ML19205A439
| ML19205A439 | |
| Person / Time | |
|---|---|
| Site: | Seabrook  |
| Issue date: | 07/24/2019 |
| From: | NRC/OGC |
| To: | Atomic Safety and Licensing Board Panel |
| SECY RAS | |
| References | |
| 50-443-LA-2, ASLBP 17-953-02-LA-BD01, RAS 55108 | |
| Download: ML19205A439 (96) | |
Text
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION ATOMIC SAFETY AND LICENSING BOARD In the Matter of Docket No. 50-443-LA-2 NEXTERA ENERGY SEABROOK, LLC ASLBP No. 17-953-02-LA-BD01 (Seabrook Station, Unit 1)
Hearing Exhibit Exhibit Number: NRC066 Exhibit
Title:
U.S. Department of Energy, Light Water Reactor Sustainability Program A Summary of Collaborative Research and Development Activities, INL/EXT-19-52416, Rev. 0 (Jan. 2019)
INL/EXT-11-23452 Revision 7 Light Water Reactor Sustainability Program Integrated Program Plan November 2018 U.S. Department of Energy Office of Nuclear Energy
DISCLAIMER This information was prepared as an account of work sponsored by an agency of the U.S. Government. Neither the U.S. Government nor any agency thereof, nor any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness, of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. References herein to any specific commercial product, process, or service by trade name, trade mark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the U.S. Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the U.S. Government or any agency thereof.
EXECUTIVE
SUMMARY
Nuclear power has safely, reliably, and economically contributed approximately 20% of electrical generation in the United States (U.S.) over the past two decades. It remains the single largest contributor (more than 60%) of non-greenhouse-gas-emitting electric power generation in the U.S.
Most currently operating nuclear power plants will begin reaching the end of the 20-year extension to their original 40-year operating license, for a total of 60 years of operation (the oldest commercial plants in the U.S. reached their 40th anniversary in 2009). Figure E-1 shows projected nuclear energy contribution to the domestic generating capacity for 40 and 60-year license periods. If currently operating nuclear power plants do not operate beyond 60 years (or new nuclear plants are not built quickly enough to replace them), the total fraction of generated electrical energy from nuclear power will rapidly decline. That decline will be accelerated if plants are shut down before 60 years of operation.
Decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements.
Figure E-1. Projected nuclear power generation for 40 and 60-year license periods.
Operation of the existing fleet of plants to 60 years, extending the operating lifetimes of those plants beyond 60 years and, where practical, making further improvements in their productivity are essential to support the nations energy needs. Recently, several utilities have submitted applications to the U.S. NRC to begin the subsequent license renewal process extending the operating license period beyond 60 years from the date of their initial licensing. This marks an important planned milestone in the history of commercial nuclear power operations in the U.S.one that underscores the long-term dependability of these plant designs and the commitment to their long-term performance by the organizations that operate them. The Light Water Reactor Sustainability (LWRS)
Program will continue to work with owner-operators to address the key issues needed to support the technical bases for continued safe, long-term operation of our nations nuclear power assets.
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The U.S. Department of Energys Office of Nuclear Energys (DOE-NEs)a primary mission is to advance nuclear power as a resource capable of making major contributions in meeting the nations energy supply, environmental, and energy security needs. DOE-NE is working to revitalize the nuclear energy sector by addressing three main priorities:
- 1. Expand the lifespan of the nations existing fleet,
- 2. Develop a new pipeline of advanced nuclear reactors, and
- 3. Strengthen the nations fuel-cycle infrastructure.
The LWRS Program is the primary programmatic activity that addresses Priority 1. For the LWRS Program, sustainability is defined as the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for as long as possible and practical. It has two facets with respect to long-term operations: 1) to provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model; and 2) to manage the aging of plant systems, structures, and components (SSCs) so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically.
The Department of Energys (DOEs) role in meeting Priority 1 is to work with industry and interface with the NRC and key industry-support groups to conduct the research needed to inform major component refurbishment and replacement strategies, performance enhancements, plant-license extensions, and age-related regulatory oversight decisions. DOEs research, development, and demonstration role focuses on aging phenomena and issues that require long-term research and/or unique DOE laboratory expertise and facilities and are applicable to a broad range of operating reactors. When appropriate, the costs of research and development (R&D) and demonstration activities are shared with industry or the NRC. Pilot projects and collaborative activities are underway at commercial nuclear facilities and with industry organizations.
