ML15208A460

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Slides of Regulatory Conference for July 8, 2015
ML15208A460
Person / Time
Site: Pilgrim
Issue date: 07/08/2015
From:
Entergy Corp, Entergy Nuclear Generation Co
To:
NRC Region 1
Marjorie McLaughlin
References
Download: ML15208A460 (45)
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Text

PILGRIM NUCLEAR POWER STATION REGULATORY CONFERENCE Safety Relief Valve SRV-3A July 8, 2015

Entergy Representatives John Ventosa COO Northeast Fleet John Dent Site Vice President David Noyes Director, Regulatory & Performance Improvement John Macdonald Sr. Manager, g , Operations p

David Mannai Sr. Manager, Fleet Regulatory Assurance Thomas White Manager, Design Engineering Everett (Chip) Perkins Manager, Regulatory Assurance Clem Littleton Senior Lead Engineer Patrick Doody Senior Lead Engineer Dr. Donald Dube PRA Consultant, ERIN Engineering Dr. M. S. Kalsi President, Kalsi Engineering Zachary Leutwyler Senior Specialist, Kalsi Engineering 2

OPENING REMARKS John Dent Site Vice President Pilgrim Nuclear Power Station 3

Agenda 4

INTRODUCTION Dave Noyesy Director, Regulatory & Performance Improvement Pilgrim Nuclear Power Station 5

Introduction

  • We are extracting g all of the learnings g from this event o Detailed root cause evaluation is underway; identified gaps in:

Equipment performance monitoring Operability evaluation and corrective action rigor Post trip review performance o We are sharing those learnings with the fleet

  • Improvements I t have h been b made d ini addressing dd i conditions in the Corrective Action Program (CAP) since 2013 o Continued improvements in CAP implementation ongoing Pilgrim performance is different in 2015 versus 2013 6

3-Stage g Target g Rock Safety y Relief Valve (SRV) 4 Safety Relief Valves 7

Full Range of SRV Operation Risk-Significant Region 8

Pilgrim Approach and Methodology

  • Best available ((including g new)) information is derived using:

o NRCs bounding analysis o Plant-specific Plant specific PRA model o Actual in-plant experience in light of physical inspection of SRVs o Results from component inspections at National Technical Systems (NTS) (f (formerly l W Wyle l Labs)

L b ) (New Information) o Analytical Work (New Information)

  • Result after considering all plant-specific risk inputs is very low safety significance 9

PILGRIM SRV PERFORMANCE HISTORY John Macdonald Senior Manager, Operations g

Pilgrim Nuclear Power Station 10

Key Events May 2011 3-Stage Target Rock SRVs installed during RFO 18 o New model SRVs included all known corrective measures for main stem/piston loosening (identified in IN 2003-01 and GE SIL 646)

Feb 8, 2013 Cooldown following loss of offsite power and scram - winter storm Nemo o Attempted use of SRV-3A to reduce reactor pressure on three occasions; response from tailpipe acoustic monitor was not as expected (114 psig, 101 psig, and 98 psig)

May 2013 SRV-3A Pilot Assembly replaced due to Intergranular Stress Corrosion Cracking of Bellows identified in 10 CFR Part 21 during RFO 19; SRV-3A cycled Aug 22, 2013 SRV-3A manually opened twice (923 psig and 823 psig) and reclosed during plant shutdown Oct 14, 2013 SRV-3A manually opened once (1080 psig) and reclosed during plant shutdown Jan 27, 2015 Cooldown following partial loss of offsite power - winter storm Juno o Two attempts to use SRV-3C to reduce reactor pressure (220 psig and 260 psig)

F b 2015 Feb SRV 3A and SRV-3A d SRV SRV-3C3C replaced l d prior i tto restart t t ffrom winter i t storm t JJuno fforced d outage t

Mar 16, 2015 Target Rock issues 10 CFR Part 21 interim report based on Pilgrim SRV performance May 2015 3-Stage SRVs replaced with 2-Stage model during RFO 20 11 11

Limited Range of Known Impact

  • Extent of Condition Risk-Significant Region 12

Key Events and Timeline

==

Conclusions:==

  • At all times maintained multiple methods of reactor pressure control
  • At all times had at least 2 SRVs available for pressure relief
  • 2 SRV SRVs (SRV-3B (SRV 3B & -3D) 3D) opened d and d closed l d reliably li bl on multiple lti l demands when called upon across the entire pressure range
  • All SRVs consistently opened at high pressure
  • SRV-3A opened 4 times at high pressure after failing to open at low pressure during winter storm Nemo - 2013
  • All SRVs S consistently reclosed under all conditions
  • SRV-3B & -3D responded as expected more than 100 times Pressure control via SRVs available at all times across the entire pressure range 13 13

DELTA CORE DAMAGE FREQUENCY DIFFERENCES Dave Noyesy Director, Regulatory & Performance Improvement Pilgrim Nuclear Power Station 14

What is different?

