ML101540033
| ML101540033 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 06/03/2010 |
| From: | FirstEnergy Nuclear Operating Co |
| To: | NRC/RGN-III |
| References | |
| Download: ML101540033 (47) | |
Text
FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Nuclear Regulatory Commission Public Meeting June 3, 2010
Agenda Opening Comments & Desired Outcomes..Barry Allen Site Vice President Cycle 16 Performance & 16th Refueling Outage Work ScopeClark Price Director, Site Performance Improvement 16th Refueling Outage Inspections..Vito Kaminskas Director, Site Engineering Results of the Root Cause Investigation...Scott Plymale Manager, Plant Engineering CRDM Nozzle Penetration Restoration.................................................Ken Byrd Basis for Return of the Reactor Vessel Head to Service Manager, Design Engineering Basis of Operation for Cycle 17 Remaining Activities Required for Restart...Brian Boles Director, Site Operations Closing CommentsBarry Allen 2
Desired Outcomes
Demonstrate Davis-Besses commitment to Safe and Reliable Operations
Recap Cycle 16 and 16 RFO Performance
Present the inspection results and discovery of the degraded Reactor Vessel Control Rod Drive Mechanism (CRDM) Nozzle Penetrations
Present the results of our Root Cause Investigation
Describe Nozzle Modifications completed to fully restore Reactor Vessel Head Integrity
Present our basis for the return of the Reactor Vessel Head to service
Demonstrate FENOCs overriding priority and commitment to ensure the safe and reliable operation of Davis-Besse 3
FirstEnergy Nuclear Operating Company Davis-Besse Nuclear Power Station Nuclear Regulatory Commission Public Meeting June 3, 2010
Agenda Opening Comments & Desired Outcomes..Barry Allen Site Vice President Cycle 16 Performance & 16th Refueling Outage Work ScopeClark Price Director, Site Performance Improvement 16th Refueling Outage Inspections..Vito Kaminskas Director, Site Engineering Results of the Root Cause Investigation...Scott Plymale Manager, Plant Engineering CRDM Nozzle Penetration Restoration.................................................Ken Byrd Basis for Return of the Reactor Vessel Head to Service Manager, Design Engineering Basis of Operation for Cycle 17 Remaining Activities Required for Restart...Brian Boles Director, Site Operations Closing CommentsBarry Allen 2
Desired Outcomes
Demonstrate Davis-Besses commitment to Safe and Reliable Operations
Recap Cycle 16 and 16 RFO Performance
Present the inspection results and discovery of the degraded Reactor Vessel Control Rod Drive Mechanism (CRDM) Nozzle Penetrations
Present the results of our Root Cause Investigation
Describe Nozzle Modifications completed to fully restore Reactor Vessel Head Integrity
Present our basis for the return of the Reactor Vessel Head to service
Demonstrate FENOCs overriding priority and commitment to ensure the safe and reliable operation of Davis-Besse 3
Cycle 16 Performance and the 16th Refueling Outage Work Scope Safe Plant Operations Clark Price - Director, Site Performance Improvement Reliable Cost-Effective Plant Operations Plant Operations 4
Plant Performance
The Station operated safely and reliably for Cycle 16
- Implemented Improved Standard Technical Specifications
- Conducted a successful NRC/FEMA Evaluated Exercise
- Implemented improved Emergency Action Levels
- Worked over 10.7 million personhours without a lost time accident
- Pressurizer Valve Code Safety Outage to improve station reliability
- Industry Top Quartile On-line Operation Dose Exposure
- Industry Top Decile Chemistry Effectiveness Index
- Capability Factor - 92.72%
- Forced Loss Rate - 0.65%
5
16RFO Major Accomplishments Successfully Completed:
All planned Alloy 600 mitigation activities
- Weld overlays on all 4 Reactor Coolant Pump (RCP) suction and discharge lines
- Weld overlays on both Core Flood Line Nozzles
- Weld Overlays on remaining Cold Leg Drains
Fuel Inspections Fuel Handling Bridge Steam Generator Inspections
Emergency Core Cooling Pump Motor Replacements Optimized weld overlay
