ML23347A144
| ML23347A144 | |
| Person / Time | |
|---|---|
| Issue date: | 12/13/2023 |
| From: | NRC/RES/DE, Organization for Economic Co-operation and Development (OECD), Nuclear Energy Agency |
| To: | |
| Jeff Poehler 301-415-8353 | |
| Shared Package | |
| ML23347A142 | List: |
| References | |
| Download: ML23347A144 (94) | |
Text
2nd Workshop on International Harvesting Cooperation Summary Report
Workshop held on November 17, 2022 in Stockholm, Sweden Organized jointly by USNRC and OECD/NEA
Table of Contents List of abbreviations and acronyms.............................................................................................................. 1 Executive Summary....................................................................................................................................... 3 Motivation and Objectives of the Workshop................................................................................................ 4 Workshop Organization and Sessions........................................................................................................... 4 Summary of Workshop presentations and discussions................................................................................ 5 Part 1: Introductory Part........................................................................................................................... 5 Part 2: Presentations on Harvesting Activities in Countries..................................................................... 7 Part 3 Discussion on Priorities and Opportunities for Collaboration and Information Sharing............. 17 Key Takeaways from the Workshop........................................................................................................... 26 References to Previous Harvested Materials Research.............................................................................. 29
Appendix A: Workshop Agenda................................................................................................................ A-1 Appendix B: Workshop Participants.......................................................................................................... B-1 Appendix C: Harvesting Priority Topics from the 2020 International Harvesting Cooperation Workshop Participants................................................................................................................................................ C-1 Appendix D: Recent and Active Harvesting Activities............................................................................... D-1 Appendix E: Planned and Potential Harvesting Activities.......................................................................... E-1 Appendix F: 2022 International Workshop Harvesting Priorities.............................................................. F-1
ii
List of abbreviations and acronyms AMP Ageing Management Programmes ANL Argonne National Laboratory APT Atom Probe Tomography BFB Baffle Former Bolts BMI Bottom Mounted Instrumentation BWR Boiling Water Reactor CASS Cast Austenitic Stainless Steel CGR Crack Growth Rate CINR Consolidated Innovative Nuclear Research CNRA Committee on Nuclear Regulatory Activities CRDM Control Rod Drive Mechanism CRIEPI Central Research Institute of Electric Power Industry (Japan)
CSNI Committee of the Safety of Nuclear Installations DHD Deep Hole Drilling DMW Dissimilar Metal Weld DOE Department of Energy (US)
DPA Displacements Per Atom EFPY Effective Full Power Years ENSI Eidgenssisches Nuklearsicherheitsinspektorat (Switzerland)
EPRI Electric Power Research Institute (EPRI)
FT Fracture Toughness HAZ Heat Affected Zone HFEF Hot Fuel Examination Facility IAEA International Atomic Energy Agency IASCC Irradiation Assisted Stress Corrosion Cracking IMCL Irradiated Material Characterization Laboratory IMTs Issue Management Tables INL Idaho National Laboratory (US)
ISI In-Service Inspections KAERI Korea Atomic Energy Research Institute KHNP Korea Hydro and Nuclear Power LANL Los Alamos National Laboratory LAS Low Alloy Steel LMNPP Life Management of Nuclear Power Plants LTO Long Term Operation MDM Materials Degradation Matrix MOU Memorandum of Understanding NDE Non Destructive Examination NEA Nuclear Energy Agency (Organization for Economic Development and Cooperation)
NFML Nuclear Fuels and Materials Library NPPs Nuclear Power Plants NRA Nuclear Regulatory Authority (Japan)
NSUF Nuclear Science User Facilities (US)
ORNL Oak Ridge National Laboratory
1
PIE Post Irradiation Examination PLR Primary Loop Recirculation PNNL Pacific Northwest National Laboratory PSR Periodic Safety Review PTU Protective Tube Unit PWR Pressurized Water Reactor PWSCC Pressurized Water Stress Corrosion Cracking RPV Reactor Pressure Vessel RTE Rapid Turnabout Experiments RVI Reactor Vessel Internals SCC Stress Corrosion Cracking SEM Scanning Electron Microscopy SG Steam Generator SPL Sample Preparation Laboratory SS Stainless Steel TEM Transmission Electron Microscopy TPD Thimble Plugging Device USNRC United States Nuclear Regulatory Commission UT Ultrasonic Testing VVER Voda Voda Energo Reactor Water-water energetic reactor
2
Executive Summary
Research in the field of aged materials harvested from operating or decommissioned nuclear power plants can provide nuclear industry plant operators and national nuclear safety regulators with an improved understanding of the ageing mechanisms of these materials. This understanding can then support the management of plant ageing, the implementation of life extension programmes and serve as an input to operating licence renewals.
In this context, the Second Workshop on International Harvesting Co-operation was an occasion to review the status of aged metal material harvesting worldwide and to discuss priorities and opportunities for international collaborative research in this field. During the workshop, held on 17 November 2022 in Stockholm, Sweden, participating experts noted that increasing opportunities exist for harvesting and performing collaborative research on aged metallic materials. It was concluded that conducting an exercise to compile currently available harvested materials for collaborative research in a library would be valuable for developing future joint projects proposals.
Discussions also highlighted that harvesting can yield significant benefits for the nuclear sector but requires diverse and extensive stakeholder involvement. Workshop participants discussed the challenge to get utilities and decommissioning companies involved. Incentives for the private nuclear sector are that enhanced knowledge of material ageing in real operating conditions should, for operating reactors, support optimization of operating conditions, of inspections and maintenance plans, of repair and replacement activities, all of which should ultimately support long term operation (LTO). Knowledge gained is also expected to be of value for supporting the development of new build reactors with planned operating life times of 60 years and beyond. All parties engaged in the development of the nuclear sector should be made more aware of the potential benefits from harvesting and related research. Relevant standing NEA committees, CNRA and CSNI, could engage further in actions to promote harvesting research activities.
The high-complexity and cost of material harvesting and related research activities make them prime candidates for collaboration. The OECD/NEA SMILE joint project conducted by Studsvik in Sweden provides a first of a kind large collaborative effort around aged metal material harvested in several decommissioning Swedish plants. Collaborative efforts could be continued and extended in the future, considering material of interest harvested in other countries for the interest to a wider range of countries/organizations.
Participants decided to continue and deepen research prioritization and identification of opportunities to perform collaborative research that were initiated at the 2020 International Harvesting Workshop. A comprehensive priorities list is provided in this report based on the 2022 Workshop discussions that should help countries involved in harvesting target materials to harvest and help determine if existing harvested materials could address priority topics.
Participants agreed that similar review initiatives should be launched for other categories of materials (e.g., concrete, polymers including cables material) in the future. NEA could support such initiatives.
3
Motivation and Objectives of the Workshop
There are increasing opportunities and on-going efforts worldwide to harvest, and do subsequent research on, service-aged components and materials (CMs) from operating and decommissioning nuclear power plants (NPPs). Harvested CMs are of unique value because, unlike accelerated ageing laboratory tests samples, they have been exposed to actual in-service plant operating conditions (with for instance, the combination of elevated temperatures, irradiation, mechanical loading, coolant chemistry, etc. during long operating time) and can provide information and data on material ageing and degradation associated with extended operation in NPPs. Investigating harvested materials provides a unique means to gain knowledge on synergetic effects of different degradation mechanisms in real operational conditions, providing a basis for comparison with results of laboratory tests and models predictions and providing a possibility for reducing unnecessary conservatism.
Insights into ageing and degradation mechanisms in harvested material can provide confirmation of the effectiveness of ageing management approaches used by the nuclear industry. Further, evaluation of harvested material properties of systems, structures and components (SSCs) from operating or decommissioned NPPs may provide insights into the actual safety margins and increase confidence that SSCs will be capable of meeting their functional requirements during extended operations.
While nuclear regulators from across the globe endorse harvesting as an important, technically robust approach to materials characterization supporting long-term operation, international collaboration on harvesting and subsequent research on harvested material currently remain limited. This is in a large part due to challenging technical and non-technical aspects, e.g. complex on-site material cutting operations in sometimes highly radioactive environments and schedule constraints pertaining to on-site activities, which make harvesting a challenging and expensive proposition. Utilities involvement is also key for conducting successfully harvesting plans.
The significant complexity of materials harvesting projects make them prime candidates for collaboration and the international nuclear community would benefit from establishing more joint ventures to maximize research value and share related costs. Collaborative ventures would also help identify which CMs would be of higher value and may have generic applicability for LTO of NPPs.
A first workshop on international harvesting cooperation was organized in 2020 by the USNRC and the OECD/NEA (International Metals Harvesting Workshop, January 21-22, 2020 at NEA headquarters in Paris, France) to discuss opportunities for international collaboration in harvesting. Since then, harvesting efforts have progressed in a number of countries and the OECD/NEA SMILE international project, based on harvesting of decommissioning Swedish NPPs, was launched in 2021. It was therefore timely to hold a workshop in 2022 to review the current developments worldwide in harvesting programs in order to; share knowledge on building research capabilities; understand current harvesting priorities, activities, and plans; and explore opportunities for future collaboration and information sharing.
Workshop Organization and Sessions
The 2nd International Harvesting Cooperation Workshop was held at Scandic Klara hotel in Stockholm, Sweden on November 17, 2022. The workshop was jointly organized by OECD/NEA and US NRC and was
4
attended by 39 experts from 13 countries (Belgium, Canada, China, Czechia, Finland, France, Germany, Japan, Korea, Slovenia, Sweden, Switzerland and United States) and 3 international organizations (European Union, IAEA, NEA). The list of participants can be found in Appendix B.
The workshop was organized with about half of the time dedicated to presentations on harvesting activities in Czechia, Japan, Korea, Sweden, Switzerland and US and the remaining time set aside for a discussion on harvesting priorities and opportunities for collaborative research.
The workshop was organized in four sessions as follows:
Part 1. Introductory part Part 2. Presentations on Harvesting Activities in Countries Part 3. Discussion on Harvesting Priorities and on Opportunities for Harvesting Collaborations and Information Sharing The detailed workshop agenda can be found at the end of this summary report in Appendix A.
Summary of Workshop presentations and discussions
The following subsections summarize the presentations and discussion in each session.
Part 1: Introductory Part Robert Tregoning (USNRC) recalled the main outcome of the first International Harvesting Cooperation Workshop which was held in 2020 to provide participants with the background to further the discussions on harvesting priorities and opportunities for international collaboration. He recalled first the focus of the 2020 workshop was on metal components, as will be the case for the present workshop. Though harvesting is also concerning other types of material (e.g. concrete and polymer material such as electrical cables insulation layers), these materials will not be discussed at this workshop. It was agreed that it would be beneficial to have dedicated discussions involving experts in ageing of materials other than metal in the future to review harvesting opportunities for these materials. There was a tentative plan in 2019 to develop a joint safety research project (OECD/NEA Harvest project) on concrete samples harvesting in the Canadian Gentilly-2 NPP but the project was finally not engaged as challenges were faced to get the plant operators engaged in the project.
The objectives of the 2020 workshop were to:
- develop an understanding of common areas of high priority for harvesting and identify opportunities for harvesting, focusing on metal components,
- identify harvesting activities that are most valuable to the international community and may be good candidates for collaboration and information sharing,
- assess interest among countries/organizations in collaborating and/or coordinating metallic component harvesting activities, including information sharing, and
- identify next steps to pursue joint harvested materials research or information sharing among interested countries/organizations.
Planned and potential harvesting activities were reviewed in 2020 and harvesting priorities were discussed for metal components. Participants agreed on a combined list of harvesting interests, prioritizing about 40 specific technical interests within the following 10 topic areas (Appendix C):
5
inspection and non-destructive examination (NDE) activities, reactor pressure vessel (RPV) low alloy steel (LAS) welds and base metal, high dose stainless steel (SS) vessel internal materials, cast austenitic stainless steel (CASS) and SS welds, primary water stress corrosion cracking (PWSCC), steam generators, wear/fretting, environmental fatigue, and reactor design specific topics for CANDU and VVER reactors.
Discussions identified several important considerations for identifying priority components and materials for harvesting such as:
- unique field aspects of ageing and degradation that are challenging to replicate in laboratory testing,
- addressing critical technical gaps in areas of common interest,
- material/component availability as well as documentation on its fabrication and ageing conditions during its service life, and
- fleet-wide applicability of the expected data.
In 2020, the discussions also identified other key aspects for harvesting. Identifying best harvesting opportunities can be challenging without plant-specific information which is often the providence of the plant owner. On-site inspection and evaluation can provide key information to guide harvesting. The coordination with companies in charge of the decommissioning is essential.
Organizations participating in the 2020 workshop expressed their interest to cooperate internationally on harvesting and to share experiences and lessons learned and to develop a common material library. Export control issues and differing priorities among organizations were seen as potential challenges in establishing cooperative research. It was discussed that possible international cooperation mechanisms could go through NEA, IAEA and the European Union. There was a general consensus on setting-up a two-tiered approach with initial information sharing and developing a first joint project around harvesting of metal components. Additionally, discussion forums should be established to facilitate communication and streamline development of new projects. These efforts could then extend to other materials (e.g.
concrete). It was said that joint projects with shared funding would foster collaborative research on harvesting. Licensees engagement should be solicited, but may be challenging to obtain due to the technical and financial challenges related to harvesting. Efforts should also be made to explain costs and investments now in harvesting activities should contribute to facilitate getting authorizations for the prolonged operation of existing plants beyond 60 years and for further development of the nuclear sector with long-life (80 to 100 years), advanced designs.
Following the 2020 workshop, participants were asked to rank each of the 40 specific technical areas of interest and feedback was received from 8 organizations, mostly research organizations, from 6 countries.
The key takeaways from this ranking exercise are that the highest priorities are g iven to the three following topic areas (with order of highest priority): irradiation effects on SS reactor internals, RPV LAS embrittlement, and PWSCC. Expectation from this second workshop in 2022 is to refine the list of components, materials and topics for prioritization; to solicit the input from participants on prioritization following the workshop with the aim to get input from more organizations than in 2020, and ultimately, to match priorities with on-going and planned activities to identify gaps and identify opportunities for addressing these gaps.
Since the 2020 workshop, the OECD/NEA SMILE project was launched in 2021 which is evaluating metal components harvested from decommissioned Swedish plants. This project represents the realization of the 2020 workshop to use an initial project to gain experience and to serve as a stepping-stone for
6
establishing future international collaborations on research on harvested material. During this 2022 workshop, participants continued the discussion on opportunities for collaborative research, based on the SMILE experience, and identified new harvesting activities.
Part 2: Presentations on Harvesting Activities in Countries The second part of the 2022 workshop provided overviews on harvesting interests, priorities and activities in a number of countries. The presentations provided an update of on-going and planned activities since the 2020 workshop. Speakers for this session included:
Miroslava Ernestová, ÚJV ez, Czechia, Taku Arai, CRIEPI, Japan, Martin Bjurman, Studsvik, Sweden, Jeff Poehler, USNRC, US, Reiner Mailnder, ENSI, Switzerland, Frank Gift, EPRI, US, Sung-Woo Kim, KAERI, Korea Peng Xu, INL, US.
Presentation Summaries and Discussions
Harvested reactor vessel internals (RVI) from EBO V1 Unit 2 (VVER 440 type), Miroslava Ernestová,
ÚJV ez, Czechia Ms. Miroslava Ernestová presented an update of harvesting activities conducted in Czechia which were done in cooperation with the Slovak Republic related to the decommissioning of two units of the Slovak EBO V1 NPP in Jaslovsk é Bohunice. The Slovak Republic was requested to decommission two VVER 440/230 units of the plant when entering the European Union. Four VVER 440/230 reactors are currently operated at the Dukovany NPP in Czechia. The interest of ÚJV ez, as well as that of VUJE and JAVYS from the Slovak Republic, is to get data on ageing of components of the VVER 440/230 PWR Russian reactors which have a specific design and use different materials than other PWRs. ÚJV ez, as well as VUJE and JAVYS, have collected materials which were harvested from unit 2 of EBO V1 to conduct investigations and testing on these materials.
The decommissioning is funded from the BIDSF (Bohunice International Decommissioning Support Fund) with an administrator EBDR (European Bank for Reconstruction and Development). The two units under decommissioning operated between 1980 to 2006 and 1981 to 2008, corresponding to 27 and 28 effective full power years, respectively. Both units provide opportunities for harvesting highly irradiated reactor vessel internals material. The decommissioning started in 2011 and is planned to be completed in phases by 2025. The opportunity to collect material from unit 1 was missed due to the schedule of the decommissioning, but experience in sectioning the RPV and RPV internals for unit 1 was useful for the cutting of unit 2 RPV and RVI and harvesting of samples from this unit. The cutting of samples and transport to ÚJV ez have now been completed. The ÚJV ez is currently starting the analysis and testing on the received samples.
Regarding the RVI, ÚJV ez has collected 8 core shroud samples in the central zone of the core where the activity and dose were maximal, 1 PTU sample from the upper plate and 2 barrel samples (high and low active samples). The core shroud samples include 9 baffle bolts. The 11 samples, weighing around 85 kg
7
in total, were conveyed in separate transports to ÚJV ez. The RVI material is mostly Ti-stabilized austenitic stainless steel, including for the bolts. The main issues of interest for these materials are IASCC and swelling and both base material and welds samples will be examined and tested to assess the change of material properties due to exposure to high doses.
Other components, including big pieces, have also been harvested such as RPV samples, the RPV head, RPV support elements, components from the pressurizer, the main recirculation pump, contaminated samples from the primary and secon dary circuit pipelines, and components from the steam generators.
These harvested components are a property of VUJE and are placed in a VUJE storage.
Following the presentation, it was clarified that the harvesting activities and the investigations on harvested material were mostly funded by ÚJV ez and VUJE, which are research organizations. The harvesting of RVI was done with the financial support of EZ who remains the owner of the collected active RVI material stored in ÚJV ez hot cells. The others harvested and decontaminated components stored in VUJE are subsequently investigated or kept in storage for eventual future testing.
CRIEPI Research Activities on Material Aging using Decommissioned Reactors Materials Sponsored by the Nuclear Regulatory Authority, Taku Arai, CRIEPI, Japan
Mr. Taku Arai, presented an on-going research project on materials collected in decommissioned reactors which is sponsored by the Nuclear Regulatory Authority (NRA) of Japan. The research project has started in 2020 and the CRIEPI has been contracted by NRA to execute the project. The objectives of the project are to obtain data on ageing degradation mechanisms of structural materials and to confirm the conservatism and validity of the current structural integrity assessment methods. Hamaoka Unit 1 (BWR 540 MW, operated between 1976 and 2009) is advanced in its decommissioning process and was selected as the target plant. It is in the second step of the decommissioning process with removal of components such as PLR pumps and valves, providing for thermally aged CASS. The dismantling of the reactor parts is scheduled later in the third part of decommissioning between 2023 and 2029.
