12/29/2016 Summary of Pre-Application Meeting with Arizona ... · 12/29/2016 Summary of Pre-Application Meeting with Arizona Public Service to Discuss a Proposed Second Emergency

UNITED STATES NUCLEAR REGULATORY COMMISSION

WASHINGTON, D.C. 20555-0001

Mr. Robert S. Bement Executive Vice President Nuclear/ Chief Nuclear Officer

Mail Station 7602 Arizona Public Service Company P.O. Box 52034 Phoenix, AZ 85072-2034

J a nuary 4, 2017

SUBJECT: SUMMARY OF DECEMBER 29, 2016, PRE-APPLICATION TELECONFERENCE WITH ARIZONA PUBLIC SERVICE COMPANY TO DISCUSS A PROPOSED SECOND EMERGENCY LICENSE AMENDMENT REQUEST TO REVISE TECHNICAL SPECIFICATION 3.8.1, "AC [ALTERNATING CURRENT] SOURCES - OPERATING," FOR PALO VERDE NUCLEAR GENERATING STATION, UNIT 3 (CAC NO. MF9019)

Dear Mr. Bement:

On December 29, 2016, a pre-application teleconference was held between the U.S. Nuclear Regulatory Commission (NRC) and representatives of Arizona Public Service Company (APS, the licensee) via teleconference. The purpose of the meeting was for APS staff to provide information on a proposed second emergency license amendment request (LAR) to modify Technical Specification (TS) 3.8.1, "AC Sources - Operating ," for Palo Verde Nuclear Generating Station (PVNGS), Unit 3. Due to the emergent and time-sensitive nature of NRC's interaction with the licensee, it would have created an undue administrative burden and would have impeded the effective implementation of the Agency's official regulatory responsibilities or responsiveness to make arrangements for public notification in the time available to open the teleconference to the public, as discussed in Management Directive 3.5, "Attendance at NRC Staff-Sponsored Meetings" (Agencywide Documents Access and Management System (ADAMS) Accession No. ML041270381 ). A list of teleconference participants is provided in Enclosure 1 to this meeting summary. The licensee electronically sent the meeting slides to the NRC Project Manager for PVNGS, which can be found in Enclosure 2.

During surveillance testing on December 15, 2016, the PVNGS Unit 3 Train B emergency diesel generator (DG) suffered a failure of the number nine right cylinder connecting rod and piston . On December 23, 2016, NRC issued Amendment No. 199 for PVNGS, Unit 3, in response to APS's application dated December 21 , 2016, as supplemented by letter dated December 23, 2016 (ADAMS Accession Nos. ML 16358A676, ML 16356A689, and ML 16358A715, respectively). The emergency amendment revised the applicable TS for a one-time extension of the emergency DG completion time (CT) described in TS 3.8.1.B.4 based on defense-in­depth guidance and deterministic criteria . Specifically, the emergency amendment extended the TS Required Action 3.8.1.B.4 CT from 10 days to 21 days for the purpose of collecting and analyzing data associated with the failure of train B DG and continue with the repair of the DG. The 21-day CT expires on January 5, 2017, at approximately 4:00 a.m. Mountain Standard Time.

R. Bement - 2 -

The licensee stated that it will propose a second emergency LAR to further extend the DG CT described in TS 3.8.1 .B.4 from 21 days to 62 days to complete the DG repairs and required DG testing . The licensee added that it was planning to submit the emergency LAR on December 30, 2016, and to request approval prior to the expiration of the current CT on January 5, 2017.