  • NRC Staff conclusions driving the low-to-moderate safety significance o Common Cause Factor (CCF) consistently applies to failure-to-open (FTO) for all SRV ((assumed SRVs d 1 iin 6 FTO probability) b bilit )

o Failure-to-close (FTC) is credible (10X FTC probability) o SRV failure probabilities increase across the full reactor pressure vessel (RPV) pressure spectrum (10X FTO factor)

  • Pilgrim Pil i conclusionsl i driving d i i the th very low l safety f t significance i ifi o SRV-3B & -3D performance affirmed by inspection and analysis supports a lower CCF rate o Based on valve performance and maintenance practices, use of a nominal FTC probability b bilit iis appropriate i t o All SRVs exhibited functionality at high RPV pressure o Based on operating history, a 2X nominal FTO probability is conservative and is based on Bayesian methodology
  • Pilgrim Pil i will ill presentt the th inspection i ti and d ttesting ti resultslt andd th the iimpactt thi this has on the risk assessment New information provides plant-specific best information 15

SRV INSPECTION / ANALYSIS RESULTS Thomas White Manager, Design Engineering Patrick Doody Senior Lead Engineer Pilgrim Nuclear Power Station 16

SRV 3A & -3C SRV-3A 3C

  • SRV-3A ((2013)) & SRV-3C ((2015)) failed-to-open p on demand at relatively low reactor pressure
  • Both valves removed from service in February 2015 and replaced l d
  • Comparison of valves:

o Similar conditions found on both SRV-3A SRV 3A & SRV SRV-3C3C o Physical inspection of SRV-3B (SN8) & SRV-3D (SN7) revealed only minor degradation (wear) that would not have impacted valve operation SRV-3A SRV 3A & -3C3C physical conditions distinctly different from SRV-3B & -3D 17

3 Stage Target Rock SRV Testing 3-Stage

  • Testing of all 4 SRVs at NTS o All valves were removed during RFO-20 (April 2015) o SRV-3A & -3C with approximately 60 days in-service o SRV-3B & -3D with approximately 2-years in-service o As-found main stage test at 50 psig instead of safety mode setpoint o Successful lifts without high pressure pre-conditioning
  • Valves were disassembled and inspected o Dimensional checks on individual parts o Observed by Entergy, Target Rock, Lucius Pitkin (LPI), and NRC
  • Inspection p and test results support pp Pilgrim g p position on reliable SRV performance for the 4 SRVs, particularly SRV-3B & -3D SRV-3B & -3D maintained full functionality y 18

Target Rock Part 21

  • New 3-Stage g SRV Model 0867F has unique q combination of internal features (used in Pilgrim 2011 - 2015) o Largest available throat diameter o Largest g total steam drive flow to p piston o Highest opening speed o Cause Analysis Disc/piston p experiences p largest g test stand impact p force resulting g in piston becoming de-torqued and de-shouldered Vibration and/or pressure pulsations cause detrimental wear on some steam supply systems o Design changes to be determined by analysis and testing 10 CFR Part 21 analysis is consistent with Pilgrim Pilgrims s analysis and experience 19

Main Stage Internal Parts Example of Ring/Guide Wear 20

SRV Analysis Model

  • Calculates the Piston Delta Pressure (DP) required to climb up the ramp
  • Required Piston DP increases as angle increases Groove
  • Required Piston DP Depth increases as friction coefficient ffi i t increases i

Ramp Angle R

Required i d Piston Pi t DP is i dependent d d t on ramp angle l and d friction f i ti coefficient 21 21

