ML13308B332

Meeting Presentation for November 4, 2013 Public Meeting Regarding Bmi Leakage

ML13308B332
Person / Time
Site:	Palo Verde
Issue date:	11/04/2013
From:	Arizona Public Service Co
To:	Jennivine Rankin Plant Licensing Branch IV
Rankin J
References
TAC MF2871, TAC MF2872, TAC MF2873
Download: ML13308B332 (54)

Text

Palo Verde Unit 3 Bottom--Mounted Instrument (BMI)

Bottom Nozzle #3 Leak November 4, 2013

Palo Verde Participants/Attendees Dwight Mims Senior Vice President, President Regulatory & Oversight Jack Cadogan* Vice President, Nuclear Engineering Ron Barnes* Director, Nuclear Regulatory Affairs Ken House House* Director, Design Engineering Mike McLaughlin Director, Technical Support Mark Fallon Director (acting), Palo Verde Communications Tom Weber* Department Leader, Nuclear Regulatory Affairs Mike DiLorenzo Department Leader, Program Engineering Brian Cable Manager, Unit 3 Operations Doug Hansen* Senior Consulting Engineer, Program Engineering / Principal Level III Ed Fernandez* Senior Metallurgist, Engineering Programs / PWROG MSC Chairman Gene Montgomery Senior Engineer, Design Engineering (Mechanical NSSS)

Scott Bauer Regulatory Affairs Manager, STARS Dave Waskey k Manager, Welding ld andd Component Repair Design, Areva Doug Killian Technical Consultant, Areva Michael Lashley Associate, Structural Integrity Associates, Inc.

2 NOTE:

• identifies presenters

Agenda

• Discovery and Initial Response Ron Barnes

• NDE Action Plan and Results Doug Hansen

• Causal Evaluation Ed Fernandez

• Repair Plan Ken House

• Relief Request Tom Weber

• Closingg Comments Jack Cadogan g

3

Desired Meeting Outcomes

• Awareness/understanding of:

- Condition and current status

- Non-destructive examination (NDE) methodologygy and results

- Causal evaluation methodology, results to date

- Repair plan/technique

- Relief Request

• APS/Palo Verde understands NRC questions/concerns

/ related l d to this h issue 4

Discovery and Initial

Response

Ron Barnes Di Director, N l Nuclear R Regulatory l Aff Affairs i

Control Element Drive Mechanism Unit 3 Reactor BMI Nozzle N l #3 2010 2013 In-core Instrumentation Nozzles 6

BMI Nozzles

• Fixed in-core instrumentation

• 61 penetrations t ti J-groove weld

• 3.0 outside diameter

• 0 0.75 inside i id didiameter

• J-groove weld

• Alloy 600 nozzle material Reactor Vessel 7

Initial Response

• Event Notification (ENS) to NRC

• Response Teams established - tiered approach

- Non-destructive Examination (NDE) Team to characterize flaw

- Engineering/Repair Team

- Project Management Team

- Causal Evaluation Team

• Communication Plan implemented

- Included NRC and industry stakeholders

• Leveraged industry expertise/experience in all aspects of response 8

BMI Leak Identified Decision Inform NRC/ Industry Tree IInformational f ti l helium test Characterize flaw UT/ECT in tube Weld indication of bubble (confirmed) 1 Circumferential flaw in tube OR AND Consult industry 1 A Axial flaw in tube Perform Determine enhanced boat J-weld OR visual (VE) sample ECT Repair and remnant location analysis 4 4 Confirmed Not Confirmed No flaw detected 2 in tube Remove boat 10%

0% eexpansion pa s o sample 2

Expand 100% Go to axial AND Flaw analysis of expansion J-weld ECT Enhanced Visual Flaw analysis A of expansion Repair NOTES OR

1. For characterized flaws, J-weld and a Boat Sample will be 1 considered.

OR 2. If the EPRI demonstrations identify significant limitations, 2 then expansion beyond 10 CFR 50.55a will be re-evaluated.

Accept as is Repair as needed 3.

3 Flaw evaluation and expansion are applicable to pressure boundary areas only.

Accept as is Repair as needed 4 4.

4 A defined As d fi d by b inspection i ti procedure.

d 9

Non-Destructive Non-Examination (NDE)

Action Plan and Results Doug Hansen S i C Senior Consulting li E Engineer, i P Program E Engineering i i Principal Level III

Bare Metal Visual Examination (VE)

• Examinations performed per Code Case N722-1 l examined

• All 61 nozzles i d

• Nozzle #3 was the only one with leakage noted 11

Wastage Examination Nozzle #3 Results

• Phased Array Ultrasonic Testing

-CConducted d t d ffrom the th outside t id

- Adjacent to nozzle #3

- Focus on degradation (wastage) in the vessel shell at the nozzle bore

• Examination demonstration:

- South Texas Project (STP) mock-up and technique

- All mock-upp flaws were detected

• Examination results:

