2017 Page 24 of 27 5.0 Output Description SKBOR generates both an ASCII output file and a binary graphics output file. The ASCII output is self-explanatory and does not warrant further discussion here. Section 5.1 describes the contents of the binary graphics output file.

5.1 Graphics Output Binary File The following variables are written to the graphics output binary file:

Variable Units Description T s Transient time.

PFRAC Normalized core power fraction.

RHOCORE Ibm/ft^ Core fluid density.

WBOIL Ibm/s Average core exit mass flow rate over timestep DT.

MBCORE Ibm Core boron mass.

MWCORE Ibtn Core water mass.

MTCORE Ibm Core total (i.e., boron + water) mass.

MBSUMP Ibm Sump boron mass.

MWSUMP Ibm Sump water mass.

MTSUMP Ibm Sump total (i.e., boron + water) mass.

WTFCORE w/f Core boron concentration, in weight fraction.

WTPCORE w/o Core boron concentration, in weight percent PPMCORE ppm Core boron concentration, in ppm WTFSUMP w/f Sump boron concentration, in weight fraction.

WTPSUMP w/o Sump boron concentration, in weight percent PPMSUMP ppm Sump boron concentration, in ppm DILSUMP ppm Sump dilution, in ppm.

PCORE psia Core pressure HEATF Btu/s-ft^ Heat Flux JFIN ft/s Superficial Velocity JGOUT ft/s Superficial Velocity COREVOID no units Void Fraction UPVOID no units Void Fraction VCORE ft^ Core mixing volume TCORE °F Core Fluid Temperature WTFfN w/f Boron Injection Weight Fraction WTPIN w/o Boron Injection Weight Percent PPMIN ppm Boron Injection ppm Methods, Technology, & Licensing Memo Template Version I -0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1, 2017 Page 25 of 27 Variable Units Descriotion TOTBINJ Ibm Total Boron Mass Injected TOTWINJ Ibm Total Water Mass Injected TOTMINJ Ibm Total Mass Injected TLP °F Lower Plenum Fluid Temperature RHOLP ibm/ft^ Density of Liquid in Lower Plenum MBLP Ibm Mass of Boron in Lower Plenum MWLP Ibm Mass of Water in Lower Plenum MTLP Ibm Total Mass in Lower Plenum WTFLP w/f Weight Fraction of Boron in Lower Plenum WTPLP w/o Weight Percent of Boron in Lower Plenum PPMLP ppm Boron ppm in Lower Plenum WBF Ibm/s Core-to-Lower Plenum Exchange Rate Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 26 of 27 6.0 Summary of SKBOR Releases Table 6-1 summarizes the releases of SKBOR on UNIX and Linux platforms.

Table 6-1: SKBOR Releases Version Release Date 3.1 5/94 4.0 4/99 5.0 7/00 6.0 6/01 7.0 11/01 8.0 11/05 9.0 5/07 10.0 11/13 Methods, Technology, & Licensing Memo Template Version 1-0

Westinghouse Non-Proprietary Class 3 MT-17-56, Rev. 1, Attachment 5 June 1,2017 Page 27 of 27 7.0 References

1. J.P. Cunningham and H.C. Yeh, Experiments and Void Correlation for PWR Small-Break LOCA Conditions, Trans. ANS 17, p. 369-370 (1973).

2. L.E. Hochreiter and H.C. Yeh, Mass Effluence During FLECHT Forced Reflood Experiments, Nuclear Engineering and Design, 60, p. 413-429 (1980).

3. H.C. Yeh, Modification of Void Fraction Calculation, Proceedings of the Fourth International Topical Meeting on Nuclear Thermal-Hydraulics, Operations and Safety, Volume 1, Taipei, Taiwan, June 6, 1988.