In its presentation , APS provided information on the status of the regulatory commitments documented in Amendment No. 199. A list of several of the licensee's commitments are shown on slide number 4 in Enclosure 2. APS then provided an overview of the AC power system at PVNGS, including the Train A and B DGs, two station blackout generators (SBOGs) and three portable DGs. The licensee described its event investigation, including its research on other Cooper-Bessemer DG failures and the proximate cause of the PVNGS Unit 3 Train B DG failure . The licensee stated that the proximate cause of fa ilure was high-cycle fatigue failure of the master connecting rod ligament which surrounds the lower part of the articulating rod pin. The licensee identified three potential flaws, which in addition to the stress of the mis-alignment of the crankshaft, led to the fatigue failure of the DG. The licensee discussed the findings and history of the PVNGS Unit 3 Train B DG, which can be seen on slide number 12. The NRC staff asked about the extent of the repairs and requested that the licensee include in its application a timeline of activities including repairs and testing. APS provided information on the comparative evaluation of the PVNGS, Unit 3 Train A and Train B DG and stated that the Train A engine has not experienced a catastrophic failure, unlike Train B had in 1986. The licensee stated that there is no common cause mode of failure to the PVNGS Unit 3 Train A DG due to the unique aspects of the Unit 3 Train B DG root cause.

APS provided an overview of the risk assessment associated with the emergency LAR, including probabilistic risk assessment (PRA) models for internal events, internal flood , internal fire , and seismic hazards. Prior to the meeting, the NRC Project Manager for PVNGS provided a set of talking points (technical issues) to the licensee, based on the PRA information contained in Enclosures 1 and 2 of the previous emergency LAR dated December 21 , 2016. The PRA-related talking points can be found in Enclosure 3 to this meeting summary. In the pre-submittal meeting, the licensee stated it will provide an attachment to the emergency LAR that will contain information addressing the technical issues in Enclosure 3. The NRC staff asked several questions during the meeting about the PRA models and requested that the licensee provide sufficient PRA information - including sensitivity studies - in its application to allow NRC staff to minimize its request for additional information and complete its review given the extremely challenging timeframe. The staff also stated that the licensee needed to provide a strong basis for why an additional 41 days is necessary (i.e., a total of 62 days for the repairs) and a justification for why APS is confident in the root cause and in its conclusion that there is no common cause failure that could similarly impact the other DGs.

R. Bement - 3 -

If you have any questions, please contact me at (301) 415-1564 or via e-mail at [email protected].

Docket No. STN 50-530

Enclosures: 1. List of Teleconference Participants 2. APS Teleconference Slides

Sincerely,

rn w aJJ;J far Siva P. Lingam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

3. NRC Talking Points (Potential Technical Issues)

cc w/encl : Distribution via Listserv

LIST OF TELECONFERENCE PARTICIPANTS

DECEMBER 29, 2016. MEETING WITH ARIZONA PUBLIC SERVICE COMPANY

REGARDING PROPOSED EMERGENCY LICENSE AMENDMENT REQUEST

PALO VERDE NUCLEAR GENERATING STATION, UNIT 3

DOCKET NO. STN 50-530

U.S. Nuclear Regulatory Commission

Maggie Watford Anne Boland John Lubinski

Mirela Gavrilas Jeff Clark Jeff Circle C.J . Fong

Todd Hilsmeier Margaret Chernoff

Vijay Goel Matthew Hamm

John Huang Mihaela Biro

Shilp Vasavada Victoria Huckabay

Tom Wengert Jennifer Whitman Diana Woodyatt

Jeff Clark Geoff Miller Mike Levine

Charley Peabody Bob Wolfgang Siva Lingam

Arizona Public Service Company

Jack Cadogan Chuck Kharrl Bruce Rash

Michael Mclaughlin Kevin Graham

Pete McSparran George Andrews Thomas Weber

Greg Brown Tom Romay

Jared Schank Tom Hook

Del Elkinton Lorraine Weaver

Nawaporn Aaronscooke Carl Stephenson

Sean Dornseif Maria Lacal

Mark McGhee Mike Dilorenzo

Enclosure 1

Enclosure 2

ARIZONA PUBLIC SERVICE COMPANY

MEETING SLIDES FOR PRE-APPLICATION TELECONFERENCE ON DECEMBER 29, 2016

ADAMS Accession No. ML 17003A033

Agenda

• Background • Regulatory Commitments • Event Investigation • Risk Assessment • Second License Amendment Request • Conclusions

First License Amendment Request (LAR)

