ML12171A525
| ML12171A525 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 01/31/2012 |
| From: | Electric Power Research Institute |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| Shared Package | |
| ML12171A508 | List: |
| References | |
| RAS 22624, 50-247-LR, 50-286-LR, ASLBP 07-858-03-LR-BD01 | |
| Download: ML12171A525 (9) | |
Text
NYS000393 Submitted: June 19, 2012 EXCERPT nuclear sector roadmaps January 2012
TABLE OF CONTENTS materials degradation and aging .................................................. 7 in use BWR and PWR Irradiated Materials Testing and Degradation Models...............................................9 Steam Generator Foreign Object Management.....................................................................................12 Steam Generator Life (Deposit) Management ......................................................................................15 Management of Jet Pump Flow-Induced Vibration .............................................................................. 18 Ensuring Reactor Pressure Vessel Integrity through Eighty Years of Operation......................................21 PWR Reactor Internals Aging Management.........................................................................................24 Welding of Irradiated Materials for PWR and BWR Internals ............................................................27 Alloy 52 Nickel-Base Filler Metal Weldability Solution .......................................................................30 High Chromium Weld Metal Development and Cracking Mechanisms ................................................33 Aging Management of Alloy 600 and Alloy 82/182 in the Steam Generator Channel Head Assembly....36 Mitigation of Boiling Water Reactor Materials Degradation................................................................39 Fatigue Management in Boiling and Pressurized Water Reactor Coolant Systems.................................42 Pipe Rupture Probability Reassessment (XLPR) ...................................................................................45 Primary Water Stress Corrosion Cracking Characterization of Alloy 690 and Weld Metals..................48 Advance Welding Process Development in the Nuclear Power Industry ................................................51 Incorporation of Fabrication, Repair and Joining Technologies in Nuclear Codes and Standards..........54 draft Steam Generator Tube Integrity Assessment.........................................................................................57 fuel reliability ............................................................................... 61 in use PWR Grid-to-Rod Fretting Fuel Failure Prevention ............................................................................63 Mitigation of Fuel Failures Caused by Foreign Material ......................................................................67 Mitigation of PWR Fuel Failures Caused by Crud and Corrosion......................................................... 70 Mitigation of BWR Fuel Failures Caused by Corrosion and Crud ........................................................73 This roadmap book is available at http://www.cvent.com/d/6cq8dc/4K.
Table of Contents 3 January 2012
Loss of Coolant Accident Regulation.....................................................................................................77 Zero Fuel Issues Strategic Matrix.........................................................................................................80 used fuel and high - level waste management ................................... 83 in use Advanced Fuel Cycles...........................................................................................................................85 Used Fuel Extended Storage ................................................................................................................88 High Burn-Up Fuel Transportation.....................................................................................................91 nondestructive evaluation ............................................................ 95 in use Assessment of Cast Stainless Steel.........................................................................................................97 Concrete Structure Characterization and Nondestructive Evaluation..................................................100 Remote Visual Examination ................................................................................................................103 Nondestructive Evaluation Modeling and Simulation Center ..............................................................106 Nondestructive Evaluation for New Plants ..........................................................................................109 Nondestructive Evaluation Reliability.................................................................................................112 equipment reliability ...................................................................... 115 in use Buried Pipe Integrity...........................................................................................................................117 Cable Aging Management....................................................................................................................120 Digital Instrumentation & Control Implementation (New/Existing Plants)..........................................123 Advanced Preventive Maintenance Technologies...................................................................................126 Guidance for Maintaining Circuit Cards, Relays, and Other I&C Components...................................129 draft Emergency Diesel Generator Systems.................................................................................................... 132 risk and safety management . ........................................................ 135 in use Probabilistic Risk Assessment for Internal Fire ....................................................................................137 This roadmap book is available at http://www.cvent.com/d/6cq8dc/4K.