Aux Feed Water Auto Suction design utilized on the four swap to Service Water 28-inch RCP cold leg
In-Service Inspections discharge nozzle welds 6
16RFO Major Accomplishments Successfully completed maintenance and modifications to improve performance:
Emergency Diesel Generator Modification Lifting of the High Pressure Turbine Rotor for Inspections
Underground Cable Replacements
Pressurizer Code Safety Valve Replacements
Replacement of Plant Process Computer Multiplexers
High Pressure Turbine Rotor Inspections
Main Feed Pump Turbine Blade Replacement
Main Condenser Inspections and Maintenance Workers disassemble a Main Feed Pump Turbine 7
16RFO Major Accomplishments
Davis-Besse is expected to start up with:
- No Open Prompt Operability Determinations
- No Open Operational Decision Making Issues
- No Open Operator Work Arounds
- No Open Outage Corrective Maintenance Orders Control Room 8
16th Refueling Outage Inspections Safe Plant Operations Vito Kaminskas - Director, Site Engineering Reliable Cost-Effective Plant Operations Plant Operations 9
In-Service Inspection Scope
Over 280 Manual Examinations
#1 Steam Generator Auxiliary Feedwater Header
Pre-Mitigation & Pre-Service Inspections for Alloy 600 Weld Overlays
Reactor Vessel Upper Core Barrel Bolts
Reactor Vessel Bottom Penetrations
Reactor Vessel Closure Head Containment Building 10
Reactor Vessel Closure Head
Installed unused head of similar design from the CRDM Nozzles cancelled Midland plant in 2003
- Nozzles constructed of Alloy 600
- Pre-service inspection performed
- Interim replacement until new reactor vessel head to be installed Reactor Vessel Closure Head in 2014
Bare metal visual inspections performed:
- 2003
- 2005
- 2006
- 2008 Reactor Vessel Closure Head
- No boron leakage 11
Reactor Vessel Head Inspections Control Rod Drive
Planned 16RFO Scope Mechanism
- Axial and Circumferential Ultrasonic Inspections of all 69 Control Rod Drive Mechanism Penetration Nozzles Insulation
- Bare Metal Visual Inspection of the Reactor Vessel (RV) Head
Supplemental 16RFO Scope
- Additional Inspections of RV Alloy 600 Nozzle Head Reactor Vessel Head
- Performed inspection of J-Groove welds by either Liquid Dye Penetrant Testing (PT) or J-Groove Weld Eddy Current Testing (ET)
Control Rod Drive Mechanism Nozzle 12
Reactor Vessel Head CRDM Nozzle Penetration Inspections
Inspection Results and Reporting
- March 12 - Identified first indications in Control Rod Drive Mechanism Penetration Nozzles during Axial Ultrasonic Testing (UT) scans
- Initiated 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report to the NRC
- March 13 - Began Bare Metal Visual examinations and discovered visible boron deposits on the Reactor Vessel Head
- Updated 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> report to the NRC 13
Reactor Vessel Head CRDM Nozzle Penetration Inspections 14
Reactor Vessel Head CRDM Nozzle Penetration Inspections 15
Reactor Vessel Head Inspection Results
Reactor Vessel Head has 69 Control Rod Drive Mechanism Penetration Nozzles
- 24 Nozzles Identified for Modification
- 12 from Ultrasonic Testing (UT) of Nozzles
- 4 from Liquid Dye Penetrant Testing (PT) of J-Groove Welds
- 8 from Eddy Current Testing (ET) of J-Groove Welds
- Remaining 45 Nozzles and J-Groove Welds are ASME Code acceptable through a combination of either Visual Examination (VT), UT, PT, or ET Examinations
- Required inspections identified boron residue on the reactor head and flaws within several CRDM nozzles
- Flaws were found before they resulted in an operational challenge
- Structural integrity of the RPV head was not compromised 16
Results of Root Cause Investigation Safe Plant Operations Scott Plymale - Manager, Plant Engineering Reliable Cost-Effective Plant Operations Plant Operations 17
Root Cause Team
Summary of Problem Statement
- 16RFO Non-Destructive examinations of the Reactor Vessel (RV) Head found twenty-four Control Rod Drive Mechanism nozzles and J-groove welds with flaws including one with active leakage onto the reactor pressure vessel head. The RV head had been in service three operating cycles.