The Japanese utilities, except Chubu Electric power company, are still reluctant to consider harvesting materials for research in the plants under decommissioning because the decommissioning schedule remains uncertain and decommissioning operations requires huge efforts and resources. Another important objective of this project is thus to establish good practices in harvesting to make other utilities more active in harvesting material for research. Since the research using decommissioned materials requires relatively long implementation period, the CRIEPI will conduct this research in cooperation with utilities and vendors, considering their decommissioning process.
Three research items, described in the next paragraphs, will be addressed in the project. Though neutron irradiation embrittlement of RPV steel is a major issue in Japan, it will not be addressed in the present project because dismantling of the RPV is not expected to occur soon.
The first research topic of interest is the verification of the applicability of the thermal ageing prediction model for CASS. This is one of the major aging phenomena for LTO. Microstructural changes due to thermal ageing were measured, showing decomposition of the ferrite phase (ferrite phase decomposition and G-phase precipitation). The applicability of H3T model (fracture toughness) to BWR materials was
8
verified with some CASS materials used in Japanese BWR and actual component materials. Eight CASS blocks (CF8M) were extracted from PLR pumps and the transportation to a research institute were successfully completed in the fiscal year 2021. Fracture toughness tests, tensile properties tests, hardness tests and Charpy impact tests are ongoing over the fiscal years2022-2023. The knowledge of ageing in CASS is expected to be significantly enhanced with these new data.
The second topic is fracture toughness degradation of core internals due to neutron irradiation. The evaluation diagram in the JSME code is established considering the reduction of fracture toughness due to neutron irradiation. A fracture toughness curve was established from data obtained on irradiated material. However, few data were obtained from real material irradiated in Japanese BWRs. Considering the decommissioning schedule, the candidate materials to fill this gap could be collected in the next decommissioning phase or later. There are also many issues that need to be resolved to realize the ongoing elastic plastic fracture toughness data. Entering the most challenging phase is planned in 2023 with dismantling of the reactor area and collecting samples. This plan is only tentative as ensuring progress on the decommissioning process remains the top priority.
The last topic is confirmation of the persistence of compressive residual stress induced by the peening process. Peening is widely used in Japan to induce compressive residual stresses on the components surfaces to prevent stress corrosion cracking. Depth of compressive stress induced by peening is less than 1 mm. It depends on the peening method and processing conditions. The CRIEPI plans to investigate the effects of the decontamination process which might reduce the layer which has the compressive stress. A mock-up was fabricated to investigate this effect with simulated decontamination tests. CRIEPI also plans to look at cutting effects on the relaxation of residual stress. Currently, the testing, related measurements (X-ray stress measurements) and samples are being prepared. Actual testing campaigns are planned to start in 2023.
SMILE - Ongoing Harvesting and Testing Activities, Martin Bjurman, Studsvik Nuclear AB, Sweden
Mr. Martin Bjurman from Studsvik Nuclear AB presented details of the OECD/NEA SMILE (Studsvik Material Integrity Life Extension) project: approach, technical highlights, schedule and current status.
Studsvik is the operating agent of the project. The SMILE project is a cooperative five-year project which was launched in 2021 under the auspices of the Nuclear Energy Agency (NEA) of the OECD with the objective to provide knowledge to support LWR operators and regulatory authorities worldwide in plant ageing management. At the end of 2022, organisations from 8 countries participate to the project which has a total budget of about 17 M. The main objectives of the project are to provide critical data and understanding of materials ageing mechanisms in support of plant ageing management, life extension programmes, and operating license renewals. The project is focusing on high priority knowledge gaps identified through prioritization reviews (e.g. the EPRI Material Damage Matrix (MDM) studies and Issue Management Tables (IMTs) but also prioritization conducted within the project itself) and is based on methodologies for material ageing management, e.g. the IAEA International Generic Ageing Lessons Learnt (IGALL).
Material harvesting was initiated in 2019, before the start of the SMILE project, and will continue for many components. SMILE leverages a unique opportunity to examine and test materials harvested from
9
Swedish reactors decommissioned after more than 40 years of operation. It includes collection, examination and testing of highly irradiated wrought stainless steel and weld metal, alloy 690 and alloy 52 weld metal from the worlds second oldest replacement steam generator (from 1989) and the oldest RPV head (from 1995). RPV specimens from low to high doses, including archive material and surveillance specimens, are also investigated in the project. Materials originate from Oskarshamn 1 and 2 BWR plants and Ringhals 2 and 3 PWR plants.
The project consists of four tasks which have been prioritized by the project members. Task 1 aims at building a materials library and includes activities for materials acquisition, for compiling the related documentation, and for calculating materials temperature and dose exposures during operating life. Task 2 focuses on irradiation embrittlement of RPV LAS. Task 3 addresses irradiation embrittlement (including for SS welds) and IASCC of SS core support structures and internals. Task 4 deals with evaluating SCC resistance in Alloy 600/182/82 austenitic pressure boundary materials and dissimilar metal welds (DMWs),
SCC resistance of stainless steels welds, and PWSCC resistance and thermal stability of alloy 690/152/52 pressure boundary ma terials and DMWs.
The strategy for selecting and harvesting material consists of several interconnected steps that impact quality of extracted sections, cost and time schedule for the SMILE project and the dismantling project which have distinctly different objectives, needs and time schedules. To gain plant owners acceptance for harvesting, the risks that can affect the dismantling time schedule need to be small. Further limitations are imposed by the subsequent transport, handling and sample cutting and preparation before testing.
Martin Bjurman then provided details of the harvesting schedule for material already available at Studsvik, material being extracted, and material which will be extracted in 2023 and later. The current harvesting schedule can be summarized as follows: internals from Oskarshamn 1 and 2 BWRs are available at Studsvik and internals from Ringhals 2 are being extracted. SS and 600/690 Ni-base DMW penetrations and piping are planned to be extracted in the first half of 2023. Alloy 690 steam generator tubes from Ringhals 2 are planned to be extracted in 2023. RPV beltline, penetrations and core shroud support of Oskarshamn 2 will be extracted in 2023. Ringhals 2 RPV material extraction is postponed until 2026.
The testing and characterizations of materials were summarized next. They include, for RPV materials, -
spectroscopy to validate dose calculations, residual stress measurements, microstructural characterization by light optical microscopy (including hardness) and SEM. In addition, tensile and fracture toughness testing will be conducted at various doses, through the thickness direction of the RPV, along with Charpy V tests, auger spectroscopy, TEM, and APT examination of fracture surfaces on a selection of materials. For irradiated RVI materials, testing and characterization include -spectroscopy to validate dose calculations, NDE to detect possible flaws, residual stress measurements of welds, density measurements of baffle plate and bolt at various doses, microstructural characterization by light optical microscopy and SEM. Additionally, TEM and APT will be performed on selected materials. Mechanical testing will include both tensile, FT, and CGR-testing at various fluences. For 600/182/82 and stainless steel alloys, testing and characterization include tensile, CGR in BWR and PWR environments, and SCC testing and hardness measurements in regions of interest. For Alloy 690/152/52, visual inspection and NDE will be used to detect flaws in SG tubes, and CRDM penetrations. In addition, residual stress measurements of CRDM penetrations, microstructural characterization by light optical microscopy
10
(including hardness) and SEM, and eventually high-resolution examinations, will be performed to determine if long range order exists in SG tubes. Mechanical testing will include tensile and CGR testing of CRDM penetrations and crack initiation testing of SG tubes.
Though the SMILE project has started in 2021, it was indicated that interested organizations who are not part of the project can still join and participate in the project.
After the presentation, participants asked about the techniques used to measure the residual stresses on-site. Deep hole drilling, iDHD, strain gauges, deflectometers are being considered for on-site residual stress measurements.
Update on NRC Materials Harvesting Activities, Jeff Poehler, USNRC, US
Mr. Jeff Poehler described NRC`s materials harvesting background and current situation. Historically, US harvesting efforts have been performed on a broad range of components of interest. Current harvesting objectives focus on materials ageing during long -term operation to confirm results from laboratory experiments and analytical simulations, and to reduce uncertainties in the current state of knowledge of ageing and NDE effectiveness in informing NRC to review ageing management program s.
In the past, harvesting efforts were limited as few plants were shutting down offering few opportunities for investigating harvested material. Recently, a number of plants have shut down and entered in the decommissioning process. This offers more supply of harvesting opportunities than in the past, with highly aged components and this situation calls for a more proactive strategic approach. Since about 2015, the NRC has been developing and implementing such a proactive approach.
The NRCs proactive harvesting strategy consists in 1) identifying and prioritizing harvesting interests, 2) considering use of previously harvested materials when possible 3) gathering information on attributes of decommissioning plants to identify harvesting opportunities. The harvesting interest s are prioritized through the use of four technical prioritization criteria, such as criticalness of the technical issue, importance of harvested materials over laboratory aging, applicability to US operating fleet, and regulatory considerations related to inspections and AMP uncertainties. Considering already harvested material can greatly reduce costs, time and complexity compared to new harvesting. Gathering adequate plant attribute information to identify harvesting opportunities across the population of decommissioning plants may be challenging because of the dependence on the licensees for sharing that information.
Jeff Poehler then listed several high priorities for metals ageing characterization and testing from NRCs perspective. These include:
- NDE and mechanical testing of alloy 690 thermally treated SG tubes with shallow flaws,
- fracture toughness and microstructure of thermally aged unirradiated CASS,
- residual stress measurements, crack initiation and growth rate testing and flaw characterization of BMI nozzles with known PWSCC indications,
- fracture toughness, IASCC CGR and microstructure of high fluence SS welds (> 2 dpa).
The NRC has catalogued previously harvested materials from prior NRC-sponsored research. PNNL has a large array of components from penetrations up to large piping sections used for NDE research, Battelle has large primary system piping and elbows and ANL smaller irradiated reactor internals materials. Other
11
sources of previously harvested materials include the US DOE Nuclear Fuels and Materials Library (NFML) the Studsvik SMILE-related and other harvested material, and the Halden Reactor Project materials.
The NRC worked with EPRI to develop a harvesting opportunities table which attempts to summarize the plant attributes for domestic and international from decommissioning or announced shutdown plants, to identify potential matches with existing priorities.
NRCs recent and current metals harvesting activities and related investigations were presented. They include investigations on neutron absorber materials from Zion, reactor internals base and weld materials from Zorita, and various metallic components in SMILE. The NRC is currently targeting harvesting from domestic and international plants focused on high value components in cooperation and coordination with DOE and EPRI.
The NRC considers partnering and information sharing essential to manage the costs and complexity associated with harvesting activities and maximizes cooperation with other organizations such as DOE and EPRI in the US and organizations outside US. The SMILE project is a good example of the advantages of international cooperation on harvesting. The NRC also periodically holds public meetings on harvesting to solicit feedback from stakeholders.
The NRC also has interest in other materials such as RPV supports, concrete, cables and flood barriers.
The NRC is actively working on developing cooperation with other organizations for these materials.
Following the presentation, NRC staff clarified that NRC collaborates with DOE and EPRI under separate, existing Memorandum of Understanding (MOU) agreements. Collaborative work under MOU is generally done in-kind. In some cases, specific agreements are established, particularly when work requires specific shared funding.
The participants discussed potential mechanisms to motivate utilities that are decommissioning plants for harvesting aged materials. NRC commented that there is currently no regulatory leverage. It is particularly challenging to motivate utilities which are decommissioning plants and have no other operating plants that are planning for LTO. Utilities which have LTO plans are more prone to contribute to harvesting. Doing more inspections during the decommissioning process could be a way to get some information but even this is challenging and the information may be of limited value. EPRI commented that they have socialized the harvesting opportunity table with utilities. However, even collecting plant attribute information to inform harvesting opportunities is challenging as some uti lities cannot easily collect and provide the needed information.
Preparatory works for harvesting at the Beznau NPP, Reiner Mailnder, ENSI, Switzerland
Mr. Reiner Mailnder from ENSI presented background information on the nuclear energy production in Switzerland and explained some reactor details of Beznau unit 1 and unit 2 (Westinghouse 2loop, PWR),
one of the oldest NPPs. First criticalities were in 1969 and 1971 and relatively high values of fluence are expected. RPV fluence will reach around 6x1019 at inner surface of the RPV of Beznau 1 after sixty years of operation. Afterwards, current ENSI planning was shown. First step is an internal evaluation about potential harvesting interests, followed by discussion with licensee (AXPO) about how to integrate harvesting operations into dismantling of the units. International cooperation will be the second step, and it is at this stage premature to discuss it.
12
During the discussion after the presentation, the meeting participants asked about the involvement of the utilities (licensees) for harvesting. A licensee who owns only one decommissioni ng plant and no other operating plants, is commonly hesitant to cooperate in harvesting.
Additional safety measures were required after the Fukushima Daiichi accident. For utilities who decided to decommission early because of the costs for those additional measures, it was difficult to devote extra resources for harvesting during decommissioning. Harvesting would be the critical path for the decommissioning schedule. This was also a big challenge for the involvement of utilities.
Incentive for utilities such as legal obligation for the long-term operation for plant operators were discussed. For extended operation of nuclear fleet beyond initial design operating life, operating experience data from decommissioning plants are important. Incorporating it in some form as a requirement of the Periodic Safety Review (PSR) could be envisaged. However, not all countries, including the US, have PSRs. It could be beneficial to establish a mechanism that would prevent decommissioning a plant without gaining any knowledge of interest to LTO.
The value of harvesting is that it could contribute to asset management, preservation of SSCs and managing the plant more efficiently. Spreading the vision to obtain data to support LTO should be an incentive for utilities. The participants reaffirmed that the co-operation of licensees is essential and needed for an effective approach.
Materials Harvesting Priorities -International Materials Research-, Frank Gift, EPRI, US
Mr. Frank Gift from EPRI provided an overview of EPRI`s materials harvesting activities. EPRI has an interest in materials harvesting because such materials are the most representative source for characterizing ageing degradation in reactor internal components, RPVs and other structural components.
EPRI provides inputs to its members for supporting their ageing management programmes. The EPRI-led Zorita Internals Research Project set the precedent for industry efforts in material harvesting. Results of the harvesting of reactor materials and testing programmes support the technical bases for modelling materials ageing. Issue Management Tables (IMTs) research gaps for a number of reactor designs (PWR, BWR, VVER and CANDU) may be addressed by these programmes. Expected benefits for EPRIs members are to alleviate conservatisms in ageing degradation assessments, potentially identify leading indicators for materials degradation, support development of ageing management strategies including optimization of the frequency and scope of ageing-related inspections. Harvesting provides an opportunity to proactively research aged material properties and to potentially demonstrate novel inspection, mitigation and repair techniques. In some cases, material harvesting is not required and inspection data from decommissioned reactors can be obtained to provide comparable materials ageing data on real plant components and materials with a lower cost and lead time to obtain results. EPRI is also monitoring progress in domestic and international programmes on harvested materials and components to inform its members.
Regarding harvesting priorities, materials which provides reactor-relevant environmental and ageing parameters for material studies, that resolve concerns with laboratory time-accelerated testing, and which address EPRI research gaps in the Materials Degradation Matrix (MDM) and Issue Management Tables (IMTs) should be harvested as a priority. Such harvested materials also provide cost savings from replication in a laboratory of long-term operation conditions and can offer unique material availability
13
(such as legacy materials that have alloy compositions and which were fabricated with metal-working techniques and thermal processing steps that cannot be reproduced today). Concrete materials (wetted, irradiated, with alkali-silica reaction, from containment (baseline)) and electrical cables are additional priority materials for harvesting in addition to the metallic materials that were the focus of discussion.
Frank Gift explained that EPRI organizes its metallic materials research in high level materials research focus areas (MRFAs) and identifies harvesting targets in applicable MRFAs, such as MRFA 2 for stainless steel alloys, MRFA 3 for nickel-based alloys, and MRFA 4 for low alloy steel. For stainless steel alloys, harvesting targets are irradiated welds, Heat Affected Zones (HAZ) and base metal for FT and CGR, high-dose and high gamma heating regions for void-swelling characterization, and stainless steels welds for characterizing original fabrication conditions and material integrity. For the nickel-based alloys, harvesting targets are high strength alloys for FT, aged/weld strained material for SCC initiation and growth, and thermally aged Ni-based alloys for PWSCC testing and assessing hardening and microstructural changes.
For low alloy steel, harvesting targets are extended belt line RPV materials (plates, welds, forging) for dosimetry and mechanical testing, vessel material and welds for evaluation of homogeneity, and base metal and welds after a long-time in service at elevated temperature for assessing thermal embrittlement.
EPRI is involved in several collaborative harvesting projects, some examples are the SMILE project and the Sherlock program (CRUAS Unit 4 Steam generator with Alloy 600 tubes with more than 30 years operational history available through an EDF program). The second phase of the research in the Sherlock program aims at examining the materials in the laboratory.
EPRI also implements research programmes that are EPRI-funded harvesting and testing, such as one programme addressing Alloy 690TT general corrosion (first data were obtained from Oconee reactor SG tubes and new data will be obtained from plugged SG tubes of Ringhals Unit 2 to enhance SG tubes corrosion models). Another possible programme is thermal shield (TS) inspections at Indian Point, which is now owned by a decommissioning company (i.e., not the original operator and licensee). The cooperation with decommissioning agents requires a different approach. Maintaining their decommissioning schedule and costs, and working efficiently are the most important aspects to permitting harvesting and research to occur in parallel. The recent operating experience at Salem with thermal shield component degradation could potentially apply to up to 30 plants. Another operating plant will perform detailed inspections with neutron noise monitoring to identify vibrations that may lead to thermal shield degradation. EPRI is coordinating planned inspections with the decommissioning agent that avoids conflicting with critical schedule work and internal interferences. Harvesting work should be done outside of the active segmentation and decommissioning work, such as in the night work or over the weekend. Frank Gift also explained an additively manufactured thimble plugging device will be harvested from Byron unit 1 (operating plant). Thimble plugging devices (TPDs) are designed and manufactured to interface with lifting tool and fuel assembly and normally have a life-span of 8-12 operating cycles. EPRI is collaborating with Constellation and Westinghouse to develop a harvesting program after 3 cycles of operation. It was fabricated from Type 316(L) stainless steel and installed in Byron Unit 1 in March 2020. In September 2024, the AM-TPD will be removed and sent to the hot cell, likely after a cool down period, for PIE.
Frank Gift pointed out the importance of forward-looking harvesting and harvesting collaboration.
Collaboration could enable high-cost projects to be implemented. He also suggested sharing the tested
14
data among utilities could be an excellent alternative to harvesting collaboration to reduce eventual duplications. EPRI welcomes discussions on harvesting, inspections and data sharing opportunities.
After the presentation, a question was raised about the publication of EPRIs research and access to the data. EPRI answered that it could be open to the public in 5-10 years. But many of the programs are currently closed, keeping the proprietary information for the funding parties. Participants asked EPRI to encourage its client utilities to participate more in harvesting activities and information sharing. EPRI agreed that its members had a lot of valuable experiences which could be beneficial to the nuclear community.