• One-time Technical Specification (TS) Change to Allow a 21 day Completion Time In Response to Failure of Unit 3 B Train DG on December 15, 2016 - Extension of 11 days needed to collect/analyze data and

continue repair - Deterministic justification based upon BTP 8-8 - Risk insights provided to support change - NRC commitments made in LAR - NRC Amendment 199 issued on December 23, 2016

Regulatory Commitments • Commitments documented in NRC

Amendment # 199 include but are not limited to: - Three, 2 MW portable DGs staged, tested and

hooked-up to Unit 3 FLEX 4.16KV connections - Diesel driven FLEX Steam Generator make-up

pump staged in Unit 3 - Suspension of discretionary maintenance on

SBOGs, Switchyard, Safety Systems - Establish protected equipment controls for Train A

equipment, SBOGs, portable equipment - Commitments monitored and tracked by OPS - Dedicated personnel

Palo Verde AC Power System SBOGs

NAN-S03 ( SOSA

( S03A

I _ _...._ _ _ I

Ul. U2 S03AB

"'--w NBN-X03 4160V

"fl:' EOG

PBA-S03

OFF SITE POWER

( SOGA NAN·S04 I a 8 k\/ ( S04A 13. ~

NBN-X04 t'YYY"'I 4160 V

~ ( S04B

"B" EOG

4160V -+---+---..._- PBB-504 Portable Diesel

Generators

~ n I nm rryn CL36B2 (L34B2 ( l32B2

Palo Verde AC Power System

( jl.ix-i-1uN-io-/,88-so4 1i cl.Ass sus--)

FLEX + 1

IT\_n I w. 'C::F7KV

I Control System : Communication

FLEX+ 2 I Unk

13.8 KV PREFERRED

SOURCE

~NBN·X04

IT\_n I I ~ I I ...... I I 'C:/F7Kv

I Control System : communication

RENTAL I Unk I

I©-.: I i~ F

TERMINAL BOX

TRANSFER SWITCH IN PBB-504

SWGRROOM I) S04K i' TRAIN B 4.16 KV CLASS 1E BUS

" BH EOG

S04B

PBB-504

Event Investigation

• Partnerships established with MPR, Goltens, Structural Integrity, EPRI, and the Cooper­Bessemer Owners Group

• Evidence of high cycle fatigue on master connecting rod

• Second major failure of 38 DG (9R) • 1986 event created localized misalignment

Operating Experience (OE)

• Cooper-Bessemer KSV-20 OE - 1986 - Palo Verde 38 DG connecting rod (9R)

failure during unit startup testing program - 1989 - South Texas Project DG 22 connecting rod

failure during a surveillance test - 2003 - South Texas Project DG 22 connecting rod

failure during a surveillance test (one-time LAR using a two-phased approach to extend allowable outage time to 113 days)

- 2016 - Palo Verde 38 DG connecting rod (9R) failure during a surveillance test

DIRECT CAUSE OF FAILURE

• High cycle fatigue failure of the master connecting rod ligament which surrounds the lower part of the articulating rod pin.

6S-76A

@ <D

® @

1. Piston Pin 2.Washer 3. Piston Pin Bolt 4. Articulated Rod 5. Articulated Rod Pin Bolt 6.Washer 7.Dowel 8. Oil Passage 9. Articulated Rod Pin

10. Bushing Dowel Pin 11.Bushing 12.Dowel 13. Connecting Rod Cap 14.Locknut 15.Stud 16. Bearing Shell 17. Master Rod 18. Nut Tightening

.-->--------------.Sequence Ligament where crack originated

c 0

·- ..., U)

cu >

c ..... ..., c cu >

w

ROOT CAUSE OF FAILURE

FLAW+ STRESS

POTENTIAL FLAWS

1. Residual tensile within master connecting rod bore due to machining process change

2. Fretting

3. "Undersized" Oversized Bearing following 1987 repair.