The following LWRS Program R&D pathways address DOE-NEs research Priority 1:
Materials Research. R&D to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants. This work will provide data and methods to assess the performance of SSCs essential to safe and sustained nuclear power plant operations. The R&D products will be used to define operational limits and aging-mitigation approaches for materials in nuclear power plant SSCs subject to long-term operating conditions, providing key input to both regulators and industry.
Plant Modernization. R&D to address nuclear-plant economic viability in current and future energy markets through innovation, efficiency gains, and business-model transformation through digital technologies. This includes addressing long-term aging and modernization or replacement of legacy instrumentation and control (I&C) technologies by research, development, and testing of new I&C technologies and advanced condition-monitoring technologies for more automated and reliable plant operation. The resulting R&D products will enable modernization of plant systems and
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processes while building a technology-centered business-model platform that supports improved performance at lower cost.
Risk-informed Systems Analysis. R&D to optimize safety margins and minimizing uncertainties to achieve high levels of safety and economic efficiencies. The pathway will: (1) deploy the method and tools of technologies that enable better representation of safety margins and the factors that contribute to cost and safety; and (2) conduct advanced risk-assessment applications with industry to support margin management strategies that enable more cost-effective plant operation. The methods and tools provided by the pathway will support effective safety margin management for both active and passive SSCs.
Measurable milestones have been developed for each of the pathways; these include both near-term (i.e., 1-5 years) and longer-term (i.e., beyond 5 years) milestones. High-level planned accomplishments in the near-term include:
Provide a mechanistic understanding of key materials degradation processes, predictive capabilities, and high-quality data to inform decisions and processes by both industry and regulators including:
Containment inspection guidelines for extended-service conditions Predictive models for swelling in light-water-reactor components, aging of cast austenitic stainless steel components, cable degradation, and nickel-base alloy stress corrosion cracking susceptibility Model of transition temperature shifts in reactor pressure vessel steels, precipitate phase stability and formation in internal primary water coolant components and reactor pressure vessel steels, and environmentally assisted fatigue in light-water-reactor components Methodology and techniques for a system for nondestructive examination of concrete sections, impact assessment of alkali-silica reaction affected concrete, radiation-induced changes, and synergistic environmental stressor damage in concrete and cable insulation Development and transfer of weld-repair techniques for irradiated materials to industry and the evaluation of new replacement alloys.
Technical basis and supporting reports and studies needed to broadly implement digital technologies and modernize plants including:
Methods and studies on migrating existing analog control rooms to hybrid integrated control-room technologies incorporating digital systems, advanced alarm systems, and computer-based control-room procedures Cost-benefit studies for deploying technologies that are the subject of R&D in actual nuclear power plants Human-performance improvement for nuclear power plant field workers based on application of technologies, such as radio-frequency identification (RFID), for management of tools and materials used in nuclear plant maintenance New analytical capabilities for reducing operational and schedule-adherence risks for nuclear refueling outages iii
Advanced online monitoring and data-analytics technologies used for applications of structural-health monitoring of nuclear-plant components to reduce maintenance and operating costs.
Integrated probabilistic risk assessments with cost analysis, and multiphysics best-estimate plus uncertainty-engineering tools to optimize the economic and safety performance of existing nuclear power plants:
Demonstrate enhanced plant resiliency with industry adoption of accident-tolerant fuel, optimal utilization of FLEX equipment, augmented or new passive cooling systems, and improved fuel cycle efficiency Demonstrate improved economic performance of existing nuclear power plants by recovering safety margins by reducing uncertainties and conservatisms of legacy licensing, design, and analysis bases through applications of the risk-informed systems analysis (RISA) toolkit Validate Terry-turbine models for input into system-level severe-accident analysis codes (e.g., MAAP, MELCOR) for evaluating extended core cooling capabilities during off-normal plant conditions.