Methods, Technology, & Licensing Memo Template Version 1-0

Enclosure Response to NRC Staff RAIs Regarding the NGF LAR Attachment 6 Affidavit from the Westinghouse Electric Company Submitted in Accordance with 10 CFR 2.390 to Consider Attachments 1, 3, and 4 as Proprietary CAW-17-4595, June 1, 2017

Westinghouse Non-Proprietary Class 3 Westinghouse Electric Company Westinghouse 1000 Westinghouse Drive Cranberry Township, Pennsylvania 16066 USA U.S. Nuclear Regulatory Commission Direct tel: (412)374-4643 Document Control Desk Direct fax: (724)940-8560 11555 Rockville Pike e-mail: greshaja@westinghouse.com Rockville, MD 20852 CAW-174595 June 1,2017 APPLICATION FOR WITHHOLDING PROPRIETARY INFORMATION FROM PUBLIC DISCLOSURE

Subject:

MT-17-56, Revision 1, Westinghouse responses to the NRC RAIs on the Palo Verde 1,2, and 3 NGF LAR and Exemption (ADAMS Accession No. ML17107A005), Revision 1 Attachments 1,3, and 4 (Proprietary)

The Application for Withholding Proprietary Information from Public Disclosure is submitted by Westin^ouse Electric Company LLC (Westinghouse), pursuant to the provisions of paragraph (bXl) of Section 2.390 of the Nuclear Regulatoiy Commissions (Commissions) regulations. It contains commercial strategic information proprietary to Westinghouse and customarily held in confidence.

The proprietary information for which withholding is being requested in the above-referenced documents is filler identified in Affidavit CAW-174595 signed by the owner of the proprietary information, Westinghouse. The Affidavit, which accompanies this letter, sets forth the basis on which the information may be withheld from public disclosure by the Commission and addresses with specificity the considerations listed in paragraph (bX4) of 10 CFR Section 2.390 of the Commissions regulations.

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

Correspondence with respect to the proprietary aspects of the Application for Withholding or the Westinghouse Affidavit should reference CAW-174595, and should be addressed to James A. Gresham, Manager, Regulatory Compliance, Westinghouse Electric Company, 1000 Westinghouse Drive, Building 3 Suite 310, Cranberry Township, Pennsylvania 16066.

James A. Gresham, Manager Regulatory Compliance

CAW-17-4595 AFFIDAVIT COMMONWEALTH OF PENNSYLVANIA:

COUNTY OF BUTLER:

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

Executed on; bj/il James A. Gresham, Manager Regulatory Compliance

CAW-17-4595 I am Manager, Regulatory Compliance, Westinghouse Electric Company LLC (Westinghouse),

and as such, 1 have been specifically delegated the function of reviewing the proprietary information sought to be withheld from public disclosure in connection with nuclear power plant licensing and rule making proceedings, and am authorized to apply for its withholding on behalf of Westinghouse.

I am making this Affidavit in conformance with the provisions of 10 CFR Section 2.390 of the Nuclear Regulatory Commissions (Commissions) regulations and in conjunction with the Westinghouse Application for Withholding Proprietary Information from Public Disclosure accompanying this Affidavit.

I have personal knowledge of the criteria and procedures utilized by Westinghouse in designating information as a trade secret, privileged or as confidential commercial or financial information.

Pursuant to the provisions of paragraph (b)(4) of Section 2.390 of the Commissions regulations, the following is furnished for consideration by the Commission in determining whether the information sought to be withheld from public disclosure should be withheld.

(i) The information sought to be withheld from public disclosure is owned and has been held in confidence by Westinghouse.

(ii) The information is of a type customarily held in confidence by Westinghouse and not customarily disclosed to the public. Westinghouse has a rational basis for determining the types of information customarily held in confidence by it and, in that connection, utilizes a system to determine when and whether to hold certain types of information in confidence. The application of that system and the substance of that system constitute Westinghouse policy and provide the rational basis required.

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

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

4 CAW-17-4595 Westinghouses competitors without license from Westinghouse constitutes a competitive economic advantage over other companies.

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

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

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

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

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

(iii) There are sound policy reasons behind the Westinghouse system which include the following:

(a) The use of such information by Westinghouse gives Westinghouse a competitive advantage over its competitors. It is, therefore, withheld from disclosure to protect the Westinghouse competitive position.

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

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

CAW-17-4595 Each component of proprietary information pertinent to a particular competitive advantage is potentially as valuable as the total competitive advantage. If competitors acquire components of proprietary information, any one component may be the key to the entire puzzle, thereby depriving Westinghouse of a competitive advantage.

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

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

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

The information sought to be protected is not available in public sources or available information has not been previously employed in the same original manner or method to the best of our knowledge and belief.