Update on Materials Harvesting Activities in Korea -Kori 1 BFB Failure Analysis and Technology Development-, Sung-Woo Kim, KAERI, Korea
Decommissioning plan of Kori 1 plant and the research project on the failure analysis of Baffle Former Bolts (BFB) of Kori-1 were described by Mr. Sung-Woo Kim. The Kori-1 plant operated between 1978 and 2017 (40 years, 30.16 EFPY) and is a Westinghouse 2 loops design, 600 MWe. The KHNP, owner of the plant prepared the decommissioning plan and submitted the final decommissioning plan (FDP) in 2021 and it expected to get approval in 2024. A materials harvesting research project has been established and is now on phase I & II (2019-2024, 2021-2025). Mr. Kim presented a summary of the project Failure Analysis of BFB of Kori-1. The goal of this research is to understand the failure mechanism and perform a root cause analysis of BFBs.
In 2015, indications were found in 8 BFBs from ISI by UT. RVI design changed from down-flow to up-flow in 1986. Two failed bolts and two intact bolts neighbouring failed bolts were pulled out and transferred to KAERI-Irradiation Materials Evaluation Facility (IMEF) and several tests have started. Test results for failed bolts (phase array UT, 3D computed tomography X-ray, fractography of crack surface) were presented. Phase array UT (on-site and in IMEF hot-cell) showed no back-wall signal for failed bolts. Mr.
Kim introduced on-going and future tasks, such as microstructure observation, IASCC testing and post-analysis. The construction of a new building for ITL (IASCC test facility) is on-going and future testing are planned. Some processes are waiting for licensing and new test data are expected after 2024.
Planning of materials harvesting is a part of this project. Target components and ageing mechanisms were described. The maximum dpa of the baffle plate is over 60 dpa and the welding part is over 5 dpa already.
KAERI published 2 documents on targeted components and ageing mechanisms. Phases I and II will end in 2024 and 2025, respectively. Future plans (Phase III) are to evaluate irradiation embrittlement of RPV nozzle steel and beltline materials, IASCC and fracture toughness of RPV plates, etc.
Questions were raised after the presentation, about the impact of the early conversion of RVI to an up-flow design on Kori-1. Based on the analysis, the impact to degradation was very small. Another question was asked regarding the possibility for international collaborations in the future. Korea is now communicating with other countries for future collaboration s. Collaborations with countries who have a significant experience in harvesting are envisaged. Korea has engaged discussions with the SMILE project.
The Korean original schedule on harvesting research is slightly delayed and Korea is considering international collaborations to address some issues. Regarding collaboration with utilities in Korea, all NPPs are owned by KHNP and KHNP is also in charge of plants decommissioning. This is an advantage for
15
the harvesting project. The lessons learned from the project can be useful for all types of plants. Some of the addressed topics are of interest to other countries.
Nuclear Fuels and Materials Library (NFML) and Irradiated Materials Harvesting with the Nuclear Science User Facilities (NSUF), Peng Xu, Idaho National Laboratory (INL), US
Mr. Peng Xu highlighted the main goals of the NSUF program which supports DOE-NE missions, in particular regarding enhancing understanding of radiation damages on material and fuels to address the related challenges for the current fleet of reactors and for the next generation. The Nuclear Fuel and Materials Library (NFML) is a library containing irradiated physical components that is owned by the DOE-NE and curated by the NSUF. Regarding harvesting, NSUF offers the potential to collect valuable reactor components during reactor decommissioning activities, without the requirements for a pre-determined test program to be in place. Through competitive proposal processes, researchers from universities, government laboratories and agencies, industry and small business can obtain access to these materials for use in DOEs funded projects, such as Consolidated Innovative Nuclear Researches (CINRs) and Rapid Turnaround Experiments (RTEs). Material can also be accessed by establishing specific agreements with DOE. NSUF also provides world class irradiation and PIE facilities for testing and characterizing harvested materials.
The NFML inventory of components/samples has significantly grown since 2017 with more than 8433 components/samples included in the library to date. Those components came in all different forms/shapes and from different sources. The NFML is the largest global open archive of high-value irradiated fuels and materials from test, commercial, and decommissioned power reactors, and valuable donations from other sources. The library includes associated information such as compositions, irradiation conditions and publications. This is one of the large archives that is available for the global research in the nuclear community.
Mr. Xu explained that currently there are 24 commercial power reactors undergoing decommissioning in the US. Irradiated components for research are important to understanding irradiation behaviour of core structural materials, for license renewal and lifetime extension for the commercial nuclear power plants.
On the other hand, the harvesting process is time consuming and costly. It must be inserted into the decommissioning critical path. Incentivizing commercial entities to support harvesting is the biggest challenge to overcome. The harvesting requires a highly cooperative, integrated approach with the decommissioning entities and the power plants. Collaboration is key to make harvesting efforts successful.
He then presented contributions to the NFML from NPPs or research reactors such as X750 Inconel alloy spacers from a CANDU reactor (CNL contribution), core shroud 347 SS Baffle Former Bolts (Westinghouse contribution), BOR-60 Russian test reactor highly irradiated steel/alloy samples (ORNL, LANL, Terrapower contribution), 304 SS core shroud samples from a commercial LWR NPP (EPRI, Southern Nuclear contribution) etc. Mr. Xu encouraged participants to contact him before materials used in project s are considered for waste disposal. NSUF is also currently working with the Halden Reactor Project staff to bring samples to the US to be added to the NFML. For Crystal River NPP, NSUF completed the phase I of the harvesting project planning and cost estimate, but it didnt move forward due to the funding.
Regarding the Zion plant which was decommissioned from 1998, harvested materials are currently being
16
tested in the LWRS program and a transfer of a portion of the Zion material to NSUF is planned for when testing is complete (at the end of 2023).
Next, Mr. Xu explained the post irradiation examination facilities at INL. Those are Hot Fuel Examination Facility (HFEF) for fuel engineering PIE, Irradiated Materials Characterization Laboratory (IMCL) for advanced characterization, and Sample Preparation Laboratory (SPL) which is being constructed for structural materials PIE. IMCL is a new facility with 15 high-end instruments for highly irradiated nuclear fuel and materials. SPL, when completed, will be the most modern reactor structural materials testing and analysis facility in the world. The facility has a larger capacity to do the structural materials PIE. The SPL will start operation in 2024-2025. NSUF manages the NFML and is engaging with all stakeholders.
Participants are highly recommended to access their database through the website.
Following the presentation, participants asked about accessibility to the NFML for non-US organizations.
Mr. Xu answered that non-US organizations can contribute by providing materials and can get access to material to do research through specific agreements with DOE. He said there are already agreements in place between DOE and some non-US organizations. He recommended that participants view the library website and provide feedback. NEA asked if NFML material could be used in international joint research projects. Presently, there are not such initiatives but Mr. Xu commented this is an interesting suggestion which could be further considered in the future.
The International Network on Life Management of Nuclear Power Plants (LMNPP Network), Ms.
Qun Yu, IAEA
Ms. Yu explained the background of the project. According to the NPPs distribution by their age, about 66
% of operating units are over 30 years. One third of the operating units is now under long term operation.
IAEA supports member states on plant life management and establish es this new network LMNPP. This network will be a very inclusive network accessible on the IAEA CONNECT platform to promote international cooperation. She described the objectives and scope of the network and showed the LMNPP public website available at https://nucleus.iaea.org/sites/connect/LMNPPpublic/SitePages/Home.aspx.
LMNPP member s can upload any related information to this website to promote their activities. She also described the members area of the LMNPP websit e and its function, the governance structure of the network and gave details about the 1st technical meeting on the LMNPP Network. Five different working groups were proposed at the 1st meeting: WG1 (good practices and lessons from LTO of NPPs), WG2 (OPEX from implementation of ageing management pre-operational phase for new nuclea r builds), WG3 (risk-informed approach in AMP and LTO decisions), WG4 (environmental impacts to NPP life management) and WG5 (equipment survivability for beyond design base accident). This network is open to everyone and if all member states request, new topics can be added as new working groups. The LNMPP can co-host, co-publish with other organization. Any idea for cooperation is welcome and harvesting is one of the topics.
Part 3 Discussion on Priorities and Opportunities for Collaboration and Information Sharing
Harvesting Priorities
17
Conducting a thorough prioritization exercise is still considered by the workshop participants as an important step as, in relation to challenges in harvesting activities, it should help focusing harvesting material with highest expected value for supporting long-term operation and identifying any potential gaps in harvesting, i.e. key aged material which have not yet been considered in harvesting and which would bring key knowledge for long-term operation.
The discussion consisted in reviewing main topics of interest for participants regarding research related to long term operation of NPPs with different designs (BWRs, PWRs, VVERs and to a lesser extent CANDU reactors) with a focus on identifying aged metal material to harvest in priority for the purpose of advancing knowledge on key ageing phenomena and enhancing related model ling, advancing knowledge on effectiveness of mitigation and repairing methods and inspection including NDE, and where relevant, of providing a basis for comparison with laboratory testing made on artificially aged material.
The discussion on prioritization was based on a first topics table which was established following the 2020 international harvesting workshop (given in Appendix C) where 10 topic areas and 40 specific technical interests had been identified. It resulted in addition of new topics which are now considered important by participants and the topic area classification was refined. Participants decided to classify the topics in four categories: (1) ageing mechanisms related to PWSCC, environmental fatigue, wear and fretting, and ageing effects on fabrication defects; (2) mitigation and inspection related to the effectiveness of repair techniques and inspection and NDE activities; (3) component specific topics on RPV internals, RPV LAS weld and base metal, CASS and SS welds and steam generators; and (4) reactor-specific topics for CANDU and VVERs.
It was decided that the updated priority topics table would be sent to the workshop participants for review and comment to establish a consolidated version that participants would then rank. The following actions were associated with implementing this process:
Action HWS2022-1: the priority topics table updated at the workshop will be distributed after the workshop by the NEA Secretariat to workshop participants for review and additional comments.
Action HWS2022-2: USNRC and NEA will consolidate the table with participants feedback.
Action HWS2022-3: (a) the NEA Secretariat will distribute the consolidated table to participants for priority ranking. (b) the participants will provide their ranking (priority in their organization) for the listed topics, reflecting increasing interest.
After completing these actions (detailed participants ranking are provided in Appendix F), the consolidated table considering the workshop discussions and the ranking assessment provided by participants (as identified by each column heading) was developed and is provided below (Table 1).
18
Table 1 2022 Priority Topics for Harvesting (1/2)
- Each response was numerically related as follows; High (5), High/Medium (4), Medium (3), Medium/Low(2), Low(1). Ave. is the arithmetic mean of the columns.
19
Table 1 2022 Priority Topics for Harvesting (2/2)
20
Table 1 lists the numerical ranking for each participating organization for each specific interest area. In their response sheets, participants were asked to rank each specific interest area high (H), medium (M),
or low (L). Some participants also ranked certain specific interests medium/low (M/L) or medium/high (M/H). Rankings of H, M/H, M, M/L and L were assigned the following numerical rankings: 5, 4, 3, 2, 1.
Some respondents did not rank certain specific interest areas, in which case a ranking of zero was assigned for the purpose of averaging. An average ranking was computed for each specific interest area.
Aging Mechanisms This category includes several aging mechanisms for which more data is desired, such as PWSCC and environmental fatigue. Some of the specific line items are for specific material and degradation mechanism combinations, such as crack growth rate testing of Alloy 690/52/152. The individual specific interest areas in this category were generally moderately rated, with no area with an average ranking of 3.3 or above.
PWSCC The highest rated specific interest areas under this topic are Evaluate mitigation effectiveness (water chemistry, peening, MSIP), and SS DMW components (stainless steel dissimilar metal weld components), both of which have an average ranking of 3.2 Respondents that rated Evaluate mitigation effectiveness highly commented that they were interested in evaluating effectiveness of mitigation of flaws (NRC), and KAERI and KHNP indicated they were interested in specific mitigation methods (weld overlay, zinc addition). For the SS DMW components, interest area, some respondents indicated interest in stainless steel weld related to the operating experience in France with SCC of stainless steel piping (IRSN, NRC, EPRI). In addition, NRC is interested in thermal aging of stainless steel welds.
Environmental Fatigue
The two specific interest areas under this topic had medium average rankings. The specific interest Validate models and design practices, was ranked H by GRS/MPA, KHNP, and Bel V. For this specific interest, NRC noted that harvesting could be useful to validate conservatism or identify over-conservatism in current models, but components selected must have a well characterized loading/thermal history and ideally would have evidence of some service induced cracking. IRSN, which ranked this area M or H depending on potential future degradation, indicated interest in prioritizing the areas with the highest usage factors and cracks in high usage factor areas. For the specific interest Evaluate irradiation effects (indicate if interest is reactor-type specific), CNSC indicated there is potential application to 304L stainless steel used in CANDU Calandria vessels, and NRC commented that priority should be on components with low-cycle fatigue loading.
Wear/Fretting The specific interest areas in this area had low average rankings. CRDM thermal sleeves had the highest average at 1.8. The NRC ranked this specific interest M/H, citing operating experience in US PWRS, and that these thermal sleeves are more difficult to inspect in-situ. NRC further stated it would only support harvesting of thermal sleeves if both degraded and undegraded thermal sleeves could be obtained to evaluate wear and cracking locations, root cause, and possibly to develop and validate models. EPRI also ranked CRDM thermal sleeves H, but only for inspection activity not material retrieval.
21
Aging Effects on Fabrication Defects This topic has only one specific interest area, Consider interaction between fabrication defects and service-induced aging (e.g., weld lack of fusion and PWSCC, RPV embrittlement and hydrogen flaking),
with an average ranking of 2.7. GRS/MPA, KHNP/KAERI, and CNSC ranked this area H. IRSN ranked this specific interest M for [RPV] inner surface defects, but ranked RPV embrittlement and flaking L. NRC ranked this specific interest M/H noting that items such as RPV materials with hydrogen flakes (already identified) and PWR bottom mounted instrumentation nozzles (weld lack-of fusion implicated in SCC) may be of interest if available. NRC further noted that the challenge is to find components that have both fabrication and service-induced aging.
Mitigation and Inspection
This category includes topics Effectiveness of Repair Techniques and Inspection and NDE Activities.
Effectiveness of Repair Techniques includes only one specific interest area, Harvest repaired components (i.e., peened, overlayed, mechanical stress improvement for evaluation and testing of long-term repair effectiveness, which had a high average rating of 3.6. Several respondents (VTT, STUK, NRC) indicate interest in evaluating the effectiveness of weld stress improvement techniques such as MSIP, weld overlay, induction heat treatment, and peening.
Under the Inspection and NDE Techniques topic, the specific interest area Evaluate existing and emerging NDE procedures and advances /developments, had the highest rating of 3.8. For this area, the NRC suggested performing the advanced NDE techniques in-situ on critical components in the decommissioning plant, followed by destructive examination of the same components after harvesting.
Another specific interest area with a fairly high average ranking (3.8) is Inspect high usage fatigue locations (including environmental factors), for which commenters indicated an interest in confirming environmental fatigue factors and examining high fatigue usage locations (NRC, IRSN). Another fairly highly ranked area (3.4) is Evaluate flaw distributions, types, and characteristics (e.g., fabrication or service induced & mechanism) in critical components/welds.
Component-specific
The next category is Component-specific. This category includes several subcategories, each of which have many specific interest areas.
RPV Internals The RPV Internals topic focuses on irradiation-assisted degradation of stainless steel, including irradiation-assisted stress corrosion cracking (IASCC), reduction in fracture toughness, void swelling, and irradiation-assisted creep. This subtopic area includes the highest overall ranked specific interest area, Evaluate IASCC initiation and growth, especially for high-fluence (e.g., > 50 dpa) materials, with an average ranking of 4.5. Several respondents noted that this area represents a data gap, including NRC and UJV, and KHNP and KAERI indicated this data is needed to support long-term operation. The next highest average ranked specific interest area under this subtopic is SS welds (>2 dpa), i.e. for CGR and fracture toughness testing (FT) testing, which received an average ranking of 3.4. EPRI indicated it was specifically interested in stainless steel base metal in the range of 5-10 dpa, NRC indicated interest in defining thresholds for higher crack growth rates and low toughness, and UJV indicated that this is a VVER IMT gap.
22
RPV LAS Weld and Base Metal This topic focuses on radiation embrittlement of low-alloy steels (LAS) used in RPVs. Thermal aging embrittlement and flaw distribution issues of LAS RPVs are also included under this subtopic. The highest ranked specific interest area under this subtopic is Validate embrittlement trend curve models (obtaining high fluence materials), with an average ranking of 3.8. EPRI noted that it agrees with the importance of this specific interest area but has existing programs to address and does not see the need for harvesting to address. NRC also indicated it believes that surveillance specimens would be most effective for validating ETCs, rather than actual RPV materials. Another relatively highly ranked specific interest area, with an average ranking of 3.4, is Near-inner surface including cladding materials (i.e., for evaluation of residual stress and shallow flaw effects). Comments from respondents (EPRI, NRC) pointed to a lack of data for flaws in this region, along with a desire to validate or refute the existence of shallow surface-breaking flaws.
CASS and SS Welds This topic involves the effects of thermal aging, irradiation, or both on cast austenitic stainless steel components, and austenitic stainless steel weld metals which have a similar duplex structure. The highest average ranking under this subtopic is for the specific interest area Evaluate thermal aging effects on fracture toughness of SS weld components (i.e., validation of lab. testing and models), with an average ranking of 3.3. Respondents which ranked this area high indicated a desire for actual component data and a lack of data for this specific interest area. The next highest ranked specific interest area under this subtopic is Evaluate thermal aging and irradiation effects on fracture toughness of CASS components (i.e., validation of lab. testing and models), with an average ranking of 2.7. However, there is less interest in this specific interest area than the similar one related to stainless steel weld material, even though both concern the synergistic effect of thermal aging and irradiation on fracture toughness of duplex stainless steel materials. This diminished interest is likely because of the large body of research associated with CASS embrittlement over the last several decades.
This topic focuses on steam generator tube degradation and defects in Alloy 600 and Alloy 690 material.
This area had the lowest overall average rankings, with the highest ranked specific interest area, Evaluate SCC in SG Tubes, having an average ranking of 2.5. For those ranking this specific interest H, KAERI/KHNP indicated that they are interested in evaluation of SCC initiation in actual components, while IRSN indicated interest in the effects of pollutants on SCC. However, NRC ranked this specific interest low, noting that it is a very broad area, and there is already quite a bit of data.
Reactor-Specific This category includes components specific to CANDU and VVER reactor designs. The specific interest areas under these topics had low average rankings, most likely because these reactor designs are only used in specific countries, thus specific interest areas are not applicable to reactor types such as BWRs and PWRs that are used in many countries.
23
CANDU The highest ranked specific interest area under this topic was Calandria vessel - 304SS at lower temperatures withvery little data. EPRI ranked this area H, and commented with respect to the Calandria vessel that it would be of high value to characterize calandria tubesheet welds under end of life fluence condition (~90 years life), so this would only be applicable with a long-service, decommissioned reactor.
EPRI further commented that interim data points from boat sampling or intermediate life reactors could potentially be helpful. CNSC also ranked the Calandria vessel H because data is needed to support extended operation. CNSC also ranked the other CANDU related specific interest areas H to support extended operation, with the exception of Investigate helium embrittlement in Ni alloys, which it ranked M.