Utilized experts from:
- FENOC
- EPRI
- EPRI Materials Reliability Program Personnel
- Westinghouse
- AREVA
- Structural Integrity Associates
- Dominion Engineering
- Nuclear Industry peers 18
Evaluating the Condition
Reviewed Reactor Vessel (RV) head pre-service inspection results
Reviewed RV head manufacturing records
Reviewed chemistry operating parameters for previous cycles
EPRI Independent Review of Ultrasonic Testing Results
Independent 3rd Party review of Bare Metal Visual Inspections conducted since 2003 19
Evaluating the Condition
Laboratory Boron Analysis
Head Vent Line Temperature Indication Calibrations
Head Temperature Model Analysis
Residual Weld Stress Calculations
Crack Growth Rate Modeling 20
Evaluating the Condition
Samples taken of welds and nozzle metal for laboratory analysis
- Indications cut out of Nozzles 4 and 10 for analysis
- Nozzle 4 and 10 ends removed for laboratory analysis Boat and Ring Samples 21
Evaluating the Condition
Laboratory Analysis work performed
- Indication Characterization
- Scanning Electron Microscope Surface Examinations
- Dye Penetrant Testing
- Micro-focus X-Ray inspections
- Material Property Testing
- Micro Hardness Measurements
- Tensile Strength
- Precision Analytical Chemistry Analysis
- Microstructure Analysis
- Carbide Distribution
- Metal Grain Size Nozzle 4 Boat Sample
- Other Analysis examined by Fluorescent PT
- Contamination in weld or nozzle material
- Weld Defects
- Micro-focus X-Ray inspections 22
Root Cause Results
What was the cause of the Davis-Besse Reactor Vessel Head Control Rod Drive Mechanism Nozzle Penetration indications after 3 operating cycles?
- Direct Cause - Primary Water Stress Corrosion Cracking (PWSCC)
3 Conditions needed for Primary Water Stress Corrosion Cracking:
- Susceptible Material
- Environment
- Tensile Stress 23
Root Cause Results
Physical Conditions which may Contribute to PWSCC
- PWSCC Potential Contributors Assessed:
- Reactor Vessel (RV) Head and Tubing Manufacturing Methods Can Introduce Residual Stresses
- Weld Size
- Alloy 600 is a PWSCC Susceptible Material
- Analysis Results of PWSCC Contributors:
- Tube and head manufacturing methods were not unique to the existing Davis-Besse RV Head
- Welding, repair and inspection methods similar to other heads
- Weld sizes are within industry weld size distribution
- Similar material to other RV Heads 24
Root Cause Results
Conclusion from Data
- Laboratory data and assembly records suggest that the Davis-Besse Reactor Vessel (RV) Head is made of similar materials and assembled with similar manufacturing methods as other industry RV heads manufactured in the 1970s
- Unique cracking characteristics were not identified for the Davis-Besse RV Head Control Rod Drive Mechanism Nozzle Penetrations 25
Root Cause Results
Reactor Vessel Head Crack Growth Analysis
- EPRI Material Reliability Program (MRP) Developed Inspection Process
- Probabilistic Fracture Mechanics Analysis for Reactor Vessel (RV)
Head Nozzle Cracking
- Utilizes years of industry Alloy 600 crack and leakage data
- Forms the basis for ASME Code Case N-729-1 which is required to be used by 10CFR50.55(a), Codes and Standards
- Inspection interval and inspection technique set to maintain a low probability for Control Rod Ejection due to cracking or loss of structural integrity due to wastage 26
Root Cause Results
Analysis of Crack Initiation and Growth (cont.)