Nuclear Sector Roadmaps 4 January 2012
Probabilistic Risk Assessment for Seismic Events ..................................................................................140 Methods for Risk Assessment for Low-Power, Shutdown, and Transition Conditions ............................143 Phoenix: Advanced, Integrated Software for Nuclear Plant Risk Management ....................................146 Risk Assessment Methodologies for External Hazards Other than Fire and Seismic ..............................149 advanced nuclear technology ...................................................... 153 in use Configuration Management for New Nuclear Plants............................................................................155 Environmentally-Assisted Fatigue........................................................................................................158 Small Modular Light Water Reactor Requirements..............................................................................161 chemistry, low level waste and radiation management .................... 165 in use Technical Solutions for Reducing Worker Radiation Exposure .............................................................167 Science and Communication of Low-Dose Radiation Risk....................................................................170 Optimized Storage and Disposition of Low and Intermediate Level Waste............................................173 Radiological Environmental Protection................................................................................................176 Water Chemistry Strategy for Reducing Radiation Fields ....................................................................179 Auxiliary System Chemistry Optimization...........................................................................................182 Development of Chemistry Software Applications.................................................................................185 Water Chemistry Guidelines for Advanced Light Water Reactors .........................................................188 Dispersant Application for Fouling Mitigation.....................................................................................191 long -term operations ................................................................... 195 in use Long-Term Operations ........................................................................................................................ 197 This roadmap book is available at http://www.cvent.com/d/6cq8dc/4K.
Table of Contents 5 January 2012
materials aging and degradations Action Plan Roadmap Summary OBJECTIVES For PWRs, the ability to monitor and demonstrate the struc-tural integrity of the reactor pressure vessel (RPV) through Metal materials degradation and aging have been problem- 80 years of operation is essential. To that end, PWRs will atic for commercial light water reactors since the mid-1970s implement EPRIs coordinated reactor vessel surveillance (e.g., PWR steam generator tube leaks, BWR recirculation program beginning in 2011, which will generate the high-pipe cracking, BWR reactor vessel internals issues, PWR fluence surveillance data and irradiated materials samples reactor pressure vessel head penetration cracking and leaks). needed to support embrittlement correlations and long-term Problems associated with actual degradation pose reliability, damage mechanism assessments. For PWR steam genera-regulatory and in some cases safety concerns, and as plants tors, recent operating experience demonstrates that knowl-age and move intolicense renewal/life-extension, the impact edge of relevant damage mechanisms may be inadequate for of the radiation environment on the susceptibility to degra- accurately projecting steam generator life. One particular dation increases. area of concern is the continuing problem of loose parts, Materials degradation and aging research at EPRI develops which can be introduced to the steam generator during the guidelines and technologies to cost-effectively manage com- manufacturing process or from the secondary side of the ponent and system degradation and aging and to inform plant. A new thermal-hydraulic computer code incorporat-strategic decisions on whether and when to replace, repair, or ing computing and technological advances over the last 30 continue operation of such components and systems. The years will be instrumental in assessing the various damage specific strategic objectives of the program are to: mechanisms. Work is underway in several areas - improved management of deposit accumulation, development accu-
- Maximize the operating life and reliability of BWR mulation computer models, loose parts identification and and PWR passive long-lived components; retrieval, etc. - for developing improved steam generator life
- Predict component degradation mechanisms and their management strategies.
rate of occurrence to inform decisions on mitigation, repair or replacement options; For BWRs, a number of plants worldwide are experiencing jet pump degradation associated with flow-induced vibra-
- Account for the impact on plant operations associated tion. Work in this area includes compiling field data (operat-with implementing materials aging management ing experience, repair history, configuration, etc.), sub-scale activities; phenomenological testing, and full scale prototypical jet-
- Develop data and physically based predictive models pump assembly testing to assess the effectiveness of proposed for remaining useful life assessments; mitigation solutions.
- Identify and disposition degradation mechanism In the area of BWR and PWR repair technology, high-chro-knowledge gaps through fundamental R&D; and mium, nickel-based weld Alloys 52 and 52M, chosen for
- Conduct research, evaluate and optimize joining, their superior resistance cracking, are used extensively for fabrication and repair processes; repair and mitigation of stress corrosion cracking (SCC) in Alloy 82/182 dissimilar metal welds joining critical reactor CURRENT ISSUES AND PLANNED RESEARCH coolant system components. Experience shows that the Near- and Mid-Term Research weldability and crack susceptibility of Alloys 52 and 52M vary widely with minor variations in material specification Research in this program for the near and midterm focuses limits. To address this issue until new/improved filler mate-on continued materials testing to better understand the phe- rials can be developed, activities are being undertaken to: 1) nomena of crack initiation and growth in light water reactor perform weldability testing to understand and rank weld-materials as well as to understand the impact of irradiation ability; 2) assess the influence of base metal composition on on both time-to-initiation and growth rate. This informa- identified weldability problems; 3) evaluate welding pro-tion will then be used to update guidance on activities to cesses and the influence of process parameters; and 4) periodically inspect and repair or replace impacted compo- develop application plans and guidance for welding nents and systems. Additionally there are activities under- vendors.
way to address specific system and component issues.