VVER For the VVER design, the only specific interest areas is horizontal steam generators which was ranked medium by UJV, with the comment that there is ongoing periodical SG cleaning; and that examination of the harvested SG from EBO V1 is ongoing. With respect to horizontal SGs, EPRI also noted that for VVER SMEs - there is increasing significance of sludge levels on DMWs (cracking), and that tubes are 321SS welded to a vertical collector pipe (CS), with unknown alloy DMWs. EPRI further noted that horizontal SGs have increased priority in the revised MDM.
Harvesting Activities The discussion then consisted of reviewing the information which had been collected after the 2020 harvesting workshop on harvesting activities in participating countries. It was discussed that there is clearly an interest to continue to collect prioritization information to be used to identify future collaborative research opportunities. In particular, the table could be used to identify if material to address priority topics has been harvested and is available for collaborative research. The following action was noted:
Action HWS2022-4: Workshop participants to update their countrys activities within the harvesting activities table to support identifying collaborative research opportunities.
The importance of completing the table was highlighted as it would constitute an important first step to get a view of harvested material and research capabilities which are already available in countries leading harvesting activities. This would also allow identifying what harvested material and related laboratory research could be proposed for collaborative research. The information updated after 2022 harvesting workshop is provided in the tables in Appendices D and E. A further valuable step would be, as was already discussed at the 2020 workshop, to address the development of a harvested metal material library providing an overview of possibilities for collaborative research.
Regarding the workshop summary report, it was said the objective would be to establish a first version of the report about six months after the workshop, providing participants would give the expected feedback in this period of time. The following action was noted:
Action HWS2022-5: NEA Secretariat and USNRC to provide workshop participants with a draft summary report of the workshop about six months after the workshop.
24
A draft report was distributed to the workshop participants before this final version with comments included was completed.
25
Key Takeaways from the Workshop
Increasing opportunities exist for harvesting and performing collaborative research on aged metallic materials
A number of nuclear power plants have recently been shut down worldwide (retirement after long term operation or closure) and are being decommissioned, offering new, unique opportunities for harvesting material aged under actual operating conditions, including material with long operation histories, corresponding to up to 30 years and beyond of full operation.
As evidenced by presentations at the workshop of on -going harvesting activities, related research and the development of supporting research capabilities in a number of countries, opportunities for collaborative research on harvested material are increasing. In a number of countries, harvesting has been conducted, and plans for further harvesting are being considered, with material with diverse interest available in research laboratories for investigations. As discussed at the workshop, conducting an exercise to compile in a library of currently available harvested materials for collaborative research would be valuable for developing future joint projects proposals. The library would include materials which can be offered for international research with no or reduced export control limitations on associated information and data and with good historical records on component fabrication; initial material properties; environmental conditions during operation ; aging management programs; and any associated inspection, repair and maintenance activities.
Harvesting can yield significant benefits for the nuclear sector but requires diverse and extensive stakeholder involvement
Harvesting material in plants is unparalleled with respect to providing prototypical materials (when compared to material artificially aged in laboratories) for use in developing ageing management strategies related to support extended reactor operation to 60 years and beyond. However, several challenging aspects exist, including complex and delicate on-site retrieval operations that can affect decommissioning schedules, transporting harvested radiological materials to external research laboratories, and marshalling the resources needed to assemble detailed information on harvested components (i.e.,
fabrication records, initial material properties, environmental conditions, aging management programs, and any associated inspection, repair and maintenance activities), make harvesting a challenging and costly proposition that requires strong involvement and close cooperation of all stakeholders (i.e.,
regulators, researchers, utilities, decommissioning companies).
Workshop participants highlighted the challenge to get utilities and decommissioning companies involved in, and supportive of, harvesting activities, particularly when there are no direct benefits for those organizations. Incentives for the private nuclear sector in general are that enhanced knowledge of material ageing in real operating conditions should, for operating reactors, support optimization of operating conditions, of inspections and maintenance plans, of repair and replacement activities, all of which should ultimately support LTO. In addition, knowledge gained is also expected to be of value for supporting the development of advanced light-water reactors with planned operating life time of 60 years and beyond. All parties engaged in the development of the nuclear sector should be made more aware of the potential benefits from harvesting and related research both for operating reactors and reactors under development.
26
Policies within countries could more clearly endorse harvesting as an important, technically-robust approach to materials characterization for long-term operation and consider how to engage more utilities in harvesting activities for the benefit of long term operation and nuclear safety. Relevant standing NEA committees, CNRA (Committee on Nuclear Regulatory Activities) and CSNI (Committee on the Safety of Nuclear Installations), could engage further in actions to promote harvesting research activities. Currently, there are no regulatory requirements that require harvesting.
The high-complexity and cost of material harvesting and related research activities make them prime candidates for collaboration
The OECD/NEA SMILE joint project conducted by Studsvik in Sweden provides a first of a kind large collaborative effort around aged metal material harvested in several decommissioning Swedish plants. In Sweden, utilities and regulators involved in plant decommissioning have worked jointly to support in harvesting and research of aged materials, in conjunction with organizations from other countries in SMILE. This project offers a unique opportunity for cooperation among different stakeholders (e.g.,
regulators, utilities, research organizations, decommissioning companies) in a number of countries to address challenges related to harvesting material on-site, such as performing on-site measurements; collecting historical information and data on harvested material; calculating environmental loads on materials during their in-service life; and collecting and preparing the most prototypic samples for further characterization and testing. It is also a unique opportunity to share experience and good practices and optimize characterization and testing methods.
Collaborative efforts could be continued and extended in the future beyond the current SMILE project, considering material of interest harvested in other countries for the interest to a wider range of countries/organizations. SMILEs access to a number of reactors at various stages of decommissioning and dismantling already provides for a large stream of samples and test results from various source materials that are applicable to BWRs, PWRs or VVERs but other sources of material would be beneficial to extend the knowledge base.
Benefits of collaborative research associated with harvesting are numerous: information sharing, sharing research costs, sharing material sources, sharing best practices in samples preparation, testing and characterization, establishing commonly shared knowledge base for models development and validation, etc.
Prioritizing research activities for maximizing benefits and addressing the widest range of interest should be continued Participants agreed on the need to continue and deepen research prioritization and identification of opportunities to collaborative research. The proposed approach, initiated at the 2020 international harvesting workshop to identify priority topics and materials of interest in relation to LTO, was continued and refined with inputs from all organizations interested in LTO. The results of this exercise was summarized in Part 3. This comprehensive priorities list should help countries involved in harvesting to target materials to harvest and help identify if existing harvested materials could address priority topics.
A next step after prioritizing research interests could be to identify plant components to be used in testing programs to generate data to address the research knowledge gaps. The identification of a ppropriate
27
components is needed to evaluate the applicability of currently-available materials or upcoming decommissioning opportunities1.
In parallel with the component identification and the plant identification steps, it would be beneficial, as earlier discussed, to develop a catalogue/library of available, previously-harvested materials and components. Matching the research priorities to necessary material conditions (e.g., service temperature, fluence) and further to the available components from decommissioning plants should help focus resources on identifying and harvesting those components with the most value.
Similar initiatives should be launched for other categories of materials (concrete, polymers incl. cables material)
Participants have expressed an interest in extending information sharing and prioritization efforts to other materials including concrete and the polymeric materials used in electrical cable insulation. It has been suggested that organizing future similar international workshops on harvesting materials other than metals would be valuable. OECD/NEA could support such initiatives.
1 As introduced at the 2020 international Harvesting workshop, such an effort was initiated by EPRI in collaboration with plant owners. EPRI intends to continue to identify and collect information from near-term decommissioning plants that could be used to address the prioritized research interests and will collate the gathered information into a simple database.
28
References to Previous Harvested Materials Research Below is a list of references to research results generated from testing of harvested materials:
- 1. J.R. Hawthorne and A.L. Hiser, Experimental Assessments of Gundremmingen RPV Archive Material for Fluence Rate Effects Studies, NUREG/CR-5201 (MEA-2286), U.S. Nuclear Regulatory Commission, October 1988.
- 2. Kurtz, R. J., et al. 1990. Steam Generator Integrated Program/Steam Generator Group Project, Final Project Summary Report, NUREG/CR-5117, PNL-6226. Prepared for the U.S. Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory, Richland, Washington.
- 3. O.K. Chopra, and W.J. Shack, Mechanical Properties of Thermally Aged Cast Stainless Steels from Shippingport Reactor Components, NUREG/CR-6275 (ANL-94/37), U.S. Nuclear Regulatory Commission, April 1995.
- 4. G. J. Schuster, S. R. Doctor, S.L. Crawford, and A. F. Pardini, Characterization of Flaws in U.S. Reactor Pressure Vessels: Density and Distribution of Flaw Indications in the Shoreham Vessel, NUREG/CR-6471 Volume 3, U.S. Nuclear Regulatory Commission, November 1999.
- 5. G. J. Schuster, S. R. Doctor, A.F. Pardini, and S.L. Crawford, Characterization of Flaws in U.S. Reactor Pressure Vessels: Validation of Flaw Density and Distribution in the Weld Metal of the PVRUF Vessel,
NUREG/CR-6471 Volume 2, U.S. Nuclear Regulatory Commission, August 2000.
- 6. D.E. McCabe, et al. Evaluation of WF-70 Weld Metal From the Midland Unit 1 Reactor Vessel, NUREG/CR-5736 (ORNL/TM-13748), U.S. Nuclear Regulatory Commission, November 2000.
- 7. A.B. Johnson, Jr., S.K. Sundaram, F.A. Garner, Program Plan for Acquiring and Examining Naturally Aged Materials and Components for Nuclear Reactors, PNNL -13930, Pacific Northwest National Laboratory, December 2001. (Prepared for the U.S. Department of Energy).
- 8. B. Alexandreanu, O.K. Chopra, and W.J. Shack, Crack Growth Rates in a PWR Environment of Nickel Alloys from the Davis-Besse and V.C. Summer Power Plants, NUREG/CR-6921 (ANL-05/55), U.S. Nuclear Regulatory Commission, November 2006.
- 9. S.E. Cumblidge, et al. Nondestructive and Destructive Examination Studies on Removed-from-Service Control Rod Drive Mechanism Penetrations, NUREG/CR-6996, U.S. Nuclear Regulatory Commission, July 2009.
- 10. S.E. Cumblidge, et al. Evaluation of Ultrasonic Time-of-Flight Diffraction Data for Selected Control Rod Drive Nozzles from Davis Besse Nuclear Power Plant, PNNL-19362, Pacific Northwest National Laboratory, April 2011.
- 11. S.L. Crawford, et al. Ultrasonic Phased Array Assessment of the Interference Fit and Leak Path of the North Anna Unit 2 Control Rod Drive Mechanism Nozzle 63 with Destructive Validation, NUREG/CR-7142 (PNNL-21547), U.S. Nuclear Regulatory Commission, August 2012.
- 12. R. Fuentes, et al. Characterization and Analysis of Boral from the Zion Nuclear Power Plant Spent Fuel Pool, SRNL-TR-2018-00244, Rev. 0, Savannah River National Laboratory, March 2019 (ML19155A215).
29
- 13. P. Ramuhalli, et al. Criteria and Planning Guidance to Ex-Plant Harvesting to Support Subsequent License Renewal, PNNL-27120, Rev. 1, Pacific Northwest National Laboratory, March 2019 (ML19081A006).
- 14. Y. Chen, et al. Crack Growth Rate and Fracture Toughness Tests on Irradiated Ex-Plant Materials. ANL-19/45, Argonne National Laboratory, July 2020 (ML20198M503).
- 15. Materials Reliability Program: Zorita Internals Research Project (MRP-440), Testing of Highly-Irradiated Baffle Plate Material. EPRI, Palo Alto, CA: 2019. 3002016015
- 16. Materials Reliability Program: Fluence Effects on Stainless Steel Welds (MRP-451): Crack Growth Rate and Fracture Toughness Testing of Zorita Weld and HAZ Materials. EPRI, Palo Alto, CA: 2020.
3002018250.
- 17. Chen, Y., W-Y. Chen, and B. Alexandreanu, Irradiated Microstructure of Zorita Materials, ANL-20/50, Argonne National Laboratory, Lemont, IL, August 2020. (ADAMS Accession No. ML20269A143).
- 18. Kombaiah, B., C. Judge, J. Charboneau, S. Smith, L. Gimenes Rodrigues Albuquerque, and V. Montes de Oca Carioni, Chemical Compositional Analysis and Microstructural Characterization of Harvested Zorita Reactor Pressure Vessel (RPV) Internals, INL/EXT-21-62220, Idaho National Laboratory, Idaho Falls, ID, March 2021. (ADAMS Accession No. ML21124A112).
- 19. Jenssen, A. J. Stjrnster, K. Kese, R. Carter, J. Smith, A. Demma, and M. Hiser, Fracture Toughness Testing of an Irradiated PWR Core Barrel Weld, Fontevraud 9 Contribution of Materials Investigations and Operating Experience to LWRs Safety, Performance and Rel iability, Avignon, France, September 17-20, 2018.
- 20. Eason, E., and R. Pathania, Irradiation-Assisted Stress Corrosion Crack Growth Rates of Austenitic Stainless Steels in Light Water Reactor Environments, 17th International Conference on Environmental Degradation of Materials in Nuclear Power SystemsWater Reactors, Ottawa, Ontario, Canada, August 9-12, 2015.
- 21. Eason, E. D., and Pathania, R. Disposition Curves for Irradiation-Assisted Stress Corrosion Cracking of Austenitic Stainless Steels in Light Water Reactor Environments, Proc. ASME 2015 Pressure Vessels &
Piping Conference, Boston MA. Paper PVP2015 -45323, July 19-23, 2015.
- 22. Jenssen, A., J. Stjrnster, C. Topbasi, and P. Chou, Specimen Size Effects on the Crack Growth Rate Response of Highly Irradiated Type 304 Stainless Steel, 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Boston, MA, August 18-22, 2019.
- 23. Chen,Y., B. Alexandreanu, C. Xu, Y. Yang, K. Natesan, and A. S. Rao, Environmentally Assisted Cracking and Fracture Toughness of an Irradiated Stainless Steel Weld, 19th International Conference on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors, Boston, MA, August 18-22, 2019.
30
- 24. Rosseel T, M Sokolov and R Nanstad. 2016a. Report on the Harvesting and Acquisition of Zion Unit 1 Reactor Pressure Vessel Segments. ORNL/TM-2016/240. Oak Ridge, Tennessee: Oak Ridge National Laboratory.
- 25. International Metals Harvesting Workshop Summary Report, Workshop held on January 21-22, 2020 at NEA headquarters in Paris, France (no reference number, internal document)
31
Appendix A: Workshop Agenda
Agenda of the 2nd Workshop on International Harvesting Cooperation
November 17, 2022, Scandic Klara Hotel, Sljdgatan 7, Stockholm, Sweden
Stor-Stockholm Norra room, 1 floor down from lobby
9:00 am - 17:45 pm
Introductory part
9:00-9:20 Introductions/Opening Remarks /Participants presentation
Robert Tregoning, USNRC, Didier Jacquemain, NEA
9:20-9:40 2020 Harvesting Workshop Summary & Conclusions
9:40-10:00 Review of 2020 Harvesting Workshop Common Priorities
Presentations on Activities in Countries
10:00-10:25 Harvested RVI from EBO V1 Unit 2 (VVER 440 type)
Miroslava Ernestová, ÚJV ez
10:25-10:45 Break
10:45-11:10 Japanese Harvested Material Research sponsored by the Nuclear Regulatory Authority
Taku Arai, CRIEPI
11:10-11:35 SMILE - Ongoing Harvesting and Testing Activities
Martin Bjurman, Studsvik
11:35-12:00 US Harvesting Activities
Jeff Poehler, USNRC
12:00-13:30 Lunch
A-1
13:30-13:40 Short Information on Potential Harvesting Interests from Beznau NPP Reiner Mailnder, ENSI
13:40-14:05 EPRI Harvesting Priorities and Activities
Frank Gift, EPRI
14:05-14:30 Update on Materials Harvesting Activities in Korea
Sung-Woo Kim, KAERI
14:30-14:55 Harvesting Activities Update for USDOE NSUF Program
Peng Xu, INL
14:55-15:10 Break
Discussion on Priorities/Opportunities
15:10- 15:40 Updated Priority Discussion All
15:40-16:00 Harvesting Opportunities Review and Update NEA\\All
16:00-16:30 Discussion of Priorities vs. Activities/Opportunities NEA\\All
Discussion on Harvesting Collaborations and Information Sharing
16:30-16:55 The International Network on Life Management of Nuclear Power Plants (LMNPP Network) Ms. Qun Yu, IAEA
16:55-17:20 Final discussion All
17:20-17:40 Wrap-up/Action items/Summary report NEA\\All
17:40-17:45 Workshop Closure NEA\\USNRC
A-2
Appendix B: Workshop Participants
BELGIUM DELEDICQUE, Vincent E-mail: vincent.deledicque@belv.be Bel V VIRTUAL
UYTDENHOUWEN, Inge E-mail: inge.uytdenhouwen@sckcen.be SCK CEN (The Belgian Nuclear Research VIRTUAL Centre)
CANADA
WASILUK, Bogdan E-mail: bogdan.wasiluk@cnsc-ccsn.gc.ca Canadian Nuclear Safety Commission VIRTUAL
CHINA WANG, Chen E-mail: wangchen1@chinansc.cn NSC, Nuclear and Radiation Safety VIRTUAL Centre
ZHEN, Zeng E-mail: zengzhen@chinansc.cn NSC, Nuclear and Radiation Safety VIRTUAL Centre
CZECHIA ERNESTOVA, Miroslava E-mail: miroslava.ernestova@ujv.cz UJV Rez IN PERSON
RUSNAKOVA, Katerina E-mail: katerina.rusnakova@ujv.cz UJV Rez IN PERSON
FINLAND KARLSEN, Wade E-mail: wade.karlsen@vtt.fi VTT Technical Research Centre of IN PERSON Finland, Ltd.