Sections 1 through 4 in this document provide a comprehensive overview of the LWRS Program and the near-term and longer-term milestones.
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CONTENTS EXECUTIVE
SUMMARY
........................................................................................................................... i ACRONYMS ............................................................................................................................................... ix
- 1. BACKGROUND ................................................................................................................................ 1 1.1 Program Overview ................................................................................................................... 6 1.2 Program Management .............................................................................................................. 7 1.3 Program Research and Development Interfaces ...................................................................... 8 1.3.1 Industry ....................................................................................................................... 9 1.3.2 Nuclear Regulatory Commission ................................................................................ 9 1.3.3 International ................................................................................................................ 9 1.3.4 Universities ............................................................................................................... 10 1.3.5 Advanced Modeling and Simulation Tools............................................................... 11 1.4 Summary ................................................................................................................................ 11
- 2. MATERIALS RESEARCH ............................................................................................................. 13 2.1 Background ............................................................................................................................ 13 2.2 Research and Development Purpose and Goals ..................................................................... 14 2.3 Pathway Research and Development Areas ........................................................................... 15 2.3.1 Assessment and Integration ...................................................................................... 16 2.3.2 Reactor Metals .......................................................................................................... 16 2.3.3 Concrete .................................................................................................................... 24 2.3.4 Cabling ...................................................................................................................... 28 2.3.5 Buried Piping ............................................................................................................ 30 2.3.6 Mitigation Technologies ........................................................................................... 30 2.3.7 Integrated Industry Activities.................................................................................... 33 2.4 Research and Development Collaborations ........................................................................... 34 2.5 Summary of Research and Development Products and Schedule.......................................... 35
- 3. PLANT MODERNIZATION ........................................................................................................... 36 3.1 Background ............................................................................................................................ 36 3.2 Research and Development Purpose and Goals ..................................................................... 38 3.3 Plant Modernization Pilot Project Descriptions and Deliverables ......................................... 43 3.3.1 Instrumentation and Control Architecture ................................................................ 43 3.3.2 Online Monitoring and Plant Automation................................................................. 45 3.3.3 Advanced Applications and Process Automation ..................................................... 47 3.4 Research Facilities and Capabilities....................................................................................... 50 3.5 Cybersecurity ......................................................................................................................... 51 3.6 Research and Development Collaborations ........................................................................... 52
- 4. RISK-INFORMED SYSTEMS ANALYSIS ................................................................................... 54 v
4.1 Background ............................................................................................................................ 54 4.2 Research and Development Purpose and Goals ..................................................................... 55 4.3 Research, Development, and Demonstration Activities and Milestones ............................... 56 4.3.1 Enhanced Resilient Nuclear Power Plant Concepts .................................................. 57 4.3.2 Cost and Risk Categorization Applications .............................................................. 59 4.3.3 Margin Recovery and Operation Cost Reduction ..................................................... 61 4.3.4 RISA Toolkit Deployment Plan ................................................................................ 63 4.3.5 Verification and Validation (V&V) of the RISA Toolkit ......................................... 65 4.3.6 Pilot demonstration using the RISA toolkit .............................................................. 65 4.3.7 RISA Toolkit Industry Deployment and Feedback ................................................... 65 4.4 Research and Development Partnerships ............................................................................... 66 FIGURES Figure E-1. Projected nuclear power generation for 40 and 60-year license periods. ................................... i Figure 1. Current U.S. electricity-generation portfolio, showing dominance of nuclear as a low-carbon-emission power source. .................................................................................................... 2 Figure 2. U.S. electrical generation capacity factors by energy source, showing high operating performance. ................................................................................................................................. 2 Figure 3. National distribution of the 98 operating nuclear power plants in the U.S. .................................. 4 Figure 4. Nuclear power plant initial license date and license extension plans (as of September 2018). ............................................................................................................................................ 4 Figure 5. Light Water Reactor Sustainability Program organization. ........................................................... 8 Figure 6. Complexity of interactions between materials, environments, and stresses in a nuclear power plant and the impact they have on operations. ................................................................. 13 Figure 7. Experimental database sets on RPV alloys relative to surveillance and vessel-service data as a function of flux and fluence (left), and (right) the comparison of experimental ATR-2 hardening (y) data to those predicted using machine learning using IVAR and previous datasets for fitting. ................................................................................................. 17 Figure 8. Results of 1D axisymmetric, 2D planar, and 3D Grizzly models of the global response of an RPV at a point in time during a PTS event........................................................................ 18 Figure 9. Example of the size dimensions of the MCT test specimen to that of a conventional Charpy V-notch test bar (shown as an already broken/tested bar) and the comparison of MCT data to other compact tension-test geometries on the Master Curve for Linde-80 weld metal from the Midland reactor. ........................................................................................ 19 Figure 10. Experimental results of atomic-force microscopy and schematic diagrams of the differences in corrosion-layer development between LiOH and KOH water chemistry.