Materials Degradation and Aging 7 January 2012
Longer-Term Research specimens from retired plants or from test reactors poses separate but unique problems: for retired plants, the need to Research in the longer term will focus on BWR and PWR assure that the specimens are representative and can be irradiated materials testing and degradation models as well obtained without interfering with plant activities;; and for as improved weld repair solutions and mitigation measures.
test reactors, the need to deal with fluence levels and other BWR and PWR reactor internals are affected by several irra-scaling factors. Where field trials and tests are needed, there diation-based degradation mechanisms: irradiation-assisted are risks associated with locating and obtaining commit-stress corrosion cracking (IASCC), irradiation embrittle-ments from suitable volunteer utilities/plants. In the area of ment, creep, stress relaxation and void swelling. A number of weld repairs and weld process development, intellectual knowledge gaps could have a major impact on decisions property and licensing risks may arise with service vendors.
related to extended plant operations (beyond current design life). Long-term irradiation effects will be characterized by SUPPORTING OTHER STRATEGIC NEEDS testing materials removed from retired plants as well as using information obtained by continued participation in world- The Materials Degradation and Aging research programs wide programs to develop irradiated specimens in various work across the Nuclear Sector to provide substantial sup-test reactors. This information will then be used to develop port to both the Advanced Nuclear Technology Program new and improve existing models to predict residual life- and Long Term Operations programs in the area of materials times of irradiated BWR and PWR reactor internals management and improvement. EPRIs materials research components. programs also receive substantial support from and interac-tion with the Nondestructive Evaluation Program, the Low As noted above, Alloys 52 and 52M weld filler metal have Level Waste and Radiation Management Program, and the been difficult to use in field applications. There is a need to Water Chemistry Program.
develop a new high-chromium, nickel-based welding alloy that has the desired mechanical and corrosion resistance properties, but also has significantly improved weldability and superior resistance to weld cracking. To that end, research and laboratory weldability testing will be performed to understand the fundamental issues causing the observed problems, new alloy composition specimens will be devel-oped and laboratory tested, and full-scale mock-up and NDE testing will be performed and validated leading to work with various manufacturers to develop a new weld-metal specification.
Finally, the continued operation of light water reactors will likely require weld repair of certain reactor internals compo-nents. Knowledge gaps exist related to the weldability of irradiated nickel alloys and RPV steel and to special welding techniques for repairing high-fluence materials. The initial phase of this work will include: 1) develop a weldability assessment for PWR designs; 2) refine conventional welding models for predicting the weldability of irradiated stainless steel; 3) develop a laser welding predictive model for welding irradiated stainless steel; and (4) develop techniques for applying laser welding to reactor internals repairs. The sec-ond phase focuses on development and testing of tools such as models and advanced welding processes for the more highly irradiated materials that will be encountered during life-extension periods.
RISKS The issues and research plans inherently involve some risks.
Both the near- and long-term research plans rely on materi-als testing that can raise questions regarding applicability of such results to actual field conditions. Obtaining suitable Nuclear Sector Roadmaps 8 January 2012
in use : aging management of alloy 600 and alloy 82/182 in the steam generator channel head assembly Issue Statement period of extended operation and after the steam generators have reached 20 years of operation.
Primary water stress corrosion cracks that initiate in Alloy 600 and associated weld materials in the steam generator Inspection and Worker Dose Drivers channel head could propagate over time to pressure bound- There are no qualified techniques to inspect the steam gen-aries such as the tube-to-tubesheet weld or the carbon steel erator channel head. Existing inspection methods used by a materials in the bowl and cause primary-to-secondary leak- utility in Europe to inspect the steam generator divider plates age. Two scenarios are under consideration. result in significant worker dose. Development of a new, In the first scenario, a primary water stress corrosion crack in more efficient technique will reduce worker dose.