VARPASUO, Pentti E. E-mail: pentti.varpasuo@afry.com AFRY VIRTUAL
B-1
FRANCE BARBIER, Gauzelin E-mail: gauzelin.barbier@irsn.fr IRSN, Institut de Radioprotection et de IN PERSON Sûreté Nucléaire
GERMANY JUENGERT, Anne E-mail: anne.juengert@mpa.uni-stuttgart.de MPA, University of Stuttgart IN PERSON
MOHR, Stefan E-mail: stefan.mohr@grs.de GRS gGmbH IN PERSON
SCHOPF, Tim E-mail: tim.schopf@mpa.uni-stuttgart.de MPA, University of Stuttgart VIRTUAL
JAPAN ARAI, Taku E-mail: arait@criepi.denken.or.jp CRIEPI IN PERSON
HASHIKURA, Yasuaki E-mail: hashikura_yasuaki_x3t@nra.go.jp NRA, Nuclear Regulation Authority VIRTUAL
E-mail: mizuta_kohei_8u6@nra.go.jp MIZUTA, Kohei IN PERSON NRA, Nuclear Regulation Authority
KOREA KIM, Jongmin E-mail: jmkim@kaeri.re.kr KAERI, Korea Atomic Energy Research VIRTUAL Institute
KIM, Sung Woo E-mail: kimsw@kaeri.re.kr KAERI, Korea Atomic Energy Research IN PERSON Institute
B-2
NA, Kyung-Hwan E-mail: kyunghwan.na@khnp.co.kr KHNP, Korea Hydro and Nuclear Power IN PERSON Company
SLOVENIA COSTA, Oriol E-mail: oriol.costa@ijs.si JSI, Jozef Stefan Institute VIRTUAL
SWEDEN
BJURMAN, Martin E-mail: martin.bjurman@studsvik.com Studsvik Nuclear AB IN PERSON
CALOTA, Elena E-mail: elena.calota@ssm.se SSM, Swedish Radiation Safety Authority IN PERSON
JENSSEN, Anders E-mail: anders.jenssen@studsvik.com Studsvik Nuclear AB IN PERSON
LIDBERG, Henric E-mail: henric.lidberg@vattenfall.com Vattenfall AB IN PERSON Ringhals
NYSTRAND, Lotta E-mail: lotta.nystrand@studsvik.com Studsvik Nuclear AB IN PERSON
WANG, Mi E-mail: mi.wang@studsvik.com Studsvik Nuclear AB IN PERSON
SWITZERLAND MAILNDER, Reiner E-mail: reiner.mailaender@ensi.ch ENSI, Swiss Federal Nuclear Safety IN PERSON Inspectorate
UNITED STATES CARTER, Robert E-mail: bcarter@epri.com EPRI, Electric Power Research Institute IN PERSON
B-3
CHEN, Xiang E-mail: chenx2@ornl.gov ORNL, Oak Ridge National laboratories VIRTUAL
CUNNINGHAM, Kelly E-mail: kelly.cunningham@inl.gov INL, Idaho National Laboratory VIRTUAL
GIFT, Frank E-mail: fgift@epri.com EPRI, Electric Power Research Institute IN PERSON
POEHLER, Jeffrey E-mail: jeffrey.poehler@nrc.gov US Nuclear Regulatory Commission IN PERSON
SMITH, Jean E-mail: jmsmith@epri.com EPRI, Electric Power Research Institute IN PERSON
TREGONING, Robert E-mail: robert.tregoning@nrc.gov US Nuclear Regulatory Commission IN PERSON
XU, Peng Nmn E-mail: peng.xu@inl.gov INL, Idaho National Laboratory IN PERSON
INTERNATIONAL ORGANISATIONS YU, Qun E-mail: q.yu@iaea.org IAEA, International Atomic Energy Agency VIRTUAL
MARTIN, Oliver E-mail: oliver.martin@ec.europa.eu European Commission - Joint Research VIRTUAL Centre
CHITOSE, Keiko E-mail: keiko.chitose@oecd-nea.org OECD Nuclear Energy Agency IN PERSON
JACQUEMAIN, Didier E-mail: didier.jacquemain@oecd-nea.org OECD Nuclear Energy Agency IN PERSON
B-4
Appendix C: Harvesting Priority Topics from the 2020 International Harvesting Cooperation Workshop Participants
Category Specific Interest Priority Comment NDE procedures and advances / developments BFB removal torque Inspection Thermal shield bolts / flexures and NDE Core barrel/shroud welds Activities High usage fatigue locations (including environmental factors)
Optimal (e.g., flawed) locations / components for harvesting Thermal aging Thermal aging with neutron embrittlement (also flux / spectrum effects)
RPV nozzle - low flux effects RPV LAS weld Through -wall fracture toughness and fluence and base (validate calculations) metal Embrittlement trend curve validation Untested surveillance specimens Orientation effects and material variability Small specimen techniques (with broken Charpys)
High dose IASCC initiation and growth stainless steel Void swelling (>30 dpa, >330C)
(SS) vessel SS Welds (>2 dpa) internal Baffle bolts materials CASS and SS Thermal aging of CASS and SS welds Welds Thermal aging and irradiation of CASS and SS weld DMW residual stress profiles PWSCC Flawed components Mitigation effectiveness (water chemistry, peening, MSIP) 690/52/152 alloys - CGR testing
C-1
Long-range ordering in Alloy 690 (pressurizer nozzle welds) 600/82/182 alloys-CGR testing SS DMWs Shallow flaws in Alloy 600 for NDE Steam Wear in Alloy 690 generators Long-range ordering in Alloy 690 Divider plate Wear / Guide cards Fretting CRDM thermal sleeves Alloy 690 SG tubes Environmental Validations of models and design practices Fatigue Irradiation effects Pressure tubes CANDU Topics Feeder tubes Calandria vessel - 304SS at lower temperatures with very little data VVER Topics Horizontal SGs
C-2
Appendix D: Recent and Active Harvesting Activities Country Plant Design Size (MWe) Years in operation Components Status Organization(s)
Canada NPD CANDU 20 25 Concrete AECL Gentilly-2 CANDU-6 675 29 Cables RPV, concrete Harvesting completed CRIEPI, Chubu in 2018 Japan Hamaoka 1 BWR-4 540 33 Internals,Primary Harvesting completed Loop Recirculation in 2029 NRA,CRIEPI Pump Harvesting completed Spain Zorita W 1-loop 160 37 Internals in 2013; DECON nearly EPRI, NRC complete Barseback ABB-II 615 28 RPV Vattenfall Sweden Oskarshamn1 ABB-I 473 45 Completed Oskarshamn2 ABB-II 638 39 Studsvik Ringhals 2 W 3-loop 900 44 RPV, cables, neutron Harvesting complete; Zion 1/2 W - 4 loop 1040 24/25 absorbers, electrical DECON finished by DOE, EPRI, NRC bus duct 2020 Crystal River B&W 860 36 Cables Harvesting complete; EPRI, NRC 3 Plant in SAFSTOR U.S. Ginna W 2-loop 580 42 (when baffle bolts Baffle bolts DOE removed)
Indian Point W 4-loop 1020 43 (when baffle bolts Baffle bolts Plant still operating EPRI, DOE removed)
DC Cook W 4-loop 1168 39 (when baffle bolts Baffle bolts EPRI, DOE removed)
D-1
Country Plant Design Size (MWe) Years in operation Components Status Organization(s)
Core shroud including baffle bolts, core Czechia/ EBO V1 Unit VVER 440 27 barrel, protective tube Already harvested UJV, VUJE Slovakia 2 440/230 unit Other SSCs (decontaminated)
D-2
Appendix E Planned and Potential Harvesting Activities
Potential component information is here
E-1
Appendix F: 2022 International Workshop Harvesting Priorities
Responses from each participants
F-1
2022 International Workshop Harvesting Priorities (Bel V / Belgium)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress profiles L Flawed components H Evaluate mitigation effectiveness (water L chemistry, peening, MSIP)
Evaluate 690/52/152 properties (i.e., CGR H testing)
PWSCC Investigate embrittlement effects in high Cr L Ni alloys (690 systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties (i.e., CGR L Aging testing)
Mechanism SS DMW components H s Alloy X750 component L Environmenta Validate models and design practices H l Fatigue Evaluate irradiation effects (indicate if L interest is reactor-type specific)
Guide cards L Wear / CRDM thermal sleeves L Fretting Other high-wear components (please specify L component of interest)
Aging effects Consider interaction between fabrication L on fabrication defects and service-induced aging (e.g., weld defects lack of fusion and PWSCC, RPV embrittlement and hydrogen flaking)
Effectiveness Harvest repaired components (i.e., peened, M of repair overlayed, mechanical stress improvement techniques for evaluation and testing of long-term repair effectiveness Evaluate existing and emerging NDE H procedures and advances /developments BFB removal torque L Mitigation Thermal shield bolts/flexures M and Core barrel/shroud welds L Inspection Inspection Inspect high usage fatigue locations (incl. H and NDE environmental factors)
Activities Conduct NDE before harvesting to identify H optimal (e.g., flawed) components/locations for harvesting Evaluate flaw distributions, types, and H characteristics (e.g., fabrication or service induced & mechanism) in critical components/welds
F-2
Mitigation Inspection Assess degradation in CRDM tubes, thermal L and and NDE sleeves, and other locations (e.g., cracking, Inspection Activities wear)
Evaluate IASCC initiation and growth, H especially for high-fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 dpa, >330C) H RPV Internals Assess irradiation assisted creep M SS Welds (>2 dpa), i.e., for CGR and Fracture M Toughness (FT) testing Baffle bolts including IASCC initiation and L crack growth mechanisms Evaluate thermal aging H Evaluate thermal aging and irradiation M embrittlement (also flux /spectrum effects)
RPV nozzle - evaluate low flux, cavity L streaming, and temperature effects Measure Through-wall fracture toughness H properties Measure through-wall fluence values (i.e., to L validate calculations)
Validate embrittlement trend curve models H (obtaining high fluence materials)
Component RPV LAS weld Validate embrittlement trend curve models H
-Specific and base (obtaining low flux/temperature materials) metal Evaluate untested surveillance specimens L Evaluate fabrication flaw distributions/types H (e.g., hydrogen flake cracking)
Assess material orientation and variability H effects including segregation (e.g., C segregation)
Small specimen techniques (from broken M Charpy surveillance specimens)
Near-inner surface including cladding M materials (i.e., for evaluation of residual stress and shallow flaw effects)
Reactor support components/material (i.e., L for PWR, VVER)
Evaluate thermal aging effects on fracture H toughness of CASS components (i.e.,
CASS and SS validation of lab. testing and models)
Welds Evaluate thermal aging and irradiation L effects on fracture toughness of CASS components (i.e., validation of lab. testing and models)
F-3
Evaluate thermal aging effects on fracture M toughness of SS weld components (i.e.,
CASS and SS validation of lab. testing and models)
Welds Evaluate thermal aging and irradiation M effects on fracture toughness of SS weld components (i.e., validation of lab. testing Component and models)
-Specific Search for shallow flaws in A600 tubes for L evaluating NDE effectiveness Steam Investigate wear in A690 SG tubes L generators Examine Long-range ordering in A690 L Evaluate SCC in SG tubes L Evaluate Divider plate degradation L Pressure tubes L Feeder tubes L Reactor-CANDU Calandria vessel - 304SS at lower L Specific temperatures w/ very little data Investigate helium embrittlement in Ni alloys L VVER Horizontal SGs L
Other Comments In our opinion, priorities should certainly be put on topics for which harvesting is the best (if not the unique) way to get some progress in our understanding. This is generally the case for components which are considered to be irreplaceable. Topics which could be addressed in other frameworks should receive a lower priority. For example, we considered that evaluating existing and emerging NDE procedures and advances /developments has a high priority, but if it is possible to address this elsewhere, given that harvesting activities ca be quite expensive and time consuming, we should do so to optimize our efforts and our available resources.
The table did not always allow to make the distinction on the type of components. The aim of the harvesting can indeed differ between components which are unique or others which are numerous.
For example, for large components (and for which harvesting will probably be possible on few components at most), we could wonder if evaluating the flaw distribution, types and characteristics presents an interest, given that one single component is not necessarily representative of the equivalent components in other NPPs (I believe a knowledge of the distribution of the hydrogen flakes in Tihange 2 for example will not present much interest for most other NPPs). On the other hand, for components which are numerous and on which a lot of harvesting activities would be possible, this analyses could make sense.
F-4
Similarly, the need to conduct NDE before harvesting to identify optimal (e.g., flawed) components/locations for harvesting depends on the aim of the analysis. Do we intend to obtain an overview of the average behavior of some component, or do we intend to gain knowledge of the worst case? Perhaps this needs a case by case discussion.
F-5
2022 International Workshop Harvesting Priorities (CNSC / Canada)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress M Possible impact for all reactor profiles designs Flawed components L Low relevance to Canadian CANDU reactors Evaluate mitigation M Possible impact for all reactor effectiveness (water chemistry, designs peening, MSIP)
Evaluate 690/52/152 properties L Low relevance to Canadian (i.e., CGR testing) CANDU reactors PWSCC Investigate embrittlement L Low relevance to Canadian effects in high Cr Ni alloys (690 CANDU reactors systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties L Low relevance to Canadian (i.e., CGR testing) CANDU reactors SS DMW components M Possible impact for all reactor Aging designs Mechanisms Alloy X750 component L Low relevance to Canadian CANDU reactors Validate models and design M Possible impact for all reactor Environmental practices designs Fatigue Evaluate irradiation effects H Potential application to 304L SS (indicate if interest is reactor-used in CANDU Calandria vessels type specific)
Guide cards L Low relevance to Canadian CANDU reactors Wear / CRDM thermal sleeves L Low relevance to Canadian Fretting CANDU reactors Other high-wear components L Low relevance to Canadian (please specify component of CANDU reactors interest)
Consider interaction between H Potential application to 304L SS Aging effects fabrication defects and service-used in CANDU Calandria vessels on fabrication induced aging (e.g., weld lack of (weld flaws, irradiation defects fusion and PWSCC, RPV embittlement, SCC) embrittlement and hydrogen flaking)
Harvest repaired components M Weld overlay for wall thinning Mitigation Effectiveness (i.e., peened, overlayed, and of repair mechanical stress improvement Inspection techniques for evaluation and testing of long-term repair effectiveness
F-6
Evaluate existing and emerging H Possible impact for all reactor NDE procedures and advances designs
/developments BFB removal torque L Low relevance to Canadian CANDU reactors Thermal shield bolts/flexures L Low relevance to Canadian CANDU reactors Core barrel/shroud welds L Not relevant to Canadian CANDU reactors Inspect high usage fatigue H Possible impact for all reactor locations (incl. environmental designs Mitigation Inspection factors) and and NDE Conduct NDE before harvesting H Can also be used to confirm Inspection Activities to identify optimal (e.g., flawed) accuracy of NDE techniques for components/locations for piping, SG tubes, etc.
harvesting Evaluate flaw distributions, M Possible impact for all reactor types, and characteristics (e.g., designs fabrication or service induced &
mechanism) in critical components/welds Assess degradation in CRDM M Possible impact for all reactor tubes, thermal sleeves, and designs other locations (e.g., cracking, wear)
Evaluate IASCC initiation and L Not relevant to Canadian CANDU growth, especially for high-reactors fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 L Not relevant to Canadian CANDU dpa, >330C) reactors Assess irradiation assisted creep L Not relevant to Canadian CANDU RPV Internals reactors SS Welds (>2 dpa), i.e., for CGR L Not relevant to Canadian CANDU and Fracture Toughness (FT) reactors Component-testing Specific Baffle bolts including IASCC L Not relevant to Canadian CANDU initiation and crack growth reactors mechanisms Evaluate thermal aging L Not relevant to Canadian CANDU reactors RPV LAS weld Evaluate thermal aging and L Not relevant to Canadian CANDU and base irradiation embrittlement (also reactors metal flux /spectrum effects)
RPV nozzle - evaluate low flux, L Not relevant to Canadian CANDU cavity streaming, and reactors temperature effects
F-7
Measure Through-wall fracture L Not relevant to Canadian CANDU toughness properties reactors Measure through-wall fluence L Not relevant to Canadian CANDU values (i.e., to validate reactors calculations)
Validate embrittlement trend L Not relevant to Canadian CANDU curve models (obtaining high reactors fluence materials)
Validate embrittlement trend L Not relevant to Canadian CANDU curve models (obtaining low reactors flux/temperature materials)
Evaluate untested surveillance L Not relevant to Canadian CANDU specimens reactors RPV LAS weld Evaluate fabrication flaw L Not relevant to Canadian CANDU and base distributions/types (e.g., reactors metal hydrogen flake cracking)
Assess material orientation and L Not relevant to Canadian CANDU variability effects including reactors segregation (e.g., C segregation)
Small specimen techniques M May have applications that can (from broken Charpy be applied to more than RPVs surveillance specimens)
Component-Near-inner surface including L Not relevant to Canadian CANDU Specific cladding materials (i.e., for reactors evaluation of residual stress and shallow flaw effects)
Reactor support L Not relevant to Canadian CANDU components/material (i.e., for reactors PWR, VVER)
Evaluate thermal aging effects on M Possible impact for all reactor fracture toughness of CASS designs components (i.e., validation of lab. testing and models)
Evaluate thermal aging and M Possible impact for all reactor irradiation effects on fracture designs toughness of CASS components (i.e., validation of lab. testing and CASS and SS models)
Welds Evaluate thermal aging effects on M Possible impact for all reactor fracture toughness of SS weld designs components (i.e., validation of lab. testing and models)
Evaluate thermal aging and H Possible impact for all reactor irradiation effects on fracture designs toughness of SS weld components (i.e., validation of lab. testing and models)
F-8
Search for shallow flaws in A600 M Possible impact on all SGs tubes for evaluating NDE effectiveness Investigate wear in A690 SG L Not relevant to Canadian CANDU Component-Steam tubes reactors Specific generators Examine Long-range ordering in L Not relevant to Canadian CANDU A690 reactors Evaluate SCC in SG tubes H Possible impact on all SGs Evaluate Divider plate H Possible impact on all SGs degradation Pressure tubes H Support extended operation Feeder tubes H Support extended operation Calandria vessel - 304SS at H Support extended operation Reactor-CANDU lower temperatures w/ very Specific little data Investigate helium M Support extended operation embrittlement in Ni alloys VVER Horizontal SGs L Not relevant to Canadian CANDU reactors
F-9
2022 International Workshop Harvesting Priorities (UJV / Czechia)
- empty cells - priority lower L (no VVER gap, different design or material in VVER NPPs)
Category Topic/Item Specific Interest Priority* Comment Measure DMW residual stress profiles M used methods, their sensitivity Flawed components Evaluate mitigation effectiveness (water chemistry, peening, MSIP)
Evaluate 690/52/152 properties (i.e., CGR testing)
PWSCC Investigate embrittlement effects in high Cr Ni alloys (690 systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties (i.e., CGR Aging testing)
Mechanisms SS DMW components M Alloy X750 component Environment Validate models and design practices M al Fatigue Evaluate irradiation effects (indicate if M PWR in general (VVER interest is reactor-type specific) is prefered)
Guide cards Wear / CRDM thermal sleeves Fretting Other high-wear components (please specify component of interest)
Aging effects Consider interaction between fabrication on defects and service-induced aging (e.g.,
fabrication weld lack of fusion and PWSCC, RPV defects embrittlement and hydrogen flaking)
Effectiveness Harvest repaired components (i.e.,
of repair peened, overlayed, mechanical stress techniques improvement for evaluation and testing of long-term repair effectiveness Evaluate existing and emerging NDE procedures and advances /developments Mitigation BFB removal torque H stress relaxation with and operation Inspection Inspection Thermal shield bolts/flexures and NDE Core barrel/shroud welds Activities Inspect high usage fatigue locations (incl. M environmental factors)
Conduct NDE before harvesting to identify optimal (e.g., flawed) components/locations for harvesting
F-10
Evaluate flaw distributions, types, and M characteristics (e.g., fabrication or service Mitigation Inspection induced & mechanism) in critical and and NDE components/welds Inspection Activities Assess degradation in CRDM tubes, thermal sleeves, and other locations (e.g.,
cracking, wear)
Evaluate IASCC initiation and growth, H VVER IMT gap (LTO especially for high-fluence (e.g., > 50 dpa) assessment) materials Investigate void swelling (>30 dpa, >330C) H VVER IMT gap (LTO assessment)
RPV Internals Assess irradiation assisted creep H VVER IMT gap (LTO assessment)
SS Welds (>2 dpa), i.e., for CGR and H VVER IMT gap (LTO Fracture Toughness (FT) testing assessment)
Baffle bolts including IASCC initiation and H VVER IMT gap (LTO crack growth mechanisms assessment)
Evaluate thermal aging L for VVER RPV material Evaluate thermal aging and irradiation L for VVER RPV material embrittlement (also flux /spectrum effects)
RPV nozzle - evaluate low flux, cavity streaming, and temperature effects Measure Through-wall fracture toughness M Component-properties Specific Measure through-wall fluence values (i.e., M to validate calculations)
Validate embrittlement trend curve M models (obtaining high fluence materials)
RPV LAS Validate embrittlement trend curve weld and models (obtaining low flux/temperature base metal materials)
Evaluate untested surveillance specimens M Evaluate fabrication flaw M distributions/types (e.g., hydrogen flake cracking)
Assess material orientation and variability effects including segregation (e.g., C segregation)
Small specimen techniques (from broken M Charpy surveillance specimens)
Near-inner surface including cladding M materials (i.e., for evaluation of residual stress and shallow flaw effects)
F-11
RPV LAS Reactor support components/material M weld and (i.e., for PWR, VVER) base metal Evaluate thermal aging effects on fracture toughness of CASS components (i.e.,
validation of lab. testing and models)
Evaluate thermal aging and irradiation effects on fracture toughness of CASS components (i.e., validation of lab. testing CASS and SS and models)
Welds Evaluate thermal aging effects on fracture Component-toughness of SS weld components (i.e.,
Specific validation of lab. testing and models)
Evaluate thermal aging and irradiation effects on fracture toughness of SS weld components (i.e., validation of lab. testing and models)
Search for shallow flaws in A600 tubes for evaluating NDE effectiveness Steam Investigate wear in A690 SG tubes generators Examine Long-range ordering in A690 Evaluate SCC in SG tubes Evaluate Divider plate degradation Pressure tubes Feeder tubes CANDU Calandria vessel - 304SS at lower temperatures w/ very little data Reactor-Investigate helium embrittlement in Ni Specific alloys Horizontal SGs M ongoing periodical SG VVER cleaning; examination of harvested SG EBO V1
F-12
2022 International Workshop Harvesting Priorities (STUK&VTT / Finland)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress profiles L Flawed components L inaccurate Evaluate mitigation effectiveness M combine with row 21?