The formation of the passivating oxide layer is inhibited by the presence of hydrated Li+ ions. These ions undergo dehydration on the surface, followed by preferential vi
adsorption of -OH from water contained in the electrical double layer (EDL). This action results in the perturbation of the latter, surface acidification, and formation of a defective oxide film that provides less substrate protection from corrosion. .......................... 21 Figure 11. Schematic examples of the different stages of crack initiation occurring in Alloy 600 (top) and Alloy 690 (bottom). ..................................................................................................... 22 Figure 12. Estimated Weibull cumulative distribution function (CDF) for fatigue in air (left) and PWR water (right) conditions, comparing modeled (grey lines) to experimental sample sets (black lines and symbols) at different strain amplitudes (a). Demonstrating that the model provides a good correlation and can be used for estimating CDF when experimental data points are not available.................................................................................. 23 Figure 13. Cut-away of a typical pressurized water reactor, illustrating large volumes of concrete and the key role of concrete performance (source: NRC). ......................................................... 25 Figure 14. Process for the MOSAIC tool to assess concrete susceptibility to radiation-induced damage starting from the structural inputs of two modulator generalized ellipsometry microscopy and X-ray fluorescence that are developed into a mineral phase-distribution image prior to passing through a non-liner fast Fourier transformation solver to simulate the damage generated in the concrete aggregate structure. .......................................... 26 Figure 15. Linear array ultrasound data collected on a thick specimen containing intentional flaws. Signal processing, using SAFT, of a given flaw is compared to that of the MBIR forward model showing improved imaging of a defect. ............................................................. 28 Figure 16. Diagram of the technical approach to cable-aging studies towards deployable NDE methods for determining remaining useful life........................................................................... 28 Figure 17. (left) Forecast of helium generation at 75 wppm boron at 60 EFPY. Red Zone: >10 appm He (not weldable with current welding processes); Yellow Zone: 0.1 to 10 appm He (weldable with heat-input control during welding repair); Green Zone: <0.1 appm He (No special process control is needed in welding repair). (Source: EPRI report, EPRI BWRVIP-97A). (right) Inside view of the laser and friction stir weld subsystems installed in the hot-cell cubicle for testing irradiated materials. ................................................. 31 Figure 18. Current instrumentation in a nuclear power plant control room is dominated by analog technology. ................................................................................................................................. 37 Figure 19. The Plant Modernization Pathway is developing an architecture that encompasses all aspects of plant operations and support, integrating plant systems, and immersing plant workers in a seamless information architecture. ......................................................................... 40 Figure 20. Pilot projects for the Plant Modernization Pathway. ................................................................. 41 Figure 21. Stages of transformation in the Plant Modernization Pathway.................................................. 42 Figure 22. HSSL: a reconfigurable hybrid control room simulator. ........................................................... 50 Figure 23. HSSL provides critical support as researchers hold operator workshops with nuclear utilities to evaluate new control-room display designs. .............................................................. 51 Figure 24. Safety and economic benefit from the RISA Pathway. ............................................................. 54 vii
Figure 25. The RISA Pathway programmatic structure. ............................................................................. 55 Figure 26. Types of analysis that are used in the RISA Pathway. .............................................................. 56 Figure 27. RISC methodology (courtesy of 10 CFR 50.69). ...................................................................... 59 Figure 28. Integrated framework for optimizing maintenance activities. ................................................... 60 Figure 29. Current software modules used to perform RISA-specific analyses. ........................................ 64 Figure 30. Notional 5-year plan for a pilot demonstration of industry application of the RISA Pathway. ..................................................................................................................................... 65 TABLES Table 1. Pilot projects related to RISA R&D focus areas. .......................................................................... 57 viii