FATIGUE

STRESS

Mis-Alignment

Unit 3 DG 'B' Findings and History HISTORY Generator - '87-'88 loose Generator poles

- sheared epoxy

I 2016 1-l!~~rMain Bearing

THREE cracks on Center-frame - near #8, #9, and #1 O

inspection windows - cracks seem old

~E, I 201tf]-

l #9 Master Rod Failure

"" I

HISTORY #9 Master Rod - '86 failed due

to iron plating - '87 replaced rod - turned down #9 journal - installed bigger bearing - stitch repair/bracing

HISTORY #6 Art Rod - '93 bent due to

over-pressure

#4 Master Rod Fret/micro crack

HISTORY #2 Main Bearing - '86 high Temp. after event - '87 replaced (5x)

'\II

#2 Main Bearing wear

#2 Master Rod FreUmicro crack

3A/3B COMPARITIVE EVALUATION

• Evaluating Wide Array of Data • Relevant Data

- Vibration - Engine Analysis - Line Bore Data - Work History - Event History

3A/3B COMPARITIVE EVALUATION

• Unit 3 "B" Emergency Diesel Generator experienced a catastrophic failure that induced crankshaft mis-alignment which increased the stress profile within the engine

• Unit 3 "A" engine has not had a catastrophic fa i I u re

• Unit 3 "A" engine vibration displacement data is consistently less and has significantly less va ria bi I ity

• Unit 3 "A" Master connecting rods are original equipment (i.e. Pre machining change)

Engineering Conclusion

There is no common cause mode of failure to Unit 3 "A" Emergency Diesel Generator due to the unique aspects of the Unit 3 "B'' Diesel Generator root cause.

Risk Assessment

• PRA models for - Internal Events - Internal Flood - Internal Fire - Seismic

• Other hazards screened out

PRA Model and Risk-Informed Application Model History

Internal Events CEOG peer review & numerous risk-informed TS changes

Internal Flood peer review

Risk-informed 7-day inverter TS approved

Internal Events self-assessment per RG 1.200 App B

TSTF-425 Surveillance Frequency Control Program approved

-- External Hazards Screening peer review

pt fire PRA peer review

Seismic PRA peer review

2nd fire PRA peer review

2013

TSTF-505 submitted

All Unit 3 Mods Comp & all ASME PRA Std SRs Met to CC II

Risk Assessment • Palo Verde PRA Aspects

- Six 100°/o capacity SG makeup pumps all supplied by onsite power sources

- Only one of these powered by B DG if loss of offsite power - RCP seal LOCAs negligible - ECCS significance minimal in

loss of offsite power events - No Pressurizer power-operated relief valves - Very low internal events CDF and LERF - consistent with STP

and Millstone 3 - Only shared systems in PRA are SBOGs and firewater - Dedicated fire department staff and equipment - Risk significant FLEX connections outside of unit - Did not need to implement NFPA-805 to address multiple

spurious operations

PRA Model Credited Changes

• Revised emergency operating procedures and night order to direct timely use of firewater to auxiliary feedwater cross-tie in total loss of feedwater event - validated in simulator - Additional dedicated auxiliary operator added to each shift to

implement cross-tie

• Post continuous fire watch in fire zone FCCOR2 (120' Corridor Building)

• Establish new transient combustible and hot work exclusion zones and conduct shiftly inspections - Fire zones FCCOR2 (120' Corridor Building) and FCCOR2A (120'

Corridor Riser Shaft) - Fire zones FCTB04 (upper level only, non-class DC Equipment,

[FCTB04-TRAN l]) - Fire zone FC86A (train A Seismic Gap, make part of train A Electrical

Protected Equipment) - Fire zone FCTBlOO zone ZTlG (SW corner, south half of 100' Turbine

between columns TA and TC)

Risk Assessment

• Defense-in-Depth Evaluation - Unavailability does not reduce the amount of

available equipment to a level below that necessary to mitigate a design basis accident

• Safety Margin Evaluation - No significant reduction in margin of safety

• Regulatory Guide 1.200, Revision 2 compliant • Regulatory Guide 1.177, Revision 1 compliant • Regulatory Guide 1.174, Revision 2 compliant

Second License Amendment Request

• Requesting a extension of TS 3.8.1 Condition B.4 Completion Time to complete the DG Repairs

• Request on Emergency Basis • Risk-informed LAR • Carrying forward the Commitments made in

Deterministic LAR • To be submitted ·Friday, December 30 • Request approval by early Thursday morning