(water chemistry, peening, MSIP)
Evaluate 690/52/152 properties (i.e., H CGR testing)
PWSCC Investigate embrittlement effects in L high Cr Ni alloys (690 systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties (i.e., L CGR testing)
Aging SS DMW components L please specify Mechanisms Alloy X750 component L Environmental Validate models and design practices M Fatigue Evaluate irradiation effects (indicate if M interest is reactor-type specific)
Guide cards L Wear / CRDM thermal sleeves L Fretting Other high-wear components (please L specify component of interest)
Aging effects Consider interaction between M For welds on fabrication defects and service-fabrication induced aging (e.g., weld lack of fusion defects and PWSCC, RPV embrittlement and hydrogen flaking)
Harvest repaired components (i.e., H Maybe reason to peened, overlayed, mechanical) stress concentrate on certain improvement for evaluation and failure cases (1 or 2?)
testing of long-term repair considering this list:
Residual effectiveness - Damage mehanism: SCC, stress profile fatigue.
Mitigation moidification - Plant phases: construction, and (Effectiveness repair.
Inspection of repair - Material: SS, LAS, 82/182, techniques) other Ni alloys.
- Environment: plant water, dry.
Proposal: Harvest weld residual stress improvement (i.e., induction treatment, weld overlay, mechanical, peening) for evaluation and testing of long-term effectiveness.
F-13
Evaluate existing and emerging NDE M procedures and advances
/developments BFB removal torque L Thermal shield bolts/flexures L Core barrel/shroud welds M Inspect high usage fatigue locations L Mitigation Inspection (incl. environmental factors) and and NDE Conduct NDE before harvesting to L Inspection Activities identify optimal (e.g., flawed) components/locations for harvesting Evaluate flaw distributions, types, and M characteristics (e.g., fabrication or service induced & mechanism) in critical components/welds Assess degradation in CRDM tubes, L thermal sleeves, and other locations (e.g., cracking, wear)
Evaluate IASCC initiation and growth, H especially for high-fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 dpa, L RPV >330C)
Internals Assess irradiation assisted creep M SS Welds (>2 dpa), i.e., for CGR and M Fracture Toughness (FT) testing Baffle bolts including IASCC initiation L and crack growth mechanisms Evaluate thermal aging L Evaluate thermal aging and irradiation M embrittlement (also flux /spectrum Component-effects)
Specific RPV nozzle - evaluate low flux, cavity M streaming, and temperature effects Measure Through-wall fracture M RPV LAS toughness properties weld and Measure through-wall fluence values L base metal (i.e., to validate calculations)
Validate embrittlement trend curve H models (obtaining high fluence materials)
Validate embrittlement trend curve L models (obtaining low flux/temperature materials)
Evaluate untested surveillance L Specify specimens
F-14
Evaluate fabrication flaw L distributions/types (e.g., hydrogen flake cracking)
Assess material orientation and L variability effects including segregation RPV LAS (e.g., C segregation) weld and Small specimen techniques (from M base metal broken Charpy surveillance specimens)
Near-inner surface including cladding M materials (i.e., for evaluation of residual stress and shallow flaw effects)
Reactor support components/material L (i.e., for PWR, VVER)
Evaluate thermal aging effects on L fracture toughness of CASS components (i.e., validation of lab.
Component-testing and models)
Specific Evaluate thermal aging and irradiation L effects on fracture toughness of CASS components (i.e., validation of lab.
CASS and SS testing and models)
Welds Evaluate thermal aging effects on M Does this mean RPV fracture toughness of SS weld cladding?
components (i.e., validation of lab.
testing and models)
Evaluate thermal aging and irradiation M Does this mean RPV effects on fracture toughness of SS cladding?
weld components (i.e., validation of lab. testing and models)
Search for shallow flaws in A600 tubes L for evaluating NDE effectiveness Steam Investigate wear in A690 SG tubes L generators Examine Long-range ordering in A690 L Evaluate SCC in SG tubes L Evaluate Divider plate degradation L Pressure tubes L Feeder tubes L Reactor-CANDU Calandria vessel - 304SS at lower L Specific temperatures w/ very little data Investigate helium embrittlement in Ni L alloys VVER Horizontal SGs L
F-15
2022 International Workshop Harvesting Priorities (IRSN / France)
Category Topic/Item Specific Interest ranking comment Measure DMW L / H if portable and (dissimilar metal welds) non destructive residual stress profiles means
Flawed components M (target to be explained)
Evaluate mitigation M / H on SS effectiveness (water chemistry, peening, MSIP)
Evaluate 690/52/152 L / M on DMW PWSCC properties (i.e., CGR - configurations crack growth rates-testing)
Investigate L embrittlement effects in high Cr Ni alloys (690 Aging systems) at higher Mechanisms temperatures(e.g.,
pressurizer nozzle welds)
Evaluate 600/82/182 L 600 remplaced by 690 properties (i.e., CGR testing)
SS DMW components H even on homogeneous SS welds Alloy X750 component L X750 in which configuration ?
Validate models and M, H depending on Prioritize the zones design practices potential future which have the most degradations important usage factors, identify the Environmental cracking zones. Access Fatigue to cracks with high usage factors.
Evaluate irradiation L on shell behaviour, effects (indicate if M in case of radiolysis interest is reactor-type playing part in specific). localized chemistry Wear / Guide cards L on 304L (french NPP),
Fretting controls seems ok
F-16
CRDM thermal sleeves L Wear / Other high-wear L Fretting components (please specify component of interest)
Aging Consider interaction interaction between Mechanisms between fabrication fabrication defects Aging effects defects and service-and service-induced on fabrication induced aging (e.g., weld aging : M for inner defects lack of fusion and surface defects PWSCC, RPV shell embrittlement :
embrittlement and L hydrogen flaking) flaking : L Harvest repaired H components (i.e.,
Effectiveness peened, overlayed, of repair mechanical stress techniques improvement for evaluation and testing of long-term repair effectiveness Evaluate existing and H emerging NDE procedures and advances
/developments BFB removal torque L Thermal shield L bolts/flexures Mitigation Core barrel/shroud welds L concerns BWR and Inspect high usage fatigue M see E15 Inspection locations (incl.
environmental factors)
Inspection Conduct NDE before H and NDE harvesting to identify optimal (e.g., flawed)
Activities components/locations for harvesting Evaluate flaw ? needs explanation distributions, types, and characteristics (e.g.,
fabrication or service induced & mechanism) in critical components/welds Assess degradation in L (content needs CRDM tubes, thermal explanation) sleeves, and other locations (e.g., cracking, wear)
F-17
Evaluate IASCC initiation M (radiation Remark for all RPV and growth, especially quantification internals topics: 1) for high-fluence (e.g., > probably based on radiation inside reactor 50 dpa) materials pure simulation, if -> indirect dpa prioritized may validation for French contribute to indirect PWR, 2) different type dpa validation for irradiation rise the French PWR) question of representativity RPV Internals Investigate void swelling M / H if tackling EPR same E30
(>30 dpa, >330C) conditions Assess irradiation L same E30 assisted creep SS Welds (>2 dpa), i.e., L same E30 for CGR and Fracture Toughness (FT) testing Baffle bolts including M same E30 IASCC initiation and crack growth mechanisms Evaluate thermal aging M / H in PZR H Component-conditions Specific Evaluate thermal aging H H for flux and spectrum and irradiation effects embrittlement (also flux
/spectrum effects)
RPV nozzle - evaluate M on EDF Fleet low flux, cavity conditions / L streaming, and otherwise temperature effects RPV LAS weld Measure Through-wall L / H tenacity /
and base fracture toughness microstructure / fluence metal properties link Measure through-wall M / H H since the fluence on fluence values (i.e., to the vessel is based on validate calculations) calcuations, the validation seems of importance if transposable to French PWR Validate embrittlement M / H of interest if the trend curve models irradiation history is (obtaining high fluence well known and several materials) cases are studied
F-18
Validate embrittlement L trend curve models (obtaining low flux/temperature materials)
Evaluate untested M to clarify surveillance specimens Evaluate fabrication flaw L distributions/types (e.g.,
RPV LAS weld hydrogen flake cracking) and base Assess material L topic already metal orientation and documented variability effects including segregation (e.g., C segregation)
Small specimen H Component-techniques (from broken Specific Charpy surveillance specimens)
Near-inner surface H including cladding materials (i.e., for evaluation of residual stress and shallow flaw effects)
Reactor support ? which support components/material components/materials (i.e., for PWR, VVER) ?
Evaluate thermal aging L effects on fracture toughness of CASS components (i.e.,
validation of lab. testing CASS and SS and models)
Welds Evaluate thermal aging L and irradiation effects on fracture toughness of CASS components (i.e.,
validation of lab. testing and models)
F-19
Evaluate thermal aging L on CASS / M on low to be seen on low effects on fracture ferrite SS --> compare ferrite stainless steel toughness of SS weld aged welds toughness welded joints. Not well components (i.e., properties vs. non-documented on 316 L validation of lab. testing aged welds CASS and SS and models)
Welds Evaluate thermal aging L and irradiation effects on fracture toughness of SS weld components (i.e.,
validation of lab. testing Component-and models)
Specific Search for shallow flaws L 600 remplaced by 690 in A600 tubes for evaluating NDE effectiveness Investigate wear in A690 M Steam SG tubes generators Examine Long-range M 600 remplaced by 690 ordering in A690 Evaluate SCC in SG tubes H effects of pollutants (secondary)
Evaluate Divider plate L degradation Pressure tubes L Feeder tubes L Calandria vessel - 304SS L CANDU at lower temperatures Reactor-w/ very little data Specific Investigate helium L embrittlement in Ni alloys VVER Horizontal SGs M interesting feedback (specific chemistry)
F-20
2022 International Workshop Harvesting Priorities (GRS&MPA / Germany)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress profiles M Flawed components H Evaluate mitigation effectiveness (water chemistry, H peening, MSIP)
Evaluate 690/52/152 properties (i.e., CGR testing) M PWSCC Investigate embrittlement effects in high Cr Ni alloys (690 M systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties (i.e., CGR testing) M SS DMW components M Aging Alloy X750 component L Mechanisms Validate models and design practices H Environmental Evaluate irradiation effects (indicate if interest is reactor-M Fatigue type specific)
Guide cards L Wear / CRDM thermal sleeves L Fretting Other high-wear components (please specify component L of interest)
Aging effects Consider interaction between fabrication defects and H on fabrication service-induced aging (e.g., weld lack of fusion and defects PWSCC, RPV embrittlement and hydrogen flaking)
Effectiveness Harvest repaired components (i.e., peened, overlayed, L of repair mechanical stress improvement for evaluation and testing techniques of long-term repair effectiveness Evaluate existing and emerging NDE procedures and H advances /developments BFB removal torque Thermal shield bolts/flexures M Mitigation Core barrel/shroud welds M and Inspection Inspect high usage fatigue locations (incl. environmental H Inspection and NDE factors)
Activities Conduct NDE before harvesting to identify optimal (e.g., M flawed) components/locations for harvesting Evaluate flaw distributions, types, and characteristics H (e.g., fabrication or service induced & mechanism) in critical components/welds Assess degradation in CRDM tubes, thermal sleeves, and H other locations (e.g., cracking, wear)
Component-Evaluate IASCC initiation and growth, especially for high-H Specific RPV Internals fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 dpa, >330C) M
F-21
Assess irradiation assisted creep H SS Welds (>2 dpa), i.e., for CGR and Fracture Toughness M RPV Internals (FT) testing Baffle bolts including IASCC initiation and crack growth M mechanisms Evaluate thermal aging M Evaluate thermal aging and irradiation embrittlement M (also flux /spectrum effects)
RPV nozzle - evaluate low flux, cavity streaming, and L temperature effects Measure Through-wall fracture toughness properties H Measure through-wall fluence values (i.e., to validate M calculations)
Validate embrittlement trend curve models (obtaining H high fluence materials)
RPV LAS Validate embrittlement trend curve models (obtaining H weld and low flux/temperature materials) base metal Evaluate untested surveillance specimens H Evaluate fabrication flaw distributions/types (e.g., H hydrogen flake cracking)
Component-Assess material orientation and variability effects H Specific including segregation (e.g., C segregation)
Small specimen techniques (from broken Charpy H surveillance specimens)
Near-inner surface including cladding materials (i.e., for H evaluation of residual stress and shallow flaw effects)
Reactor support components/material (i.e., for PWR, M VVER)
Evaluate thermal aging effects on fracture toughness of M CASS components (i.e., validation of lab. testing and models)
Evaluate thermal aging and irradiation effects on fracture M toughness of CASS components (i.e., validation of lab.
CASS and SS testing and models)
Welds Evaluate thermal aging effects on fracture toughness of M SS weld components (i.e., validation of lab. testing and models)
Evaluate thermal aging and irradiation effects on fracture M toughness of SS weld components (i.e., validation of lab.