ACRONYMS ASR alkali-silica reaction ASSW austenitic stainless steel welds ATF accident-tolerant fuel ATR Advanced Test Reactor BWR boiling water reactor CASL Consortium on Advanced Simulation of LWRs CASS cast austenitic stainless steel CDF cumulative distribution function CFR Code of Federal Regulations DOE Department of Energy DOE-ID Department of Energy Idaho Operations Office DOE-NE Department of Energy Office of Nuclear Energy EMDA Expanded Materials Degradation Assessment EPRI Electric Power Research Institute ESSAI Energy Systems, Strategies, Assessments, and Integration FLEX Diverse and Flexible Coping Strategy HSSL Human Systems Simulation Laboratory I&C Instrumentation and Control IAEA International Atomic Energy Agency IASCC irradiation-assisted stress corrosion cracking ICWE Integrated Computational Welding Engineering II&C instrumentation, information, and control INL Idaho National Laboratory LOCA loss of coolant accident LWR light-water reactor LWRS Light Water Reactor Sustainability MAI Materials Aging Institute MCT miniature compact tension NDE nondestructive examination ix
NEA Nuclear Energy Agency NEAC Nuclear Energy Advisory Committee NEAMS Nuclear Energy Advanced Modeling and Simulation NEI Nuclear Energy Institute NEUP Nuclear Energy University Program NRC (U.S.) Nuclear Regulatory Commission NUGENIA Nuclear GENeration II & III Association NUREG NRC Technical Report ORNL Oak Ridge National Laboratory PRA probabilistic risk assessment PWR pressurized water reactor R&D research and development RAVEN Risk Analysis and Virtual Control Environment (simulation controller for risk-informed safety margin characterization)
RELAP Reactor Excursion and Leak Analysis Program (followed by version number)
RIME Radiation Induced Microstructural Evolution (code)
RIMM risk-informed margins management RISA Risk-Informed Systems Analysis RPV reactor pressure vessel SCC stress corrosion cracking SSC systems, structures, and components U.S. United States UWG Utility Working Group x
Light Water Reactor Sustainability Program Integrated Program Plan
- 1. BACKGROUND Nuclear energy is an important contributor to meeting national electricity generation objectives. It provides reliable base-load capacity at historically high availability rates while supporting national greenhouse gas emission goals. The United States (U.S.) commercial nuclear power industry has demonstrated a substantial history of safe operation and serves as a vital element that ensures the stability of the nations electricity grids.
The Light Water Reactor Sustainability (LWRS) Program is a research and development (R&D) program sponsored by the U.S. Department of Energy (DOE), performed in close collaboration and cooperation with industry. The LWRS Program provides technical foundations for the continued operation of the nations nuclear power plants, utilizing the unique capabilities of the national laboratory system. This involves leveraging national laboratory facilities, staff, and expertise to conduct research needed to inform decisions, demonstrate technical solutions, and provide methods needed for the long-term management and operation of nuclear power systems.
Electric power is a vital component of the nations economy and is essential to continuing improvements in the quality of life. Currently, almost 70% of domestic electricity generation relies on fossil fuels. Nuclear energy is the nations largest contributor of electric power generation that does not emit greenhouse gases, comprising over 60% of non-emitting sources (see Figure 1). Energy efficiency, renewable energy, and carbon capture and storage technologies are playing increasing roles in providing clean and reliable energy. Nevertheless, nuclear energy is an essential part of the nations long-term future energy mix, beyond just its ability to reduce greenhouse-gas emissions, for the following reasons:
Fuel source diversity: An appropriate balance of more than one type of energy resource within the electricity supply system is prudent to mitigate short-term scarcity and price volatility Electric supply reliability: An electrical-power supply in the U.S. must be stable in time and space and have an adequate capacity margin Environmental sustainability: This includes minimal free-release emissions, no carbon emissions, small environmental footprints, minimal solid waste, and sustainable water use National security: In order to have a major role in setting international standards for safeguards, physical security, and safety, the U.S. must be a major player in domestic nuclear energy to influence the directions taken worldwide.