- Reactor Vessel (RV) Head Temperature is a key input into analysis
- BWOG analysis utilized Reactor Coolant System Thot as RV Head Temperature
- Davis-Besse RV Head temperature is warmer than Thot
- Small Continuous Vent Line flow has no significant impact on temperature Reactor Vessel 27
Root Cause Results
Analysis of Crack Initiation and Growth (cont.)
- 16RFO Inspection findings were compared to the Material Reliability Programs probabilistic crack analysis
- Statistical analysis shows that indications and leakage could be present as early as 16RFO
- Flaw Evaluation Analysis
- Deterministic analysis performed to determine the time it would take for postulated initial flaws to grow to the top of the J-groove weld and beyond, causing leakage
- Preliminary Model shows that the flaw is not expected to grow through wall 28
Root Cause Results
Conclusions from the Root Cause
- Direct Cause - Primary Water Stress Corrosion Cracking (PWSCC)
- No identified material or manufacturing issues that suggest the Davis-Besse head is unique
- Using the higher head temperature the MRP Statistical Model more clearly projects the probability for PWSCC crack initiation as found during 16RFO
- Crack Growth Models shows leakage is not expected over the next operating cycle
- Inspection requirements of 10CFR50.55(a) were effective at detecting early degradation of the Alloy 600 material 29
CRDM Nozzle Restoration and Basis for Return of Reactor Vessel Head to Safe Plant Operations Service Ken Byrd - Manager, Design Engineering Reliable Cost-Effective Plant Operations Plant Operations 30
CRDM Nozzle Modification Nozzle Before Modification Nozzle After Modification Control Rod Drive Modification moves the weld Mechanism Nozzle attachment to a region with no Primary Water Stress Corrosion Cracking (PWSCC) or flaws Alloy 600 Nozzle Reactor Vessel Head Reactor Vessel Head J-Groove Weld Original J-Groove Weld New Weld Location 31
CRDM Nozzle Modifications Control Rod Drive Mechanism Control Rod Drive Mechanism Penetration Nozzle Penetration Nozzle Before Repair After Repair 32
CRDM Nozzle Modifications
Roll Expansion
- Secures nozzle in place
Machining
- Machine nozzle to move attachment weld to a region with no PWSCC degradation The nozzle was rolled and the bottom portion containing the flaw machined away 33
CRDM Nozzle Modifications
Structural Weld
- New weld material resistant to PWSCC
Post Weld Non Destructive Examination
- Ultrasonic and Liquid Dye Penetrant Examinations performed on modified area
Abrasive Water Jet (AWJ) After machining the new nozzle Remediation was welded in this location
- AWJ induces compressive stress near the surface
- Process inhibits PWSCC initiation 34
CRDM Nozzle Modifications
Summary of Key Points
- Modification process moves the higher stressed weld attachment to a region with no PWSCC or flaws
- Method includes post repair remediation to provide extended life for repair
- Proven industry practice Control Rod Drive Mechanism Penetration Nozzle After Repair 35
Basis for Return of Reactor Vessel Head to Service
All nozzles with indications of degradation were modified
Remaining Nozzles and J-Groove Welds meet or exceed regulatory and ASME Code requirements through a combination of either Visual Examination (VT), UT, PT, or ET Examinations
Reactor Vessel (RV) Head is being returned to service in a fully restored condition
RV Head will support safe and reliable operation 36
Basis of Operation for Cycle 17 Safe Plant Operations Ken Byrd - Manager, Design Engineering Reliable Cost-Effective Plant Operations Plant Operations 37
Basis of Operation for Cycle 17
Davis-Besse plans to perform next Reactor Vessel Head inspection in early 2012
Inspection frequency is governed by ASME Boiler &
Pressure Vessel Code Case N-729-1
- Reinspection frequency is in accordance with Re-inspection Year (RIY) parameter
- Inspection requirements of 10CFR50.55(a) were effective at detecting early degradation of the Alloy 600 material
- RIY will be maintained less than the required 2.25
- RIY is a function of head temperature and accumulated Effective Full Power Years (EFPY) 38
Basis of Operation for Cycle 17
Effective Full Power Years (EFPY)
- Accumulated Effective Full Power Years (EFPY) will be controlled by cycle length
- Revised plan for Cycle 17 will provide a shorter cycle. EFPY accumulation will be almost 100 days less than the original plan.