SRV Assessment of Stroke OPEN SRV-3B - New information

  • Inspection p revealed no measureable wear and gguide damage

~10° 15°

  • Inspection revealed shallow wear groove and ramp

.0025 depth angle

  • Shallow ramp easily climbed by piston ring

.020 SRV-3A and -3C 56° depth

  • Both did not open at low pressures (98 to 260 psig), but both opened at higher system pressures New Information for SRV-3B & -3D
  • Both opened at NTS at 50 psig; meeting the valve design specification
  • Internal inspections explain why the valves operated per the design SRV-3B and -3D operated at all pressures - SRV-3A and -3C had reduced operating range due to ramp angle 22 22

Inspection p Results Condition Affected Part Demand Cycles 50+ 50+ 8 (2 PWTs) 3 (1 PWT)

Grooves Worn in NO Minor wear from Deep Ring Deep Ring Guide piston contact Grooves Grooves Max Depth Max Depth Max Depth 0.0025 inches 0.020 inches 0.018 inches Max Ramp Angle Not Applicable ~10° ~39° ~56° Gap from Piston to 3-4 3 4 mils 3-4 3 4 mils Large Gap Large Gap Disc Shoulder 13-16 mils Not Measured Piston OK Shoulder not Wear on OD Wear on OD and Threads Galled Threads Galled Rings Normal Wear Normal Wear Fretting Wear Fretting Wear Disc / Stem Shoulder not Shoulder not Threads Galled Threads Galled Disc length short Disc length short Disc machined off Severe thread Pitch damage pitch damage SRV-3A & -3C physical conditions distinctly different from SRV-3B & -3D 23

SRV 3B Guide SRV-3B No Guide Damage g No Grooves Guide is re-useable Ring shadows on Guide surface are causeddbby h heatup t

SRV-3B opening and closing unaffected - No wear 24

SRV 3D Guide SRV-3D Magnified Impression of Localized Wear in Guide on SRV-3D Max Localized Depth 0.0025 Inches Minor Wear Location 10° OPEN SRV-3D opening and closing unaffected by very minor wear 25

SRV 3A Guide SRV-3A Magnified Impression from top and bottom ring grooves located at 1200 position Max Depth 0.020 inches 8 times deeper than SRV-3D Sh ll Shallower ramp anglel th than SRV-3C 0.290 inches wide 0.240 inches wide

~39° OPEN Top Groove OPEN Bottom Groove SRV-3B & -3D comparatively little or no Guide wear 26

SRV 3C Guide SRV-3C Top Groove Magnified Impression from 0 003 inches 0.003 Bottom Ring Groove located at 0530 position Max Depth 0.018 inches 7 times deeper than SRV-3D St Steeper ramp angle l th than SRV 3A SRV-3A 0.215 inches wide

~56° OPEN Bottom B tt Groove G

SRV-3B & -3D comparatively little or no Guide wear 27

Kalsi Engineering Results

  • Valve modeling to assess opening/closing performance (using as-found conditions)
  • Force balance analysis takes into account:

o Differential pressures o Coefficient of friction between internal parts o Spring force o Parts geometry Threads Stem and piston shoulders Effective ramp angle of grooves

  • Third-Party Thi d P t ReviewR i P Performed f db by MPR and d LPI Comprehensive modelling of all critical parameters affecting SRV operation 28

Kalsi Engineering Conclusions Opening p g Analysis y

  • SRV-3A & -3C minimum opening pressure estimated by analysis
  • Minimum opening pressures:

SRV/Opening Pressure Effective Ramp Angle

  • 3A: Est.

Est >200 psig ~39°

  • 3C: Est. >300-400 psig ~56°
  • 3B: Actual <50 psig None
  • 3D: Actual <50 psig ~10° SRV-3A SRV 3A & -3C 3C - Reduced ability at low pressure SRV-3B & -3D - No loss of capability 29

Kalsi Engineering Conclusions Closing g Analysis y

  • Based on design, approximately two times more closing force available than opening force
  • Ability to close was not compromised
  • As-found conditions identified in Hatch and Browns Ferry 2-stage SRVs that failed to close in-service were evaluated (Ref IN 2003 (Ref. 2003-01, 01 GE SIL 646 646, Browns Ferry Licensee Event Report, Hatch and Browns Ferry Root Cause Analysis)
  • Pilgrim piston tilting limited by existing thread and shoulder clearances on all SRVs Cl i capability Closing bilit nott compromised i d for f any valvel 30