The other forms of low carbon dioxide-emitting and renewable energy production methods (e.g., hydroelectric, wind, geothermal, solar) have the potential to produce substantial energy; however, intermittent sources are of limited use for baseload power until energy storage becomes economical.
Hydroelectric power is the most widely used renewable energy source in the U.S.; however, there is limited opportunity for expansion. Wind, geothermal, and solar power are demonstrating promise in meeting the nations growing energy demands, though these sources currently contribute only a small fraction. In addition, wind and solar power are inherently dilute, with low power density, and are intermittent, resulting in low capacity factors. Geothermal is not intermittent, but is limited to locations or regions, such as the geysers in the State of California, where very hot water is easily accessible. Figure 2 provides a graph of the current capacity factors by energy source. The very high capacity factor for nuclear power makes it the most reliable, non-carbon-dioxide-emitting source of baseload power available.
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Figure 1. Current U.S. electricity-generation portfolio, showing dominance of nuclear as a low-carbon-emission power source.
Figure 2. U.S. electrical generation capacity factors by energy source, showing high operating performance.
Construction of new nuclear power plants is a clear option for new electrical generating capacity.
However, bringing new nuclear power plants online meets substantial challenges and uncertainties, including high up-front capital costs, high financing costs, long construction times, and competition from low natural-gas prices and renewable energy sources. Currently, two nuclear power plants are under construction in the U.S.: Vogtle Units 3 and 4. Watts Bar Unit 2 began commercial operation in October 2016, making it the first U.S. reactor to begin commercial operation in the twenty-first century.
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In January 2013, 104 nuclear power plants operated in 31 states. However, since that time, six plants have been shut down (several due to economic reasons), with more additional shutdowns announced to occur by 2025. In fact, in March 2018, FirstEnergythe owner of three nuclear units in Ohio and Pennsylvaniaannounced plans to shut them all down within the next three years. In most of the recent and planned shutdown decisions, the inability to profitably compete in the current environment of inexpensive, abundant natural gas and subsidized renewable energy sources was a major factor. This current situation has significantly increased emphasis on reducing the costs of producing energy from commercial nuclear sources. It has also prompted a national dialogue on the value of existing nuclear power plants and possible electricity market reforms incentivizing the contribution of electricity supplied by nuclear power plants. The Nuclear Energy Institute (NEI) is leading efforts within the industry through an initiative called Delivering the Nuclear Promise,b which began in late 2015 with the goal of a 30%
reduction in electric generating costs by 2018. Because of the clear and continuing economic challenges faced by nuclear power plants, the LWRS Program has refocused some of its R&D efforts and is considering how to leverage the results from other ongoing R&D activities to improve the economic performance of light-water reactors (LWRs) in current and future energy markets.
The six plant closures since January 2013 brings the number of total operating nuclear power plants in the U.S. to 98 (see Figure 3). The existing fleet of U.S. nuclear power plants has continued to maintain outstanding levels of nuclear safety, reliability, and operational performance over the last several decades and operates with an average capacity factor over 90%. In fact, the capacity factors of nuclear power plants have improved from around 50% in the early 1970s to over 90% today, and industry initiatives on continuous improvement maintain that high record of performance. Over the same period of time, the operational safety of nuclear power plants has improved substantially, as measured by predicted core-damage frequency in postulated accident scenarios, and as seen by the reductions in the rates of initiating events and system failures. Significant improvements in performance, reliability, and safety have made nuclear power plants considerably more economical to operate. Major improvements were made in all areas of plant operation, including operations, training, equipment maintenance and reliability, technological improvements, and improved understanding of component degradation. More broadly, these improvements reflect effective management practices, advances in technology, and the sharing of safety and operational experience among utilities.