Conclusions

• Direct cause of the 3B DG failure has been determined

• No common mode failure applicability to 3A DG

• Continue to have diverse and redundant sources of AC power and steam generator makeup

• PRA risk acceptable in accordance with Regulatory Guides 1.174 and 1.177

• No significant hazards consideration criteria satisfied

Potential Technical Issues Regarding the Palo Verde, Unit 3, License Amendment Request for a One-Time Extension of the 38 Emergency Diesel Generator Completion

Time (CAC No. MF9019)

Discussion: Palo Verde submitted a deterministic license amendment request (LAR) on 12121116 requesting a completion time (CT) extension for emergency diesel generator (EOG) 38. Although not used by the staff to make a safety decision, the licensee chose to provide risk information in Enclosures 1 and 2 of that LAR. The licensee also stated an intent to submit a risk-informed LAR in the future containing or referencing risk insights similar to what was provided in the aforementioned enclosures provided that common cause issues could be resolved. If I when that LAR is submitted (date is to be determined), the following are potential technical issues that may need to be evaluated. The list is not meant to be exhaustive nor does it represent an official agency position as the licensee has not yet decided whether to submit a risk-informed LAR.

Potential Issue 1

The license amendment request (LAR) for Palo Verde Nuclear Generating Station (PVNGS) , Unit 3, dated December XX, 2016, states that the plant-specific risk assessment of the proposed change to the Technical Specification (TS) completion time (CT) follows the guidance in Regulatory Guide (RG) 1.17 4, Revision 2, and RG 1.177, Revision 1. Both of these regulatory guides endorse the guidance in RG 1.200, Revision 2, as an acceptable approach for determining whether the technical adequacy of the PRA is sufficient for use in regulatory decision-making (e.g. , changes to a plant's licensing basis).

Section 4.2, "Licensee Submittal Documentation ," of RG 1.200 provides detailed guidance on what information should be included in a risk-informed submittal to demonstrate the technical adequacy of the PRA, including a discussion of the resolution of peer review (or self­assessment, for peer reviews performed using the criteria in NEI 00-02) findings and observations that are applicable to the submittal. Also stated in RG 1.200 is that the objective of the peer review is to demonstrate that the requirements in an NRG-endorsed standard (e.g., ASME/ANS RA-Sa-2009) have been met.

Attachment 4 of Enclosure 2 to the LAR lists those "Findings" from the peer review of the internal events PRA model (IEPRA) conducted in accordance with NEI 00-02. It is unclear whether these findings were dispositioned in a manner that demonstrates that the requirements in ASME/ANS RA-Sa-2009 (the "ASME PRA Standard") have been met. Therefore, address the following items related to these findings :

(a) Finding HR-01 cites concern that operation input into the human reliability analysis (HRA) may be marginal. The corresponding disposition explains that this finding was addressed by updating the HRA documentation, but does not explain whether the degree of operational input and review meets PRA standard ASME/ANS RA-Sa-2009, as qualified by RG 1.200, Revision 2. Supporting requirement (SR) HR-E3 of ASME/ANS RA-Sa-2009 requires "talk-throughs" of the procedures with plant operation and training personnel to ensure a consistent interpretation . Explain whether "talk-throughs" (i.e., detailed review) of the procedures with plant operation and training personnel were performed.

Enclosure 3

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Otherwise, justify why not performing "talk-throughs" is judged to have no significant impact on the quantification results used in this application.

(b) Finding HR-03 cites concern about not modelling "miscalibration and common cause miscalibration of critical sensors." The corresponding disposition states that common cause modelling was updated in the PRA to "match the NRC common cause database treatment. " It is not clear from this statement how this finding was resolved in the PRA. Clarify how miscalibration errors were resolved in the IEPRA.

(c) Finding AS-03 asks why the plant response to small loss of coolant accidents (LOCAs) and induced small LOCAs were modelled differently. The corresponding disposition states that the finding has been resolved and closed by an update of the PRA model and documentation. Describe the update of the IEPRA model to resolve this finding , and, if applicable, explain and justify why the plant responses are different for these LOCAs.