testing and models)
Search for shallow flaws in A600 tubes for evaluating NDE M Steam effectiveness generators Investigate wear in A690 SG tubes H Examine Long-range ordering in A690 M
F-22
Component-Steam Evaluate SCC in SG tubes H Specific generators Evaluate Divider plate degradation M Pressure tubes L Feeder tubes L Reactor-CANDU Calandria vessel - 304SS at lower temperatures w/ very L Specific little data Investigate helium embrittlement in Ni alloys L VVER Horizontal SGs M
F-23
2022 International Workshop Harvesting Priorities (NRA / Japan)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress profiles M Flawed components M Evaluate mitigation effectiveness (water chemistry, H peening, MSIP)
Evaluate 690/52/152 properties (i.e., CGR testing) H PWSCC Investigate embrittlement effects in high Cr Ni alloys H (690 systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties (i.e., CGR testing) M SS DMW components M Aging Alloy X750 component M Mechanisms Validate models and design practices L Environmental Evaluate irradiation effects (indicate if interest is M Fatigue reactor-type specific)
Guide cards L Wear / Fretting CRDM thermal sleeves M Other high-wear components (please specify L component of interest)
Aging effects on Consider interaction between fabrication defects and L fabrication service-induced aging (e.g., weld lack of fusion and defects PWSCC, RPV embrittlement and hydrogen flaking)
Effectiveness of Harvest repaired components (i.e., peened, overlayed, H repair mechanical stress improvement for evaluation and techniques testing of long-term repair effectiveness Evaluate existing and emerging NDE procedures and H advances /developments BFB removal torque M Thermal shield bolts/flexures M Core barrel/shroud welds M Mitigation and Inspect high usage fatigue locations (incl. H Inspection Inspection and environmental factors)
NDE Activities Conduct NDE before harvesting to identify optimal H (e.g., flawed) components/locations for harvesting Evaluate flaw distributions, types, and characteristics M (e.g., fabrication or service induced & mechanism) in critical components/welds
Assess degradation in CRDM tubes, thermal sleeves, L and other locations (e.g., cracking, wear)
F-24
Evaluate IASCC initiation and growth, especially for H high-fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 dpa, >330C) M Assess irradiation assisted creep M RPV Internals SS Welds (>2 dpa), i.e., for CGR and Fracture H Toughness (FT) testing Baffle bolts including IASCC initiation and crack growth H mechanisms
Evaluate thermal aging M Evaluate thermal aging and irradiation embrittlement H (also flux /spectrum effects)
RPV nozzle - evaluate low flux, cavity streaming, and M temperature effects Measure Through-wall fracture toughness properties M Measure through-wall fluence values (i.e., to validate M calculations)
Validate embrittlement trend curve models (obtaining H high fluence materials)
Validate embrittlement trend curve models (obtaining H Component-RPV LAS weld low flux/temperature materials)
Specific and base metal Evaluate untested surveillance specimens H Evaluate fabrication flaw distributions/types (e.g., L hydrogen flake cracking)
Assess material orientation and variability effects M including segregation (e.g., C segregation)
Small specimen techniques (from broken Charpy M surveillance specimens)
Near-inner surface including cladding materials (i.e., H for evaluation of residual stress and shallow flaw effects)
Reactor support components/material (i.e., for PWR, L VVER)
Evaluate thermal aging effects on fracture toughness H of CASS components (i.e., validation of lab. testing and models)
Evaluate thermal aging and irradiation effects on H CASS and SS fracture toughness of CASS components (i.e.,
Welds validation of lab. testing and models)
Evaluate thermal aging effects on fracture toughness H of SS weld components (i.e., validation of lab. testing and models)
F-25
Evaluate thermal aging and irradiation effects on H CASS and SS fracture toughness of SS weld components (i.e.,
Welds validation of lab. testing and models)
Search for shallow flaws in A600 tubes for evaluating L Component-NDE effectiveness Specific Steam Investigate wear in A690 SG tubes H generators Examine Long-range ordering in A690 H Evaluate SCC in SG tubes M Evaluate Divider plate degradation L Pressure tubes L Feeder tubes L Reactor-CANDU Calandria vessel - 304SS at lower temperatures w/ L Specific very little data Investigate helium embrittlement in Ni alloys L VVER Horizontal SGs L
F-26
2022 International Workshop Harvesting Priorities (KAERI, KHNP / Korea)
Category Topic/Item Specific Interest Priority Comment KAERI KHNP Measure DMW residual stress M H profiles Flawed components M M Evaluate mitigation H H Effectiveness of Zn addition, effectiveness (water chemistry, WOL peening, MSIP)
Evaluate 690/52/152 properties H H CGR variations of heat-to-(i.e., CGR testing) heat, Lab-to-Lab PWSCC Investigate embrittlement M M effects in high Cr Ni alloys (690 systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties M H (i.e., CGR testing)
Aging SS DMW components M M Mechanisms Alloy X750 component L L Validate models and design M H Environmental practices Fatigue Evaluate irradiation effects H M (indicate if interest is reactor-type specific)
Guide cards M M Wear / CRDM thermal sleeves M M Fretting Other high-wear components L L (please specify component of interest)
Consider interaction between H H Aging effects fabrication defects and service-on fabrication induced aging (e.g., weld lack of defects fusion and PWSCC, RPV embrittlement and hydrogen flaking)
Harvest repaired components H H Effectiveness of repair after Effectiveness (i.e., peened, overlayed, long-term aging of repair mechanical stress improvement Mitigation techniques for evaluation and testing of and long-term repair effectiveness Inspection Inspection Evaluate existing and emerging M M and NDE NDE procedures and advances Activities /developments BFB removal torque M M
F-27
Thermal shield bolts/flexures M H Core barrel/shroud welds H H Inspect high usage fatigue M M locations (incl. environmental factors)
Conduct NDE before harvesting M L to identify optimal (e.g.,
Mitigation Inspection flawed) components/locations and and NDE for harvesting Inspection Activities Evaluate flaw distributions, M H types, and characteristics (e.g.,
fabrication or service induced &
mechanism) in critical components/welds Assess degradation in CRDM M M tubes, thermal sleeves, and other locations (e.g., cracking, wear)
Evaluate IASCC initiation and H H Evaluation for long term growth, especially for high-operation of WH plants fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 H M Evaluation for long term dpa, >330C) operation of WH plants RPV Internals Assess irradiation assisted M M creep SS Welds (>2 dpa), i.e., for CGR H M Evaluation for long term and Fracture Toughness (FT) operation of WH plants testing Baffle bolts including IASCC H H Evaluation for long term initiation and crack growth operation of WH plants Component mechanisms
-Specific Evaluate thermal aging M M Evaluate thermal aging and H H Evaluation in actual irradiation embrittlement (also components flux /spectrum effects)
RPV nozzle - evaluate low flux, M M cavity streaming, and RPV LAS weld temperature effects and base Measure Through-wall fracture H M metal toughness properties Measure through-wall fluence H M values (i.e., to validate calculations)
Validate embrittlement trend H M curve models (obtaining high fluence materials)
F-28
Validate embrittlement trend H L curve models (obtaining low flux/temperature materials)
Evaluate untested surveillance M M specimens Evaluate fabrication flaw M M distributions/types (e.g.,
hydrogen flake cracking)
Assess material orientation and M L RPV LAS weld variability effects including and base segregation (e.g., C metal segregation)
Small specimen techniques M H (from broken Charpy surveillance specimens)
Near-inner surface including M M cladding materials (i.e., for evaluation of residual stress and shallow flaw effects)
Reactor support M M components/material (i.e., for PWR, VVER)
Component Evaluate thermal aging effects M H
-Specific on fracture toughness of CASS components (i.e., validation of lab. testing and models)
Evaluate thermal aging and H H Evaluation in actual irradiation effects on fracture components toughness of CASS components (i.e., validation of lab. testing CASS and SS and models)
Welds Evaluate thermal aging effects M H on fracture toughness of SS weld components (i.e.,
validation of lab. testing and models)
Evaluate thermal aging and H H Evaluation in actual irradiation effects on fracture components toughness of SS weld components (i.e., validation of lab. testing and models)
Search for shallow flaws in M L Steam A600 tubes for evaluating NDE generators effectiveness Investigate wear in A690 SG H L SCC initiation from wear tubes
F-29
Examine Long-range ordering in M L Some concerns of LRO A690 Component Steam Evaluate SCC in SG tubes H M Evaluation of SCC initiation in
-Specific generators actual components Evaluate Divider plate M L degradation Pressure tubes M L Feeder tubes M L Calandria vessel - 304SS at M L Reactor-CANDU lower temperatures w/ very Specific little data Investigate helium M L embrittlement in Ni alloys VVER Horizontal SGs L L Not applicable in Korea
F-30
2022 International Workshop Harvesting Priorities (JSI / Slovenia)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress profiles H Flawed components H Evaluate mitigation effectiveness (water M chemistry, peening, MSIP)
Evaluate 690/52/152 properties (i.e., CGR M testing)
PWSCC Investigate embrittlement effects in high Cr Ni L alloys (690 systems) at higher temperatures(e.g., pressurizer nozzle welds)
Evaluate 600/82/182 properties (i.e., CGR M testing)
Aging SS DMW components M Mechanisms Alloy X750 component L Environmental Validate models and design practices H Fatigue Evaluate irradiation effects (indicate if interest M Not necessarily is reactor-type specific)
Guide cards L Wear / CRDM thermal sleeves L Fretting Other high-wear components (please specify L component of interest)
Aging effects Consider interaction between fabrication M on fabrication defects and service-induced aging (e.g., weld defects lack of fusion and PWSCC, RPV embrittlement and hydrogen flaking)
Effectiveness Harvest repaired components (i.e., peened, M of repair overlayed, mechanical stress improvement for techniques evaluation and testing of long-term repair effectiveness Evaluate existing and emerging NDE procedures M and advances /developments BFB removal torque L Mitigation Thermal shield bolts/flexures L and Core barrel/shroud welds L Inspection Inspect high usage fatigue locations (incl. M Inspection and environmental factors)
NDE Activities Conduct NDE before harvesting to identify M optimal (e.g., flawed) components/locations for harvesting Evaluate flaw distributions, types, and H characteristics (e.g., fabrication or service induced & mechanism) in critical components/welds
F-31
Mitigation Inspection and Assess degradation in CRDM tubes, thermal L and NDE Activities sleeves, and other locations (e.g., cracking, Inspection wear)
Evaluate IASCC initiation and growth, especially H for high-fluence (e.g., > 50 dpa) materials Investigate void swelling (>30 dpa, >330C) L RPV Internals Assess irradiation assisted creep L SS Welds (>2 dpa), i.e., for CGR and Fracture M Toughness (FT) testing Baffle bolts including IASCC initiation and crack H growth mechanisms Evaluate thermal aging M Evaluate thermal aging and irradiation M embrittlement (also flux /spectrum effects)
RPV nozzle - evaluate low flux, cavity M streaming, and temperature effects Measure Through-wall fracture toughness H properties Measure through-wall fluence values (i.e., to M validate calculations)
Validate embrittlement trend curve models M (obtaining high fluence materials)
RPV LAS weld Validate embrittlement trend curve models M Component and base (obtaining low flux/temperature materials)
-Specific metal Evaluate untested surveillance specimens M Evaluate fabrication flaw distributions/types H (e.g., hydrogen flake cracking)
Assess material orientation and variability L effects including segregation (e.g., C segregation)
Small specimen techniques (from broken L Charpy surveillance specimens)
Near-inner surface including cladding materials M (i.e., for evaluation of residual stress and shallow flaw effects)
Reactor support components/material (i.e., for L PWR, VVER)
Evaluate thermal aging effects on fracture L toughness of CASS components (i.e., validation of lab. testing and models)
CASS and SS Evaluate thermal aging and irradiation effects L Welds on fracture toughness of CASS components (i.e.,
validation of lab. testing and models)
Evaluate thermal aging effects on fracture L toughness of SS weld components (i.e.,
validation of lab. testing and models)
F-32
CASS and SS Evaluate thermal aging and irradiation effects L Welds on fracture toughness of SS weld components (i.e., validation of lab. testing and models)
Search for shallow flaws in A600 tubes for L Component evaluating NDE effectiveness
-Specific Steam Investigate wear in A690 SG tubes L generators Examine Long-range ordering in A690 L Evaluate SCC in SG tubes L Evaluate Divider plate degradation L Pressure tubes L Feeder tubes L Reactor-CANDU Calandria vessel - 304SS at lower temperatures L Specific w/ very little data Investigate helium embrittlement in Ni alloys L VVER Horizontal SGs L
JSI comments based on priority:
H In-line with current research topics and/or interests M In-line with past research topics and/or interests L Not a research topic and/or interst
F-33
2022 International Workshop Harvesting Priorities (NRC / USA)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual M/L Residual stresses are important to stress profiles understanding PWSCC susceptibility.
Some data will be obtained from SMILE.
There's also been quite a bit of previous work associated with measuring DMW WRS profiles so it's not clear what the objective of any new work in this area would be.
Flawed components H Flawed components are often of high interest to validate NDE effectiveness and gain a better understanding of degradation mechanisms.
Evaluate mitigation H Important to NRC since many licensees are effectiveness (water using different mitigation methods and chemistry, peening, MSIP) requesting relief from inspections based on mitigation. Obtaining components that had flaws and have been mitigated in some way would be of particular interest.
Evaluate 690/52/152 L Sufficient data exists.
properties (i.e., CGR Aging testing)
Mechanisms PWSCC Investigate embrittlement M/H Could be a topic of interest for long term effects in high Cr Ni alloys operation beyond 60 years, particularly if (690 systems) at higher the right materials (lower Fe) and temperatures(e.g., temperatures (> 320 °C) could be pressurizer nozzle welds) identified.
Evaluate 600/82/182 L There is already ample data on this.
properties (i.e., CGR testing)
SS DMW components H NRC is interested in austenitic stainless steel weld metal thermal aging and SCC growth. Not much harvesting of these type materials. SS DMWs will use different weld metals (e.g. 309) than similar metal SS welds, so it would also be of interest to harvest these, in addition to similar metal austenitic stainless steel welds, which are covered by another line item.
Alloy X750 component M/H Not that much data on CGR's for X750.
There has been some OE with SCC in RVI of X-750 components, so would be helpful to characterize CGRs.
F-34
Validate models and M Harvesting could be useful to validate design practices conservatism or overconservatism of current models. However, components selected must have a well characterized Environmental loading/thermal history and ideally would Fatigue have evidence of some service induced cracking.
Evaluate irradiation M/H Priority should be on components with effects (indicate if interest low-cycle fatigue loading, but high-cycle is reactor-type specific) fatigue components could also be of interest.
Guide cards L Wear can be measured in-situ.
CRDM thermal sleeves M/H OE in US PWRs with this issue, more difficult to inspect in-situ. NRC would support harvesting of thermal sleeves only Aging if both degraded and undegraded thermal Mechanisms Wear / sleeves could be obtained to evaluate Fretting wear/cracking locations, root cause, and possibly to develop/validate models Other high-wear L There are a number of components in components (please reactor vessel internals that are specify component of susceptible to wear, but wear is generally interest) adequately managed by periodic visual examination.
Consider interaction M/H Items such as RPV materials with between fabrication hydrogen flakes (already identified) and Aging effects defects and service-PWR bottom mounted instrumentation on fabrication induced aging (e.g., weld nozzles (weld lack-of fusion implicated in defects lack of fusion and PWSCC, SCC) may be of interest if available. The RPV embrittlement and challenge is to find components that have hydrogen flaking) both fabrication and service-induced aging.
Harvest repaired H MSIP welds are of high interest to NRC to components (i.e., peened, examine residual stress state and see if Effectiveness overlayed, mechanical SCC was prevented. Likewise peened of repair stress improvement for reactor vessel head penetrations. There techniques evaluation and testing of are probably other components/repair long-term repair methods as well. Components that had Mitigation effectiveness flaws prior to mitigation are of particular and interest.
Inspection Evaluate existing and M/H If advanced NDE can be performed on emerging NDE procedures critical components in a decommissioning Inspection and advances plant, then the same components are and NDE /developments harvested and destructively analyzed, this Activities would be highly valuable. Components and NDE techniques must be carefully selected.
F-35
BFB removal torque L Lots of BFB data has been gathered in US.
Removal torque should be measured when replacing BFBs but probably not worth the time if plant is decommissioning.
Thermal shield M/H There has been recent OE with cracking of bolts/flexures flexures in US. Destructive examination of harvested flexures could confirm the failure mechanism for these components.
Core barrel/shroud welds H PWR core shroud welds are desirable to NRC, there is little or no data on stainless steel welds with greater than 50 dpa.
Toughness, crack growth and void swelling tests are of interest to NRC.
Inspect high usage fatigue H Useful to determine degree of locations (incl. conservatism of environmental factors Mitigation Inspection environmental factors) currently used to determine and and NDE environmentally-adjusted CUFs.
Inspection Activities Conduct NDE before H NDE prior to harvesting should be harvesting to identify performed if possible to characterize optimal (e.g., flawed) flaws, if appropriate for the component components/locations for and degradation type of interest.
harvesting Evaluate flaw H Understanding of flaw distributions could distributions, types, and help improve PFM models.
characteristics (e.g.,
fabrication or service induced & mechanism) in critical components/welds Assess degradation in M/H This item is very broad - NRC interest CRDM tubes, thermal depends on what is available. Harvesting sleeves, and other these components is only of interest if locations (e.g., cracking, components with known inservice wear) degradation can be harvested and then destructively evaluated.
Evaluate IASCC initiation H Very little data on IASCC CGR for very high and growth, especially for fluence stainless steel. Data exists up to high-fluence (e.g., > 50 50 dpa so obtaining data with dpa) materials higherfluences ( 60 dpa) would be most valuable; however, any data above 20 dpa Component-is valuable.
Specific RPV Internals Investigate void swelling M Available high dose data showed lower
(>30 dpa, >330C) amounts of swelling compared to models, but more data is needed under PWR high fluence conditions to validate. Materials with both high fluence and high operating temperature are needed for void swelling assessment to be worthwhile.
F-36
Assess irradiation assisted M Irradiation assisted creep and irradiation-creep assisted stress relaxation are mechanisms that can cause loss of functionality of RVI.
This is an important topic but it's difficult to assess post-service without good knowledge of pre-service material/component state.
SS Welds (>2 dpa), i.e., for H Important to define thresholds for higher CGR and Fracture CGRs and low toughness.
Toughness (FT) testing RPV Internals Baffle bolts including M/L Baffle-former bolt cracking in US is fairly IASCC initiation and crack well managed, so not as much of a driver growth mechanisms for harvesting. It would be nice to gain a better understanding of the crack growth mechanisms. Initiation data would be useful for improving predictive models.
In-service loading of the bolts would need to be known, and while harvesting both with or without degradation would be useful, bolts with degradation can provide better quantitative calibration of bolting initiation models.
Evaluate thermal aging L Domestic harvesting effort in US of Component-pressurizer material will provide some Specific data on this. It is expected that thermal aging in RPV will be less significant than in pressurizer due to lower temperature.
Evaluate thermal aging L Diffficult to separate synergistic effects; and irradiation thermal aging at RPV temperatures not embrittlement (also flux expected to be significant.
/spectrum effects)
RPV nozzle - evaluate low M Cavity streaming is a topic of interest for flux, cavity streaming, and NRC. Not usually well modeled by typical RPV LAS weld temperature effects RPV neutron transport models. RPV and base nozzles have been dispositioned as metal generally not limiting for P-T limits, but higher fluences than currently predicted could change this conclusion. There must be sufficient baseline fracture toughness information available for benchmarking, which may not be the case for nozzles.
Measure Through-wall M/L Several prior studies done. Testing is fracture toughness expensive. More data coming from VTT properties under SMILE and ORNL for Zion Measure through-wall M/L Would be somewhat useful to validate fluence values (i.e., to attenuation models. However, the only validate calculations) way to measure is through retrograde dosimetry which seems highly uncertain.
F-37
Validate embrittlement H More data from high fluence RPV trend curve models materials is needed, sparse data from (obtaining high fluence surveillance and test reactors for fluences materials) > 6E19 n/cm2. Surveillance specimen data would be most effective for validating embrittlement models, versus actual RPV materials.
Validate embrittlement M/H Could be helpful to obtain more data in trend curve models parts of the variable space that are more (obtaining low thinly populated. There is shortage of low flux/temperature temperature (< 525 deg F/274 deg C) materials) embrittlement data. Some SMR designs plan to operate at lower temperatures than current LWRs.
Evaluate untested M/H Interest depends on fluence levels - high surveillance specimens fluence of more interest.
Evaluate fabrication flaw M Hydrogen flaking is of interest due to OE distributions/types (e.g., at Belgian PWRs, and questions about hydrogen flake cracking) whether similar-vintage plants also have flaking. General flaw distribution assessments could help inform PFM models.
RPV LAS weld Assess material M Carbon macrosegregation was a significant Component-and base orientation and variability issue in some LAS forgings in nuclear Specific effects including plants. Material orientation effects have metal segregation (e.g., C been well-studied; segregation effects segregation) have been studied, but less so; would likely have to be expansive program to provide new insights.
Small specimen L A lot of work has already been done in this techniques (from broken area and more is ongoing.
Charpy surveillance specimens)
Near-inner surface H Would be useful to look for shallow including cladding surface-breaking flaws to help validate or materials (i.e., for refute the existence of these type of flaws.
evaluation of residual stress and shallow flaw effects)
Reactor support H The extent of embrittlement of RPV components/material (i.e., supports has been an ongoing issue at for PWR, VVER) NRC for long-term operation. It would be helpful to have actual data from an operating plant to compare to models.
Low temperatures and different steel grades than RPV pressure boundary could cause a different response to irradiation making current RPV ETCs not applicable to supports.
F-38
Evaluate thermal aging M/H More data from materials aged in effects on fracture operating plants is needed to validate toughness of CASS models. Materials would need to be long-components (i.e., term, high temperature components with validation of lab. testing a chemistry of interest.
and models)
Evaluate thermal aging H Little or no data on CASS having and irradiation effects on simultaneous thermal aging and fracture toughness of irradiation in operating plants.
CASS components (i.e.,
validation of lab. testing CASS and SS and models)
Welds Evaluate thermal aging H There is little data on fracture toughness effects on fracture of SS welds aged in operating plants. This toughness of SS weld is an emerging topic of interest for NRC in components (i.e., the long-term operation area.
validation of lab. testing and models)
Evaluate thermal aging H Little data on synergistic effects, could be and irradiation effects on gained by fracture toughness testing of fracture toughness of SS welds from RVI.