- No indication of wastage was detected 12

Helium Bubble Test 13

Video Here 14

J-Groove Weld Eddy C

Current t Testing T ti

• On the decision tree

• Technology not available to place coil at the helium area 15

Ultrasonic & Eddy Current Testing

• Similar to reactor vessel closure head examinations

• Conducted from the nozzle inside diameter g p

• Single probe with multiple p techniques q

• Techniques demonstrated at EPRI:

- TOFD (time of flight diffraction)

• Both axial and circumferential

- Additional techniques used:

• Inside diameter eddy current

• 45 degree shear-wave UT; looking down

• Zero degree UT; looking perpendicular to the surface 16

Ultrasonic Testing Demonstration (EPRI NDE Center)

• Mock-up design and f b i ti fabrication early l 2013 Flaw locations: inside and outside diameter Flaw orientations: axial, off angle, and circumferential

• Results: all flaws d t t d detected 17

Nozzle #3 TOFD Graphic 360 Degree slice Deepest 0.378 Zoomed in to show

~0.4 of nozzle of 1.125 total thickness Axial Indications Nozzle OD Weld Weld Area Area 18

BMI Nozzle #3 Bubbles visually observed at 42 degrees 4

1 2 3

Deepest:

eepest 00.378 3 8 Longest: 1.88 Overall width: 72º (1.87) 19

Conclusions/Summary

• BMI nozzle #3

- Only visual leaking nozzle

- Helium validated leak location

- No inside diameter indications detected

- Multiple axial ultrasonic indications

• APS Principal Level III

• Two WesDyne Level IIIs

• WesDyne Chief Engineer

• EPRI independent reviewer

• All other nozzles (60) p

- No unacceptable indications 20

Causal Evaluation Ed Fernandez S i M Senior Metallurgist, ll i E i Engineering i P Programs PWROG MSC Chairman

Causal Evaluation Team Composition

• Station core team

- Consisting of station personnel and industry peers including Structural Integrity, Westinghouse and AREVA

• Industry groups

- PWROG M Materials t i l S Subcommittee b itt (MSC)

- EPRI Material Reliability Program (MRP)

- INPO 22

Causal Evaluation Process

• Failure modes and effects analysis (FMEA)

- Palo Verde Corrective Action Program (CAP)

Cause Analysis Manual

- Reviewed and informed by EPRI MRP-206 Inspection and Evaluation Guidelines for Reactor Vessel Bottom Bottom-Mounted Mounted Nozzles Nozzles along with Operating Experience lessons learned

- Developed a summary of potential causal factors based on input from EPRI, Westinghouse, AREVA, STP and Structural Integrity Associates 23

Causal Evaluation FMEA Nozzle Leak Primary Water Environmental Axial-Radial RPV Surface Environment Fatigue Circ-Axial Weld ID/OD Axial ID/OD Circ Weld or Butter Breaking Lack Primary Water or Butter Flaw Flow in Tube Flaw in Tube Flaw of Fusion Environment Off Water Natural Circulation Chemistry Inside the Nozzle C diti Conditions iin P Pastt T b Tube Startup Water Operating Chemistry T-hot Conditions Volumetric Surface Defects in Functional Tube Material Defects in Nozzle Nozzle Tube from Alloy 600 Heat Weld Repairs Repairs p

Tube from Mat Matll Processing Treatment Processing Fabrication Previous Chemistry Mechanical Excursions/

Vibration Contamination Weld hot Cracking Nozzle Nozzle Roll and Other Weld Lack of Weld Grinding of Surface Straightening Straightening Fabrication Fusion Areas from Nozzle Tube or Cold Working Operational p Impacts p

Contaminants After During Mat Matll R i IIntrusions Resin t i Defects/

f / Fabrication b Weldld of Rx Work Installation Processing Contaminants Material Stress Environment 24

Probable Cause

• Probable cause

- Crack initiation was likely due to a weld defect exposed to primary water environment, environment resulting in primary water stress corrosion cracking (PWSCC)

• Probable causal factors

- Material

• Alloy 600

• Near surface weld defect

- Stress

• Weld residual stress

• Weld repairs and grinding

- Environment

• Primary water

• Temperature

• Operating environment 25

Causal Evaluation Additional Analyses

• Collection of boat sample

• Sample content

- RCS leak entrance point

- Weld defect

- Axial crack

- Area of high reflectivity

- Unaffected Alloy 600 and 182 material t i l 26

BMI Nozzle #3 Bubbles visually observed at 42 degrees 4

1 2 3

Deepest:

eepest 00.378 3 8 Longest: 1.88 Overall width: 72º (1.87) 27

Boat Sample Dimensions 28

Causal Evaluation Boat Sample

• Metallurgical analysis and test plan

- Visual inspections

- Liquid penetrant (PT)

- X-ray radiography

- High-resolution replication

- Scanning Electron Microscopy (SEM)

- Energy Dispersive Spectroscopy (EDS)

- M t ll Metallography h

29

Conclusions/Summary

• The UT results are characteristic of PWSCC

• The initiation likely occurred at a weld defect which was exposed to the primary water environment resulting in PWSCC

• Boat sample removal and metallurgical analysis and testing are planned 30

Repair Plan Ken House Director Design Engineering

Director,

Repair Options Considered

• Half-nozzle repair selected

- Code compliant repair

- Proven technology extensive industry p

experience

- ALARA

- Permanent repair

• Other options considered:

E t

- Externall mechanical h i l plug l

- Inner diameter temper bead (IDTB) repair 32

Half-Nozzle Repair Alloy 600 182 Filler Temper p Bead Pad 52M Filler Alloy 690 Temper Bead Pad 33

Extensive Mock-Up Preparation 34

Bore Machining Mock-Up 35

Temper Bead Pad Mock-Up 36

Weld Pad 37

Repair Timeline

• Commenced work: 10/27/2013 T B d Pad

• Temper-Bead P d Complete:

C l t 11/02/2013

• Half-Nozzle Complete: 11/07/2013 38

Repair Analyses Corrosion Assessment ASME Analyses 39

ASME Section III Class 1 Analysis

• Stress and fatigue analysis consistent with original reactor vessel design specification requirements

- Stress loads (normal/upset/emergency/faulted conditions)

- Fatigue loads (thermal transient) 40

Corrosion Assessment

• Small gap between original Alloy 600 nozzle and new Alloy 690 nozzle will exist following repair

• Low-alloy steel corrosion rate due to interaction with primary coolant in operating reactors has proven to be extremely small WCAP 15973 documents method for evaluating

• WCAP-15973 corrosion of low alloy steel following half-nozzle repairs

• Palo Verde plant-specific analyses are in progress, which follow the WCAP methodology 41

Remnant Analysis Remnant Analysis 42

Remnant Analysis for Relief Request

• Finite Element Analysis previously done for a representative Palo Verde BMI nozzle configuration

- Pressure, thermal and residual stresses

• AREVA performed a fracture mechanics analysis on a postulated maximum remnant fl flaw including i l di conservative ti fl flaw extension t i ffor crack growth during one operating cycle

• Analysis demonstrates the weld flaw maintains structural integrity and is acceptable 43

Summary/Conclusions

• Palo Verde implementing half-nozzle repair on Unit 33, nozzle #3 Code compliant repair Permanent repair backed by extensive industry experience Repair bounds probable cause(s)

• The remnant analyses support the relief request 44

Relief Request Tom Weber Department Leader, Leader Nuclear Regulatory Affairs

Palo Verde Applicable ASME Codes

• Design Code for reactor vessel Edition Winter 1973 Addenda

- ASME III 1971 Edition,

• Construction Code Edition Winter 1975 Addenda

- ASME III 1974 Edition,

• Repairs/Replacements Edition 2003 Addenda

- ASME XI 2001 Edition,

• Palo Verde 3rd ISI Interval 7-17-2018

- Unit 1 thru 7 17 2018

- Unit 2 thru 3-17-2017

- Unit 3 thru 1-10-2018 46

Relief from ASME Code

• Two separate relief requests

- Restart analysis

• Duration of one operating cycle o g te

- Long-term evaluation e a uat o

• Detailed analysis including fatigue crack growth

• Operation beyond next operating cycle 47

Relief from ASME Code

• Removal of defects 4421 Defects shall be removed or

- IWA 4421, mitigated in accordance

- IWA 4422.1a,, A defect is considered removed when it as been reduced to an acceptable size 4422 1b Alternately,

- IWA 4422.1b, Alt t l the th defect d f t removall area and any remaining portion of the defect may be evaluated and the component accepted in accordance with the appropriate flaw evaluation provisions of Section XI 48

Relief from ASME Code

• Characterization of flaws in J-groove weld 3100(a) Evaluation

- IWA 3100(a), Evaluation shall be made of flaws detected during an inservice examination as required by IWB-3000 for Class 1 pressure retaining components components

- IWA 3300(b), Flaws shall be characterized in accordance with IWA-3310 through IWA-3390...

- IWA 3420, Each detected flaw or group of flaws shall be characterized by the rules of IWA-3300 to establish ab the dimensions d o ofo the flaws.

a These dimensions shall be used in conjunction with the acceptance standards of IWB-3500.

49

Relief from ASME Code

• Successive examinations 2420(b) If a component is accepted for

- IWB 2420(b),

continued service in accordance with IWB-3132.3 or IWB-3142.4, the areas containing flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the inspection program of IWB-2400 50

Proposed Alternative per 10 CFR 5050.55a(a)(3)(i) 55a(a)(3)(i)

• Implement design repair on outside surface of Unit 3 reactor vessel

• Relocate pressure-retaining weld

• Analyze a postulated maximum flaw in remnant J-groove weld

• Duration of relief to be one operating cycle 51

Basis for Relief Request

• ASME Code compliant half-nozzle repair N ttechnology

• No h l available il bl ffor characterization of flaws in J-groove weld

• Analysis of postulated maximum flaw demonstrates remnant flaw remains acceptable for one operating cycle 52

Relief Request Summary

• Proposed alternative provides an acceptable level of quality and safety for the next operating cycle

• Separate ASME Relief Request to address successive ASME Code examinations and operation p beyond y the next operating p g cycle y

53

Closing Comments Jack Cadogan Vice President President, Nuclear Engineering