the divider plate assembly (Alloy 600) could reach the chan-RESULTS IMPLEMENTATION nel head, which is a pressure boundary. The channel head is carbon steel and is not susceptible to primary water stress Upon completion of this work, it is expected that:
corrosion cracking (PWSCC), but the stresses in this region are unknown and could be sufficient to cause growth via 1. Nuclear plants will update aging management plans, fatigue. This is applicable to U.S. (30 units) and non-U.S. and EPRI will update the steam generator guideline steam generators. documents based on research results related to divider plate crack propagation and cladding crack propagation; In the second scenario, PWSCC in the tubesheet cladding could propagate over time to the tube-to-tubesheet weld, 2. Vendors will offer qualified inspection techniques to which is the pressure boundary in some steam generator identify cracking in the steam generator channel designs. The applicable steam generator designs have Alloy head; and 690TT tubing and a cladding that is Alloy 600 weld mate-
- 3. Nuclear plants will update steam generator programs rial. The susceptibility of the weld between the tubing and and plant procedures to reflect research results and the cladding to PWSCC is unknown. This is applicable to operating experience.
U.S. (25 units) and non-U.S. steam generators PROJECT PLAN Lack of understanding hinders the ability to make sound decisions regarding monitoring and potential mitigation in Divider Plate Crack Propagation the channel head region.
Objectives: To determine the integrity of the steam genera-DRIVERS tor when cracks propagate to the channel head and to develop and demonstrate an inspection technique to deter-Aging Management Drivers mine if cracks exist in the channel head.
PWSCC in susceptible materials could grow over time and Review and Compilation of Existing Information reach non-susceptible or less susceptible materials that form the pressure boundary in the channel head assembly. The Other issue programs, such as the Boiling Water Reactor industry lacks understanding of the impact of such cracks on Vessel Internals Project (BWRVIP) and the Materials Reli-pressure boundary materials, which is especially important ability Program (MRP), have studied cracking behavior in ensuring safe operation as steam generators age. Research when it comes in contact with material that is not susceptible is needed to address this issue in aging management plans. to PWSCC. This information will be compiled, and existing research results will be investigated to determine the applica-Regulatory Drivers bility to the divider plate crack propagation issue.
Based on operating experience from two utilities in Europe, Analytical Modeling the U.S. Nuclear Regulatory Commission is requiring plants with Alloy 600 material in the channel head assembly Finite element modeling will be used to determine the maxi-(divider plate, stub runner, tubesheet cladding, and associ- mum stress distributions in a steam generator channel head ated welds) to 1) include the material in their aging manage- assembly. This will be used as input to determine a critical ment plans and 2) commit to inspection after entering the flaw size for the channel head material and the allowable Nuclear Sector Roadmaps 36 January 2012
flaw size considering factors of safety. Fatigue crack growth mine if the crack would ultimately penetrate the weld and analyses will be performed for the channel head to deter- lead to a through-wall crack.
mine the operating period required for the postulated initial Mockup Testing flaw to reach the allowable flaw size.
Test welds from existing mockups or from mockups built by Effective Inspection EPRIs Welding and Repair Technology Center will be ana-Existing technology to inspect the divider plate assembly lyzed for chromium content by measuring across the weld uses a combination of visual, liquid penetrant, and ultrason- cross-section. The measured chromium distributions will be ics inspections from inside the steam generator bowl. These compared to the distributions predicted using the dilution methods are slow and dose intensive. To ensure that crack- model to determine the most representative mockups to use.
ing has not propagated into the pressure boundary base If the results of the testing indicate that the tube-to-tubesheet material of the channel head assembly, a more effective solu-weld is susceptible to PWSCC, the industry would develop tion will be developed. A feasibility study will be conducted an alternate repair criteria for 690TT tubing similar to H*
to determine if existing ultrasonic methods/transducers can for Alloy 600TT tubing that would move the pressure be used from the outside of the bowl to inspect for cracking boundary from the tube end weld to some defined distance that propagates through the clad and into the base material below the top of the tubesheet.
of the steam generator bowl. If successful, mockups will be located or developed to demonstrate the inspection tech- RISKS nique. If unsuccessful, an investigation will begin to develop a technique to inspect the divider plate by going inside the Availability of Information bowl using phased array ultrasonics.