New Information - Summary/Conclusions

  • Additional Inspections, Testing, and Modelling/Analyses Support Minimal Loss of SRV Functional Capability o CCF Significant physical differences between SRV-3A/-3C and SRV-3B/-3D This explains the performance differences at low pressure SRV 3A & -3C SRV-3A 3C functioned at high pressure Differences should be reflected in the CCF o FTC:

New data demonstrates failure mode baseline probability is unaffected All SRVs demonstrated closing capability at all pressures o FTO:

Testing and inspections of SRV-3B & -3D demonstrates failure mode baseline probability p y is unaffected Both SRVs demonstrated opening capability at all pressures New information provides best available information for risk assessments 31

PILGRIM PERSPECTIVE ON NRC STAFF BOUNDING ANALYSIS Dave Noyesy Director, Regulatory & Performance Improvement Pilgrim Nuclear Power Station 32

PRA Keyy Assumptions p Comparison p and Reconciliation - MLOCA Run Title PNPS (CDF) NRC (CDF) Factor Comment Difference Delta Core Damage 1.83E-08 1.34E-06 ~ 100 Result: 2 orders of magnitude Frequency (CDF) difference MLOCA Base Case 2.08E-07 1.19E-07 ~1 Similar Success criteria of 1 out of 4 Close agreement SRV required to open MLOCA Conditional 2.27E-07 1.46E-06 ~ 10 NRC CCF = 1.74E-01 Large difference in PNPS CCF = 1.39E-03 CCF SPAR model kept CCF group of 4 valves PNPS separated CCF group into 2 valves (SRV-3B & -3D)

Major difference is how CCF is treated 33

Pilgrim Perspective - MLOCA Run Title PNPS ((CDF)) NRC ((CDF)) Factor Comment Difference Delta CDF 1.83E-08 1.34E-06 ~ 100 NRC CCF = 1.74E-01 Large difference PNPS CCF = 1.39E-03 in CCF Delta CDF (New) 1.30E-07 ~ 10 PNPS CCF = 1.0E-02

  • Not credible to postulate a 1-in-6 chance that all SRV SRVs s will fail to open o Valves SRV-3B & -3D opened and closed more than 100 times o Disassembly of valves SRV-3B & -3D revealed minimal physical degradation which would not have prevented operation of valves over entire pressure range o Given the actual performance of SRV-3B & -3D (operated ~ 100 times), inspection results, and analysis we have assigned a conditional CCF value of 1.0E-02 CCF iis overly l conservative ti withoutith t consideration of plant-specific information 34

CCF Sensitivity - MLOCA

  • SPAR CCF is 2 orders of magnitude greater than Pilgrims o NRC treats all 4 SRVs as a CCF group of 4; Application of SPAR CCF generates a 1.74E-01 value which assumes loss of function of 2 or more valves (upper bounding condition) o Initial Pilgrim PRA analysis treated SRV-3B & -3D as a CCF group of 2 b

based d on actual t l performance, f iinspection, ti andd analysis; l i thithis generated t da 1.39E-03 value (lower bounding condition) o Best-available information compels consideration of actual performance o Given actual p performance of SRV-3B & -3D ((operated p ~ 100 times), ),

inspection results and analysis we assigned conditional CCF value of 1.0E-02 o A CCF value of 1.0E-02 is conservative considering the performance of SRV-3B & -3D combined with functionality of SRV-3A & -3C at high pressure SRV-3A & -3C degraded performance only at pressures below risk-significant range o This yields a Pilgrim Delta CDF for all MLOCA events of approximately i t l 1.30E-07 1 30E 07 35

PRA Keyy Assumptions p Comparison p and Reconciliation - Non-MLOCA Run Title Pilgrim NRC Factor Comment (CDF) (CDF) Difference Delta CDF 3.60E-07 2.20E-06 ~ 10 2X FTO most appropriate treatment (2X FTO) (10X FTO & No evidence supporting a 10X FTC 10X FTC)

Non MLOCA 2.89E-06 2.08E-05 ~ 10 SPAR generally more conservative Base Conditional 3.25E-06 Conservative application of 2X FTO in accordance with data (2X FTO) contained in NUREG/CR NUREG/CR-7037 7037 and the methodology for Bayesian updating in NEI 99-02 Conditional See 2.30E-05 Overly conservative - no observed failure mechanism that (FTO 10X & Comment would tend to increase FTC rate; not realistic to use FTC with FTC 10X) 10X multiplier 2X FTO most appropriate and any FTC increase is not credible 36