Potential Issue 2

The PVNGS LAR, dated December XX, 2016, states that the plant-specific risk assessment of the proposed change to the TS CT follows the guidance in RG 1.17 4, Revision 2, and RG 1.177, Revision 1. Both of these regulatory guides endorse the guidance in RG 1.200, Revision 2, as an acceptable approach for determining whether the technical adequacy of the PRA is sufficient for use in regulatory decision-making (e.g., changes to a plant's licensing basis).

Section 4.2, "Licensee Submittal Documentation ," of RG 1.200 provides detailed guidance on what information should be included in a risk-informed submittal to demonstrate the technical adequacy of the PRA, including a discussion of the resolution of peer review (or self­assessment, for peer reviews performed using the criteria in NEI 00-02) findings and observations that are applicable to the submittal. Also stated in RG 1.200 is that the objective of the peer review is to demonstrate that the requirements in an NRG-endorsed standard (e.g., ASME/ANS RA-Sa-2009) have been met.

Attachment 7 of Enclosure 2 to the LAR lists facts and observations (F&Os) from the peer review of the seismic PRA model. The disposition to seismic PRA F&O SPR-B10 indicates that the finding has been resolved , but does not discuss the resolution. Discuss how this F&O was resolved.

Potential Issue 3

The PVNGS LAR, dated December XX, 2016, states that the plant-specific risk assessment of the proposed change to the TS CT follows the guidance in RG 1.17 4, Revision 2, and RG 1.177, Revision 1. Both of these regulatory guides endorse the guidance in RG 1.200, Revision 2, as an acceptable approach for determining whether the technical adequacy of the PRA is sufficient for use in regulatory decision-making (e.g., changes to a plant's licensing basis) .

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Section 4.2, "Licensee Submittal Documentation," of RG 1.200 provides detailed guidance on what information should be included in a risk-informed submittal to demonstrate the technical adequacy of the PRA, including a discussion of the resolution of peer review (or self­assessment, for peer reviews performed using the criteria in NEI 00-02) findings and observations that are applicable to the submittal. Also stated in RG 1.200 is that the objective of the peer review is to demonstrate that the requirements in an NRC-endorsed standard (e.g., ASME/ANS RA-Sa-2009) have been met.

The LAR discusses the peer review of the seismic PRA and fire PRA. The staff also reviewed information provided to the NRC by the licensee in its Risk-Informed Completion Time (RICT) application dated July 31 , 2015 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML 15218A300). It is unclear to the staff whether these peer reviews were full-scope reviews and what guidance documents were used to perform them. For example, the RICT application states the following , which suggests the 2012 peer review of the fire PRA was not a full-scope review:

A peer review of the PVNGS internal fire PRA was conducted in October 2012 ... Subsequently, a focused-scope peer review of the internal fire PRA was conducted in December 2014 (Reference 11 of this Attachment) to address ASME PRA Standard SRs not-met to Capability Category II requirements and those SRs not-reviewed in the prior October 2012 internal fire PRA peer review.

Confirm that peer reviews performed for the seismic PRA and fire PRA were full -scope reviews meeting industry guidance for a peer review and that they were reviewed against capability category II (in accordance with RG 1.200). In addition, discuss which organization performed the review, and list the guidance documents followed for each review, including the guidance used for the peer review process (e.g., NEI 07-12, "Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines"). As applicable, provide the F&Os, including their dispositions, from the 2014 focused-scope peer review of the fire PRA determined not met to Capability Category 11 .

Potential Issue 4

The PVNGS LAR, dated December XX, 2016, states that the plant-specific risk assessment of the proposed change to the TS CT follows the guidance in RG 1.17 4, Revision 2, and RG 1.177, Revision 1. Both of these regulatory guides endorse the guidance in RG 1.200, Revision 2, as an acceptable approach for determining whether the technical adequacy of the PRA is sufficient for use in regulatory decision-making (e.g., changes to a plant's licensing basis) . RG 1.200 endorses, with clarifications and qualifications, the ASME PRA Standard. Section 5-2.3, "Seismic Plant Response Analysis ," of the ASME PRA Standard states:

The restoration of safety functions can be inhibited by any of several types of causes; these include damage or failure , access problems, confusion , loss of supporting staff to other post[-]earthquake-recovery functions, and so on . Careful consideration of these must be given before recoveries are credited in the initial period after a large earthquake. This is especially true for earthquake­caused loss of off-site power (LOSP), given that the damage could be to switchyard components or to the off-site grid towers, which are generally difficult

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to fix quickly. While this part does not require the analyst to assume an unrecoverable LOSP after a large earthquake, the general practice in seismic PRAs has been to make such an assumption .