Component-weld components (i.e.,
Specific validation of lab. testing and models)
Search for shallow flaws in M/H This is an ongoing issue in remaining SGs A600 tubes for evaluating in US with A600 tubes.
NDE effectiveness Investigate wear in A690 L Wear can be measured with NDE, so not SG tubes clear what additional information could be obtained through harvesting given the robustness of NDE.
Examine Long-range M/L Could be a potential LTO issue. Would ordering in A690 only be of interst if long-service tiem tubes Steam with lower Fe content could be found.
generators SMILE project is already doing this with the Ringhals plant.
Evaluate SCC in SG tubes L This is very broad. A lot of data on SCC in SG tubes generally.
Evaluate Divider plate M/L Divider plates are robust, not much OE on degradation this. There has been some divider plate cracking associated with crevice conditions; finding and evaluating degraded or crack divider plates would be of interest - less interest in random divider plate harvesting.
Reactor-CANDU Pressure tubes L This reactor type is not found in US - NRC Specific Feeder tubes L doesn't regulate any.
F-39
Calandria vessel - 304SS L at lower temperatures w/ This reactor type is not found in US - NRC Reactor-CANDU very little data doesn't regulate any.
Specific Investigate helium L embrittlement in Ni alloys VVER Horizontal SGs L This reactor type is not found in US - NRC doesn't regulate any.
F-40
2022 International Workshop Harvesting Priorities (INL / USA)
Blue letters represent comments by Simon Pimblott (reference to Nuclear Science User Facilities (NSUF))
Category Topic/Item Specific Interest PX/GB Peng Xu and Grace Burke Comments Priority Measure DMW residual stress Are there indications of PWSCC in those profiles DMWs? If so, priority would be high.
Would be useful if NDE has already identified the flaw. Studying these materials advances the understanding of Flawed components M/H materials failure mechanisms and helps the design of advanced materials and components.
I agree, however, this scope is not within the mission space of the NSUF Evaluate mitigation Depends on the component history.
effectiveness (water chemistry, peening, MSIP)
Evaluate 690/52/152 properties L a lot going on already (i.e., CGR testing)
Investigate embrittlement Is the temperature differential sufficient PWSCC effects in high Cr Ni alloys (690 to warrant further study? Presence of Fe systems) at higher L and Mn in A690 would inhibit Ni2Cr.
temperatures(e.g., pressurizer OK, agree Aging nozzle welds)
Mechanisms Evaluate 600/82/182 properties Science project - good for testing (i.e., CGR testing) mechanistic understanding of PWSCC and precursor reactions.
SS DMW components Are there indications of PWSCC in those DMWs? If so, priority would be high.
B effects? What is the fluence? Might there be any use in Gen IV. The precipitation process in X750 is very sensitive to heat treatment conditions and Alloy X750 component M initial microstructures. The understanding of the competing effect between precipitation and dissolution is necessary.
In other words we need a detailed pedigree, if we take the samples.
Validate models and design L Environment practices al Fatigue Evaluate irradiation effects The combined effect of irradiation and (indicate if interest is reactor-M/H fatigue has not been well understood.
type specific)
Wear / Guide cards Fretting CRDM thermal sleeves
F-41
Wear / Other high-wear components Fretting (please specify component of interest)
Consider interaction between Fabrication-induced defects play an Aging Aging fabrication defects and service-important role in the microstructure Mechanisms evolution during aging and irradiation.
effects on induced aging (e.g., weld lack of L/M Understanding the effects of fabrication fabrication fusion and PWSCC, RPV defects will help optimize the fabrication defects embrittlement and hydrogen processes.
flaking) OK.
Effectiveness Harvest repaired components We have supported a lot of "fundamental of repair (i.e., peened, overlayed, work" on fabricated repairs - so this is techniques mechanical stress improvement M/H probably a yes.
for evaluation and testing of long-term repair effectiveness Evaluate existing and emerging Developing advanced NDE techniques NDE procedures and advances M/H with higher resolution is important to
/developments detect small flaws.
Not within the scope of the NSUF BFB removal torque Thermal shield bolts/flexures Core barrel/shroud welds Inspect high usage fatigue Are there indications of PWSCC in those locations (incl. environmental H DMWs?
factors) I agree, however, this scope is not within Mitigation the mission space of the NSUF and Sounds appropriate - if the NDE Inspection Inspection Conduct NDE before harvesting techniques are well-qualified. It is and NDE to identify optimal (e.g., flawed) H necessary to conduct NDE to identify the Activities components/locations for location of interest before harvesting.
harvesting I agree, however, this scope is not within the mission space of the NSUF Non-destruction examination of flows in Evaluate flaw distributions, components and repairs before and after types, and characteristics (e.g., service provides the guidance on the fabrication or service induced & M/H loading cycles. - but will require mechanism) in critical DE/characterization as well components/welds I agree, however, this scope is not within the mission space of the NSUF Assess degradation in CRDM tubes, thermal sleeves, and other locations (e.g., cracking, wear)
F-42
Evaluate IASCC initiation and The data of high-dose irradiated RPV growth, especially for high-internals is very limited. Understanding fluence (e.g., > 50 dpa) materials H the initiation and propagation of IASCC is imperative for their assessment in long-term service. Are there such materials available?
Investigate void swelling (>30 Although there are some data available in dpa, >330C) the literature, more detailed M microstructure and microchemistry characterization are required, especially using advanced analytic electron RPV microscopes.
Internals Assess irradiation assisted creep The study of time-dependent mechanical properties is very limited. Understanding M/H and measuring irradiation creep properties could prevent the catastrophic failure of in-core components. Need to know where it come from.
SS Welds (>2 dpa), i.e., for CGR and Fracture Toughness (FT) M/H testing Baffle bolts including IASCC NSUF already has this sort of material.
Component-initiation and crack growth Specific mechanisms and segregation effects via examination of harvested pressurizers; would provide a good thermal baseline for the extrapolation of high fluence (100 mdpa Evaluate thermal aging H or so) hardening features; also answer "LBP" claims with respect to embrittlement models; I agree, however, this scope is not within the mission space of the NSUF RPV LAS Would be good to coordinate between weld and NRC/NE/EPRI; value may be limited if base there are no archive materials for metal Evaluate thermal aging and comparison. Thermal and irradiation irradiation embrittlement (also M/L embrittlement is not well understood, flux /spectrum effects) many hypotheses have been proposed, but there in no consensus on the embrittlement mechanisms, more studies are necessary.
RPV nozzle - evaluate low flux, What are to max fluences?
cavity streaming, and temperature effects Measure Through-wall fracture hasnt this been done already?
toughness properties
F-43
Measure through-wall fluence Hasnt this been done already?
values (i.e., to validate calculations)
Validate embrittlement trend Good for combined mech curve models (obtaining high H properties/embrittlement and detailed fluence materials) materials characterization (to elucidate mechanistic details).
Validate embrittlement trend low irradiation temp steels/welds will be curve models (obtaining low H useful for SMRs (small PWRs) flux/temperature materials)
Evaluate untested surveillance specimens Evaluate fabrication flaw I did not think that there was any flaking-distributions/types (e.g., related fracture (flake formation hydrogen flake cracking) occurred very early in life of the steel).
But it makes a good case for always RPV LAS keeping original NDE techniques for weld and examination along with the latest and base greatest ones.
metal Assess material orientation and Not an issue for forgings I presume.
variability effects including However, The minor elements, such as C, segregation (e.g., C segregation) S, P, Si, et al., could be important to the L/M microstructural evolution under Component-irradiation and thermal aging, for Specific example, changing the carbon content can change the void swelling rate in HT-9.
Small specimen techniques Small scale characterization and testing (from broken Charpy M are necessary for fundamental surveillance specimens) understanding of materials failure.
Near-inner surface including cladding materials (i.e., for evaluation of residual stress and shallow flaw effects)
Reactor support depends on operational life of harvested components/material (i.e., for M/H components; fluence/flux/irradiation PWR, VVER) Temp Evaluate thermal aging effects on fracture toughness of CASS components (i.e., validation of lab. testing and models) depends on length of service operation.
CASS and Evaluate thermal aging and The synergistic effect of thermal aging and SS Welds irradiation effects on fracture irradiation changes the microstructural toughness of CASS components M/H evolution, and thus controlling the fracture (i.e., validation of lab. testing model of SS. Experimental and modeling and models) are required to develop predictive models for fracture toughness.
Fair enough
F-44
Evaluate thermal aging effects on fracture toughness of SS weld components (i.e.,
validation of lab. testing and models)
CASS and Evaluate thermal aging and The synergistic effect of thermal aging SS Welds irradiation effects on fracture and irradiation changes the toughness of SS weld microstructural evolution, and thus components (i.e., validation of M/H controlling the fracture model of SS.
lab. testing and models) Experimental and modeling are required to develop predictive models for fracture toughness.
Component-Search for shallow flaws in A600 I agree, however, this scope is not within Specific tubes for evaluating NDE M the mission space of the NSUF effectiveness Investigate wear in A690 SG Is it an issue? Would check with EPRI tubes The order-disorder transformation Steam Examine Long-range ordering in M determines the mechanical properties.
generators A690 I agree, however, this scope is not within the mission space of the NSUF Is there any? Other than Pb-caustic?
Evaluate SCC in SG tubes L/M I agree, however, this scope is not within the mission space of the NSUF Evaluate Divider plate Is it an issue? Are plates A600 or A690?
degradation L I agree, however, this scope is not within the mission space of the NSUF Pressure tubes Feeder tubes Calandria vessel - 304SS at ?
lower temperatures w/ very little data He embrittlement turns to be very severe Reactor-CANDU in the components after long-term Specific service. The effect of long-term Investigate helium irradiation and initial microstructure on embrittlement in Ni alloys H the embrittlement need more investigations.
I agree with the comment; however, I am not sure that CANDU reactors are within our scope.
VVER Horizontal SGs
F-45
2022 International Workshop Harvesting Priorities (EPRI / USA)
Category Topic/Item Specific Interest Priority Comment Measure DMW residual stress L often these are very joint profiles configuration/fabricator/plant life history specific to be broadly useful Flawed components L The specific interest is too generic to validate priority; if flaws were detected during service in unexpected material/component - could be high priority Evaluate mitigation M this is partially repeated below on effectiveness (water effectiveness of repair techniques (peened chemistry, peening, MSIP) components are of interest to study; possibly materials of varying exposure to Zn
& noble chem would be interesting to have more fully characterized under various conditions)
PWSCC Evaluate 690/52/152 M Long-term service materials at higher properties (i.e., CGR testing) temperature would be useful to evaluate CGRs Investigate embrittlement M Consider "long term aging effects" instead effects in high Cr Ni alloys (690 of "embrittlement effects" Aging systems) at higher Mechan temperatures(e.g., pressurizer isms nozzle welds)
Evaluate 600/82/182 L Ample laboratory data properties (i.e., CGR testing)
SS DMW components H SS welds and SS DMWs - given EDF OE; harvesting location would need to be strategic Alloy X750 component H High for BWR and PWR IMT; interest in SCC, stress relaxation and fatigue behavior (irradiated X-750)
Validate models and design n/a What is the correlation of this interest with Environm practices harvesting?
ental Evaluate irradiation effects L What is the availability of EAF endurance Fatigue (indicate if interest is reactor-testing for irradiated materials?
type specific)
Guide cards M Mid-level guide cards in the CRGT; harvest ion-nitridated RCCA Wear / CRDM thermal sleeves H Inspection activity not material retrieval Fretting (thermal sleeve flange and housing)
Other high-wear components n/a Annotate ranking table to remind (please specify component of Participants that harvesting and inspection interest) are both options to obtain needed data.
F-46
Consider interaction between n/a Defects are addressed in other rows.
Aging Aging fabrication defects and Mechan effects on service-induced aging (e.g.,
isms fabrication weld lack of fusion and defects PWSCC, RPV embrittlement and hydrogen flaking)
Harvest repaired components M Peened components (i.e., peened, overlayed, Effectiveness mechanical stress of repair improvement for evaluation techniques and testing of long-term repair effectiveness Evaluate existing and H emerging NDE procedures and advances /developments BFB removal torque L Thermal shield bolts/flexures M Inspection activity ongoing diminishing priority for data from decommissioned plants Core barrel/shroud welds H Mitigation Inspect high usage fatigue M Inspection activity with potential for and locations (incl. environmental harvesting Inspection factors)
Inspection Conduct NDE before n/a and NDE harvesting to identify optimal Activities (e.g., flawed) components/locations for harvesting Evaluate flaw distributions, n/a types, and characteristics (e.g., fabrication or service induced & mechanism) in critical components/welds Assess degradation in CRDM n/a Consider inspections of core barrel guide tubes, thermal sleeves, and lug / RV keyway interface (including bolts, other locations (e.g., cracking, shims, etc.)
wear)
Evaluate IASCC initiation and H Harvest high-fluence (> 50 dpa) wrought or growth, especially for high-weld materials (use this verbiage in lieu of fluence (e.g., > 50 dpa) the laboratory testing objective of materials harvested material, note parenthetically Component RPV the purpose of harvesting)
-Specific Internals Investigate void swelling (>30 M Harvest high-fluence, high temperature dpa, >330C) stainless steel (just note paranthetically it is for void swelling research)
Assess irradiation assisted L Stress relaxation is a higher priority than creep creep for LWRs
F-47
SS Welds (>2 dpa), i.e., for M High-priority (H) for SS base metal FT CGR and Fracture Toughness testing (5 to 10 dpa)
(FT) testing RPV Baffle bolts including IASCC L Suggest breaking Baffle bolts into 2 lines:
Internals initiation and crack growth (1) In-tact/sound baffle bolts for IASCC mechanisms initiation / CGR testing material, and (2)
BFBs with indications for evaluating degradation in service Evaluate thermal aging L L - U.S. / M - International; EPRI already has a program harvesting/testing LAS from PZR Evaluate thermal aging and L EPRI has several programs evaluating (see irradiation embrittlement comments on other line items)
(also flux /spectrum effects)
RPV nozzle - evaluate low M Not much data available; PWROG flux, cavity streaming, and addressed this analytically for U.S. PWRs temperature effects and select international PWRs P-T Limits Measure Through-wall H ORNL study on Zion may complement this fracture toughness properties need Measure through-wall fluence n/a See above; uncertain how this is done via values (i.e., to validate harvesting. Must be done in conjunction calculations) with FT data Validate embrittlement trend L EPRI has programs that address: PSSP and curve models (obtaining high CRVSP (MRP-412 and 326, Rev. 1, Component fluence materials) respectively); EPRI agrees with importance
-Specific of topic, but harvesting unlikely to contribute at this time.
RPV LAS Validate embrittlement trend L EPRI has programs that address: PSSP and weld and curve models (obtaining low CRVSP (MRP-412 and 326, Rev. 1, base metal flux/temperature materials) respectively); EPRI agrees with importance of topic, but harvesting unlikely to contribute at this time.
Evaluate untested surveillance L Significant surveillance data on RV already specimens exists Evaluate fabrication flaw L Doel/Tihange issue dispositioned (MRP-367, distributions/types (e.g., Rev. 1 proved analytically no impacts if H hydrogen flake cracking) flakes found; PWROG project confirmed that no flakes were present in U.S. PWR forgings).
Assess material orientation and M Still a concern for EDF fleet; EPRI is variability effects including researching with MAI and has additional segregation (e.g., C segregation) projects pending; prior work to disposition covered in MRP-417, Rev. 1)
Small specimen techniques L This is well-defined and accepted by many (from broken Charpy international programs accepting miniature surveillance specimens) CT specimens, etc. for use in test programs Near-inner surface including H Lack of data available and research done on cladding materials (i.e., for cladding in general; MRP-437 documents evaluation of residual stress prior EPRI disposition (analytical) for this and shallow flaw effects) issue
F-48
Reactor support H Long-term operation potential concern.
RPV LAS components/material (i.e., for Harvesting could help support resolution, weld and PWR, VVER) since conservative analysis is having base metal difficulty to generically disposition for PWRs. Could be an issue with other reactor designs as well.
Evaluate thermal aging effects L BWR IMT not identified as an issue; PWR on fracture toughness of CASS and VVER IMT ranked low for SCC impacts components (i.e., validation of to CASS components (FT not a focus);
lab. testing and models) however, it is a focus for some in EPRI MRP to validate thermal embrittlement models.
LTCP in CASS identified as lacking data in the IMTs Evaluate thermal aging and L BWRVIP-234-A, BWRVIP-315 resolved; irradiation effects on fracture closed in BWR IMT; PWR IMT identifies toughness of CASS synergistic effects as well-characterized for CASS and components (i.e., validation of CASS; VVER IMT does not identify issue SS Welds lab. testing and models)
Evaluate thermal aging effects L IMTs do not uniquely identify this as an Component on fracture toughness of SS issue for welds
-Specific weld components (i.e.,
validation of lab. testing and models)
Evaluate thermal aging and M PWR IMT ranked high; VVER IMT ranked irradiation effects on fracture Medium; BWR IMT does not list as an issue toughness of SS weld components (i.e., validation of lab. testing and models)
Search for shallow flaws in L Very low A600 tubes for evaluating NDE effectiveness Investigate wear in A690 SG L Operational inspection data are abundant tubes Steam Examine Long-range ordering M Useful if hot leg, long-term aged material -
generators in A690 particularly High-Cr Ni-based welds.
Evaluate SCC in SG tubes L Evaluate Divider plate L Suggest adding an additional line for degradation secondary side of SG tubing to assess water cleanliness/maintenance in plants (e.g.,
oxide scale examinations)
Pressure tubes L Pressure tubes are already periodically harvested and surveilled (single fuel Reactor-channel replacement) as well as scraped for Specific CANDU hydrogen / deuterium measurements; there is no need to evaluate these. They are also extracted and replaced fully every
~30 years with full refurbishment project.
F-49
Feeder tubes L Feeder "pipes" is a more accurate (M for descriptor. Not irradiated component, and DMWs) there is limited degradation with cracking, only experienced at one unit - was identified as potentially hydrogen-assisted creep cracking. FAC experience also well known and characterized. Feeder pipes are also extracted and replaced every ~30 years with full refurbishment project.
Some dissimilar metal welds are examined in plant-specific programs; potential benefits for harvesting DMWs for PWSCC in CANDU collaborative form.
Calandria vessel - 304SS at H High value to characterize calandria Reactor-lower temperatures w/ very tubesheet welds under EOL fluence Specific little data condition (~90 years life), so this would only be applicable with a long-service, decommissioned reactor. Interim data points could potentially be helpful - boat sampling or intermediate life reactors.
Investigate helium L Garter springs are periodically harvested embrittlement in Ni alloys (along with the pressure tubes) through single fuel channel replacements.
Degradation mode is well characterized and well managed. EPRI MDM ranking low in latest revision.
Horizontal SGs M VVER SMEs - increasing significance of sludge levels on DMWs (cracking) - Tubes VVER are 321SS welded to a vertical collector pipe (CS), with unknown alloy DMWs.
Increased priority in revised MDM.
F-50