As-built information about the channel head assembly is Steam Generator Guidelines needed to build the mockups to demonstrate the inspection technique. Utilities and vendors will need to provide the as-EPRI will update the Steam Generator Integrity Assessment built information. The information may not be easily Guidelines and the Steam Generator Examination Guide-accessible.
lines to incorporate inspection and integrity assessment guidance. External Stakeholder Participation Tubesheet Cladding Crack Propagation Utility involvement is needed to build the database and develop the mockups for the tube-to-tubesheet welds. If this Objectives: To determine the range of potential chromium information is not made available to EPRI in a timely man-content in autogenous gas-tungsten-arc welds between Alloy ner, the progress of this project would be affected.
690 tubing and Alloy 82/182 cladding material and to deter-mine the susceptibility of those welds to PWSCC. RECORD OF REVISION Review and Compilation of Existing Information This record of revision will provide a high level summary of Using EPRIs Alloy 82/182 weld material databases and the major changes in the document and identify the Road-nuclear plant data on 690 tubing material, field tube-to- map Owner.
tubesheet weld compositions will be estimated. A literature revision description of change search will be conducted to determine the acceptable level of chromium for resistance to primary water stress corrosion 0 Original Issue: August 2011 cracking. Roadmap Owner: Heather Feldman 1 Original Issue: December 2011 Analytical Modeling Roadmap Owner: Heather Feldman Change: Flowchart updated. Alloy 82/182 Weld dilution models will be developed to estimate chro-was added to the roadmap title.
mium levels for autogenous gas-tungsten-arc welds between Alloy 690 tubing and Alloy 82/182 cladding. The results of this model in conjunction with the results of the literature review will be used to determine if the tube-to-tubesheet weld is susceptible to PWSCC. Finite element modeling will be used to determine the stresses in the tubesheet area. To determine how a crack in the cladding will propagate, the finite element analysis will be modified to include the initia-tion of a crack in the cladding. The model will then deter-Materials Degradation and Aging 37 January 2012
Nuclear Sector Roadmaps Aging Management of Alloy 600 and Alloy 82/ 182 in the Steam Generator Channel Head Assembly
,., } ,., ) ,.,,
J J ,.,. - } -
~0'11
'*0 of t)'\OO(Iecl operet.on Yt'llh &team gen&~:atOI$ that have reached 20 yea!S of operabOn with 1
.. jPiantlenl&llnQ the AllOy 600 Dlvidof Ploto () 0 0 <> 0 o0
~
5 r AllOy 821182 T u - C~d>ng 690TTTub>ng Mo-RAI~R.._.o l ~ ~' Teclvlique-Fte1d Tnal Clsddl>g . K -
Rewse~~1P11n
~ l 'loma"' tep>Jt crilO<>o (ARC) for 690]
tubing lnlplen~rt J
Stum Generator M.MIIJ4HMnl Pf051.~ ( SGM P p.-.F ~ - ... le-gend 0>
i--on non rwscc- Olvder Pw** FJnlle Elomoro Ano~yu~LOM<
0>...-"'-
o-<<
l d o t !Ptale-l i y -. '
c
- SG FOIIgUeG<cwth An*'Y!"! FlawS.... j 0 fW!CiedVIOOC.
~ , L.
c
.!1 e
'-ddll"'g LiteraNte RevloM on acceptabte etw04'Num lo=:
~
COn'!ttiC CIOOdono-e......- 01 Tu-IStrOI...
I 1 0 .................
38
~ Claddong Compilo Cttdctrng *Mode'lng to Ettim9M IC~*ng*Analy~e weld modcups for 1 0 won-:
FI.Wided
- ~ 690TT b<ng mat0f1ol J
., _ p~n,.. ChfOmlum Conttnt In $90182 Wto!d1 chronwm cont9f\t and fl'IOdel resutts Jln~ty Auosatnent GuldtMM RIJvltaon i-- t- l 0 ......,_ "'"'
Woldlnu Rep11lr Technology Contor (WRTCJ CIOdelong-<:ompllo 821 c-..~~c~..,~~~y *** ..,.,.P I 82m...,olpt- 800182- ...... ...,.
I Ul 0
z
~*
__,..-y UT ltiPICJtian Chonnol-t- <>y DMde< P"to-Dovelopnew lnspec:tion 1echnique d needed Divider Ptato-Oomonstrahon- UT tnsp&et1on Channat 0
0: 91-anooa! S~m GeneralcX Task Force Meetmga z
l5 Channel-c!-.ctln_.
--lot~
January 2012
~ I o-ntlry -C<odc
~ .......!
~lfiSM.£_,.,.11'V r-~ .... H~ 1- < l w Ml
~2011