Pilgrim Perspective - Non-MLOCA Non MLOCA Run Title Pilgrim (CDF) NRC (CDF) Factor Comment Difference Delta CDF 3.60E-07 2.20E-06 ~ 10 2X FTO most appropriate treatment (2X FTO) (10X FTO & No evidence supporting a 10X FTC 10X FTC)

  • Overly conservative to assume a 10X increase in probability of FTO o Applying a 10X FTO to all SRVs is overly conservative (Note that the resultant FTO probability of 2.77E-02 would predict ~ 3 instances of SRV FTO during event when 100 demands were experienced) o Conservative application of 2X FTO in accordance with data contained in NUREG/CR-7037 NUREG/CR 7037 and the methodology for Bayesian updating in NEI 99-02
  • Not credible to postulate a 10X increase in probability of FTC o RASP Handbook: The expected higher failure probability estimate can be assessed using engineering judgment through expert elicitation; In some cases, cases the estimate may be derived through prior operating experience of the component o No industry history of FTC for valves modified and maintained in accordance with IN and GE SIL o Kalsi Engineering review of disassembled SRVs affirmed no projected FTC above nominal probability Differences due to overly conservative FTO/FTC assumptions 37

Delta CDF Comparison Run Title Pilgrim (CDF) NRC (CDF) Factor Comment Difference Total Delta 4.90E-07 3.54E-06 ~10 Differences in application of best-available plant plant-specific specific information Delta CDF 1.30E-07 1.34E-06 ~10 CCF application MLOCA difference Delta CDF 3.60E-07 2.20E-06 ~10 FTO/FTC application Non MLOCA Differences due to overly conservative assumptions 38

OPERATIONAL RISK MITIGATING FACTORS John Macdonald Senior Manager, Operations Pilgrim Nuclear Power Station 39

Operational Risk Mitigating Factors

[HPCI], Reactor Core Isolation Cooling [RCIC],

Feedwater, Control Rod Drive, and Standby Liquid Control)

  • EOP-01, RPV Control, provides alternate depressurization systems: HPCI, RCIC, Main Steam line d i drains, R Reactor W Water Clean-Up Cl U in i lletdown d mode d
  • Our procedures and training direct operators to initiate emergency g y depressurization p when RPV level approaches pp top of active fuel versus at the minimum steam cooling reactor water level
  • Availability of diverse shutdown methods demonstrated 40

CONCLUSIONS Dave Noyesy Director, Regulatory & Performance Improvement Pilgrim Nuclear Power Station 41

Conclusions

  • Public health and safety always protected, with margin
  • Our initial conclusions on risk were based on empirical data regarding performance of the SRVs in the plant o We performed testing, detailed inspections, and expert analysis o These validated the experience of the SRVs in the plant
  • Updated plant-specific PRA evaluation provides best available determination of actual risk, based on:

o SRV-3B & -3D were distinctly different from SRV-3A & -3C o Two T SRVs SRV (SRV-3B (SRV 3B & -3D)3D) always l worked k d o Narrow range of scenarios of concern with operation for SRV-3A & -3C Functional at high pressure o Increased FTC probability is not a credible concern for Pilgrim Twice as much closing force Appropriate to remove the 10X factor o 2X FTO is appropriate

  • Result:

R lt Risk Ri k iis off very llow safetyf t significance i ifi 42

Appendix M Analysis - Summary Number Subject Pilgrim Position 4.2.1.1 Defense in depth Other mitigating strategies remained available; alternative pressure control and high pressure injection 4212 4.2.1.2 Reduction in safety margin Very small; SRV-3B SRV 3B & -3D 3D remained operable quantified at all times; Increased FTC probability of any valves not credible; 2X FTO is appropriate 4.2.1.3 Extent of PD impact on None other equipment 4.2.1.4 Degree of degradation 2 valves FTO at low pressure below the risk-significant region 4.2.1.5 Period of fault exposure 12 months time 4.2.1.6 Likelihood of success of Other mitigating g g strategies g remained available; licensees recovery actions alternative pressure control and high pressure injection 43

CLOSING COMMENTS John Dent Site Vice President Pilgrim Nuclear Power Station 44

QUESTIONS AND DISCUSSION 45

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