(Vi) The proprietary information sought to be withheld in this submittal is that which is appropriately marked in MT-17-56, Revision 1, Westinghouse responses to the NRC RAls on the Palo Verde 1, 2, and 3 NGF LAR and Exemption (ADAMS Accession No. ML17107A005), Revision 1 Attachments 1, 3, and 4 (Proprietary), for submittal to the Commission, being transmitted by Arizona Public Services (APS) letter. The proprietary information as submitted by Westinghouse is that associated with the NRC review of the Palo Verde 1, 2, and 3 NGF LAR and Exemption, and may be used only for that purpose.

This information is part of that which will enable Westinghouse to support APS for the use of Combustion Engineering 16x16 Next Generation Fuel in the Palo Verde Units.

6 CAW-17-4595 (b) Further, this information has substantial commercial value as follows:

(i) Westinghouse plans to sell the use of similar information to its customers for the purpose of transitioning them to Combustion Engineering 16x16 Next Generation Fuel.

(ii) Westinghouse can sell support and defense of industry guidelines and acceptance criteria for plant-specific applications.

(iii) The information requested to be withheld reveals the distinguishing aspects of a methodology which was developed by Westinghouse.

Public disclosure of this proprietary information is likely to cause substantial harm to the competitive position of Westinghouse because it would enhance the ability of competitors to provide similar technical evaluation justifications and licensing defense services for commercial power reactors without commensurate expenses. Also, public disclosure of the information would enable others to use the information to meet NRC requirements for licensing documentation without purchasing the right to use the information.

The development of the technology described in part by the information is the result of applying the results of many years of experience in an intensive Westinghouse effort and the expenditure of a considerable sum of money.

In order for competitors of Westinghouse to duplicate this information, similar technical programs would have to be performed and a significant manpower effort, having the requisite talent and experience, would have to be expended.

Further the deponent sayeth not.

PROPRIETARY INFORMATION NOTICE Transmitted herewith are proprietary and non-proprietary versions of a document, furnished to the NRC associated with the NRC review of the Palo Verde 1, 2, and 3 NGF LAR and Exemption, and may be used only for that purpose.

In order to conform to the requirements of 10 CFR 2.390 of the Commissions regulations concerning the protection of proprietary information so submitted to the NRC, the information which is proprietary in the proprietary versions is contained within brackets, and where the proprietary information has been deleted in the non-proprietary versions, only the brackets remain (the information that was contained within the brackets in the proprietary versions having been deleted). The justification for claiming the information so designated as proprietary is indicated in both versions by means of lower case letters (a) through (f) located as a superscript immediately following the brackets enclosing each item of information being identified as proprietary or in the margin opposite such information. These lower case letters refer to the types of information Westinghouse customarily holds in confidence identified in Sections (4)(ii)(a) through (4)(ii)(f) of the Affidavit accompanying this transmittal pursuant to 10 CFR 2.390(b)(1).

COPYRIGHT NOTICE The reports transmitted herewith each bear a Westinghouse copyright notice. The NRC is permitted to make the number of copies of the information contained in these reports which are necessary for its internal use in connection with generic and plant-specific reviews and approvals as well as the issuance, denial, amendment, transfer, renewal, modification, suspension, revocation, or violation of a license, permit, order, or regulation subject to the requirements of 10 CFR 2.390 regarding restrictions on public disclosure to the extent such information has been identified as proprietary by Westinghouse, copyright protection notwithstanding. With respect to the non-proprietary versions of these reports, the NRC is permitted to make the number of copies beyond those necessary for its internal use which are necessary in order to have one copy available for public viewing in the appropriate docket files in the public document room in Washington, DC and in local public document rooms as may be required by NRC regulations if the number of copies submitted is insufficient for this purpose. Copies made by the NRC must include the copyright notice in all instances and the proprietary notice if the original was identified as proprietary.

ML17153A373: Difference between revisions

Latest revision as of 16:34, 4 February 2020

Contents

Text

Dear Sirs:

Enclosure:

Response

References:

Response

Response

Response

Response

References:

Response

References:

Response

References:

Response

References:

Subject:

Navigation menu

ML17153A373: Difference between revisions

Latest revision as of 16:34, 4 February 2020

Text

Dear Sirs:

Enclosure:

Response

References:

Response

Response

Response

Response

References:

Response

References:

Response

References:

Response

References:

Subject:

Navigation menu

Search