Attachment 11 of Enclosure 2 to the LAR discusses a seismic assumption/uncertainty regarding LOSP recovery. The licensee states:

It is realistic to consider that offsite power recovery is available for low magnitude seismic events. The selection of the [safe shutdown earthquake] SSE as a threshold between recovery/no-recovery of offsite power is arbitrary and conservative. Therefore, no sensitivity analysis is required for this application .

Provide additional justification for this assumption and why it is considered conservative. Explain whether it is conservative in terms of baseline risk (i.e., CDF and LERF) or delta risk (i.e., ICCDP and ICLERP). Include in this discussion your assumptions about damage to switchyard components, offsite power transformers, or to the off-site grid towers, which are generally difficult to fix quickly. Alternatively, alter the credit for offsite power recovery in the seismic PRA as part of the sensitivity analysis.

Potential Issue 5

The PVNGS LAR, dated December XX, 2016, states that the plant-specific risk assessment of the proposed change to the TS CT follows the guidance in RG 1.17 4, Revision 2, and RG 1.177, Revision 1. Section 2.5.5, "Comparisons with Acceptance Guidelines," of RG 1.17 4 states that when the contributions from the contributors modeled in the PRA are close to the risk acceptance guidelines, the argument that the contribution from the missing items is not significant must be convincing and in some cases may require additional PRA analyses (e.g., bounding analyses, detailed analyses, or by a demonstration that the change has no impact on the unmodeled contributors to risk) . When the margin is significant, a qualitative argument may be sufficient. In addition, Section 2.5.3, "Model Uncertainty," of RG 1.17 4 states that the impact of using alternative assumptions or models may be addressed by performing appropriate sensitivity studies or by using qualitative arguments.

Section 2.3 of Enclosure 2 to the LAR states that the fire PRA model is consistent with the NUREG/CR-6850 (dated September 2005) methodology with "no exceptions." However, there have been numerous changes to fire PRA methodology since 2005, including the following :

• The NRC staff has formally accepted methods during resolution of unreviewed analysis methods (UAMs) for fire PRAs, as well as NUREG/CR-6850 (as supplemented in September 2010) , or frequently asked question (FAQ) guidance developed for the National Fire Protection Association Standard (NFPA) 805, "Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants." FAQs that may be relevant for the fire PRA include:

FAQ 13-0004, (ADAMS Accession No. ML 13322A085) FAQ 13-0005, (ADAMS Accession No. ML 133198181) FAQ 13-0006, (ADAMS Accession No. ML 133318213) FAQ 14-0008, (ADAMS Accession No. ML 141908307)

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FAQ 14-0009, (ADAMS Accession No. ML 15119A176) FAQ 12-0064, (ADAMS Accession No. ML 12346A488) FAQ 08-0053. (ADAMS Accession No. ML 121440155) FAQ 08-0052, (ADAMS Accession No. ML092120501) FAQ 08-0050, (ADAMS Accession No. ML092190555) FAQ 08-0049, (ADAMS Accession No. ML092100274) FAQ 08-0045, (ADAMS Accession No. ML091240311) FAQ 08-0044, (ADAMS Accession No. ML092110516) FAQ 08-0043, (ADAMS Accession No. ML092120448) FAQ 08-0042, (ADAMS Accession No. ML092110537) FAQ 07-0035, (ADAMS Accession No. ML091620572) FAQ 07-0031 , (ADAMS Accession No. ML072840658) FAQ 06-0018, (ADAMS Accession No. ML072500273) FAQ 06-0017, (ADAMS Accession No. ML072500300) FAQ 06-0016, (ADAMS Accession No. ML072700475)

• The NRC has also issued a letter, "Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires" (ADAMS Accession No. ML 12171A583), June 21 , 2012, providing staff positions on 1) frequencies for cable fires initiated by welding and cutting , 2) clarifications for transient fires , 3) alignment factor for pump oil fires , 4) electrical cabinet fire treatment refinement details, and 5) the EPRI 1022993 report.