39
Basis of Operation for Cycle 17
Head Temperature
- Temperature in the head region is dependent on the cycle core design
- The current core design for Cycle 17 will provide about a 3 degree reduction in average Head Temperature throughout Cycle 17 as compared to Cycles 14 through 16
- Reinspection Year (RIY) will be conservatively based on the highest core exit temperature in center region instead of an average temperature
- Reactor temperatures will be periodically monitored to support RIY calculation 40
Basis of Operation for Cycle 17
Cycle 17 Operational Changes that will minimize PWSCC Growth
- Shorter cycle length
- Lower head temperature through core design
Safety for Cycle 17 will be assured by compliance with 10CFR 50.55(a) and ASME Code Case requirements with several conservative factors providing additional safety assurance
- Re-inspection Year (RIY) determination based on highest temperature in center region of the core
- Liquid Dye Penetrant Testing (PT) or Eddy Current Testing (ET) examinations of all J-Groove welds during 16RFO exceed ASME Code Case requirements 41
Inspections for 17RFO
Code Required Inspections
- Examinations will meet the requirements as modified in 10CFR50.55(a), which endorses ASME Code Case N-729-1 for all 69 nozzles
Additional Inspections
- Surface examinations beyond the code requirements on the Control Rod Drive Mechanism Penetration J-groove welds 42
Remaining Activities Required for Restart Safe Plant Operations Brian Boles - Director, Site Operations Reliable Cost-Effective Plant Operations Plant Operations 43
Remaining Activities for Restart
Re-install the Control Rod Drive Mechanisms on the Reactor Vessel Head
Reload the Fuel into the Reactor Vessel
Reactor Vessel Head Cleaning and Visual Inspection
Install the Reactor Head on the Vessel
Perform remaining Refueling Outage Testing
Perform a Restart Readiness Plant Operator adjusts equipment in Assessment support of plant lineup activities for establishing Condenser vacuum
Return the Unit to safe and reliable Operation 44
Summary of Actions Going Forward
Prior to Restart
- Provide the NRC with the Re-Inspection Year (RIY) calculation for Cycle 17
Cycle 17
- Provide the NRC with the final Crack Growth Rate Analysis for the RPV Head
- Update the Root Cause Analysis Report and include the final Laboratory Results and Crack Growth Rate Analysis
- Provide the NRC with an untested Nozzle Ring Sample
- Provide the NRC with updated RIY calculations after every 6 months of operation
- Will not exceed 2.25 RIY
RFO 17
- Perform volumetric examinations of CRDM Nozzles per 10CFR50.55(a)
- Perform Surface Examinations of J-Groove welds left in service 45
Safe Plant Operations Closing Comments Barry Allen - Site Vice President Reliable Cost-Effective Plant Operations Plant Operations 46
Conclusion
Inspection requirements of 10CFR50.55(a) were effective at detecting early degradation of the Alloy 600 material
The cause of the Control Rod Drive Mechanism Nozzle cracking has been thoroughly investigated
We found no unique characteristics with the replacement head
We developed a new method to model Reactor Vessel Head temperatures
We restored the design margin for the Reactor Vessel Head with the modification and inspections performed
With this information, we have developed actions and operational parameters to ensure we have satisfactory operating margin going forward
Davis-Besse is ready to return to safe and reliable operation 47