• The NRC has published NUREG/CR-7150, "Joint 6ssessment of Cable Damage and Quantification of _!;_ffects from Fire (JACQUE-FIRE) ," Volume 2, which is supported by a letter from the NRC to NEI, "Supplemental Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis" (ADAMS Accession Nos. ML 14086A165 and ML 14017A135).

• The NRC has published NUREG-2169, "Nuclear Power Plant Fire Ignition Frequency and Non-Suppression Probability Estimation Using the Updated Fire Events Database: United States Fire Event Experience Through 2009" (ADAMS Accession No. ML 15016A069).

• Guidance on the credit taken for very early warning fire detection system (VEWFDS) is available in NUREG-2180, "Determining the Effectiveness, Limitations, and Operator Response for Very Early Warning Fire Detection Systems in Nuclear Facilities, (DELORES-VEWFIRE)" (ADAMS Accession Nos. ML 16343A058). The guidance provided in FAQ 08-0046, "Closure of National Fire Protection Association 805 Frequently Asked Question 08-0046 Incipient Fire Detection Systems" (ADAMS Accession No. ML093220426) , has been retired.

Based on the PVNGS LAR, dated December XX, 2016, the calculated incremental conditional core damage probability (ICCDP) and incremental conditional large early release probability (ICLERP) are close to the RG 1.177 risk acceptance guidelines. However, the integration of NRG-accepted fire PRA methods and studies described above that are relevant to th is submittal

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could potentially result in an exceedance of the risk acceptance guidelines. For example, previous risk-informed LARs have shown that integration of NRC approved methods can lead to a calculated risk increase of up to approximately 3 in some cases. Therefore, in accordance with Section 2.5.5 of RG 1.17 4, additional analysis is necessary to ensure that contributions from this influence would not change the conclusions of the LAR. The NRC staff requests the licensee address one of the following :

(a) Provide a detailed justification for why the integration of NRG-accepted fire PRA methods and studies in the fire PRA would not change the conclusions of the LAR. As part of this justification, identify the fire PRA methodologies used in the fire PRA that have not been formally accepted by the NRC staff. For these methodologies, provide technical justification for their use and evaluate the significance of their use on the results of this submittal.

(b) Alternatively, demonstrate through a sensitivity study of the proposed TS change, which credits the compensatory measures, that the risk results (i.e ., total ICCDP and total ICLERP) meet the risk acceptance guidelines by a large margin. This sensitivity study may take into consideration, as necessary, credit for compensatory measures such as deployment of the portable AC diesel generators and the diesel-driven FLEX steam generator makeup pump, provided that they are modeled in a way that is consistent with RGs, 1.17 4, 1.177, and 1.200. Provide the associated risk results (i.e. , those results in Attachments 2 and 3 of Enclosure 2 to the LAR) and a discussion of how the compensatory measures were credited in the PRA models, including :

• Discuss the conservatisms in the analysis.

• Discuss which accident scenarios were credited for the compensatory measures.

• Explain how the failure rates/probabilities of hardware failures (e.g., random failures, unavailability due to testing and maintenance) associated with setup and operation were estimated.

• Explain how the timelines for operator actions were established. Describe the cues or indications operators will use to initiate use of credited FLEX equipment and how the time available and time required to complete operator actions were estimated.

R. Bement - 3 -

If you have any questions, please contact me at (301) 415-1564 or via e-mail at [email protected].

Docket No. STN 50-530

Enclosures: 1 . List of Teleconference Participants 2. APS Teleconference Slides

Sincerely,

/RA/ Margaret Watford for

Siva P. Lingam, Project Manager Plant Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation

3. NRC Talking Points (Potential Technical Issues)

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