Figure 4 shows that: 1) the oldest operating nuclear power plant started operation in 1969, while the newest plant received its operating license in 2015; 2) the first group of nuclear power plants were brought online between 1969 and 1979, and the second group between 1980 and 1996; and 3) almost all operating nuclear power plants have been issued, are applying for, or plan to apply for a 20-year license extension. This license extension will result in a licensed operating period of up to 60 years. Note, however, that receiving a license extension doesnt necessarily mean that the plant will continue to operate, as evidenced by the decisions of Dominion to shut down their Kewaunee plant and Entergy to shut down their Vermont Yankee plant prior to entering their license extension periods. Business decisions on extended operation ultimately rely on economic factors; however, economics can often be improved through technical advancements.
- b. Delivering the Nuclear Promise: Advancing Safety, Reliability and Economic Performance, February 2016, NEI.
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Figure 3. National distribution of the 98 operating nuclear power plants in the U.S.
Figure 4. Nuclear power plant initial license date and license extension plans (as of September 2018).
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Near the year 2030, unless second license renewals are granted, decommissioning of the current fleet of nuclear power plants will begin. Over the three decades beyond 2030, decommissioning of the existing fleet would result in a loss of nearly 100-GWe of emission-free electrical generating capacity, leaving a shortfall of emission-free generating capacity. Early (prior to 60 years of operation) shutdowns due to economic factors will increase this shortfall. Hence, the continued safe and economical operation of current plants to and beyond the current license limit of 60 years is an important option for supplying needed electricity and maintaining the existing level of emission-free power generation capability at a fraction of the cost of building new plants.
To receive a 20-year license extension, a nuclear power plant operator must ensure the plant will operate safely for the duration of the license extension. The 40-year initial operating license period established in the Atomic Energy Act was based on antitrust and capital depreciation considerations, not technical limitations. The 20-year license-extension periods are presently authorized under the governing regulation of 10 CFR Part 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants.c This rule places no limit on the number of times a plant can be granted a 20-year license renewal as long as the licensing basis is maintained during the renewal term in the same manner and to the same extent as during the original licensing term (e.g., the licensee can demonstrate continued safe and secure operation during the extended period).
This regulatory process ensures that licensed nuclear power plants can continue to be operated safely and efficiently during future renewal periods. The license-extension process requires both a safety and an environmental review, with multiple opportunities for public involvement. The license extension applicant must demonstrate how the licensee is, or is planning to be, addressing aging-related safety issues through technical documentation and analysis, which the U.S. Nuclear Regulatory Commission (NRC) confirms before granting a license extension. A solid technical understanding of how SSCs age is necessary for nuclear power plant licensees to demonstrate continued safe operation. A well-established knowledge base for the current period of licensed operation exists; however, additional research is needed to obtain the same robust technical basis required for continued operational evaluations beyond 60 years.
In early 2007, DOE, with Idaho National Laboratory (INL) engaging the Electric Power Research Institute (EPRI) and other industry stakeholders, initiated planning that led to the LWRS Program. The aim was to develop an R&D strategy that addresses nuclear-energy issues within the framework of the National Energy Policy and the National Energy Policy Act of 2005. Based on considerable analysis and information gathering, the Strategic Plan for Light Water Reactor Research and Development,d was developed and reviewed by an independent committee of experts. The plan, which recommended ten top-priority areas for an R&D program with costs shared between government and industry was issued in November 2007.
Building on the strategic plan and collaborative relationships that were developed while preparing it, DOE and INL immediately started developing the LWRS Program. In February 2008, DOE and NRC co-sponsored a workshop, which identified necessary R&D for long-term operation and licensing of nuclear power plants.e Participation by industry, along with other stakeholders, provided an important definition of needs and focused program objectives on long-term operation of existing nuclear power plants. A follow-on workshop was held in February 2011 to review progress and discuss challenges with R&D for long-term operation.
- c. 10 CFR 54, Requirements for Renewal of Operating Licenses for Nuclear Power Plants, Code of Federal Regulations.
- d. INL/EXT-07-13543, Strategic Plan for Light Water Reactor Research and Development, INL, November 2007.
- e. Life Beyond 60 Workshop Summary Report, NRC/DOE Workshop U.S. Nuclear Power Plant Life Extension Research and Development, NRC and DOE, prepared by Energetics Inc., February 19 through 21, 2008.
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In developing the strategic plan and the more specific program plans, it became apparent that a government/industry collaborative cost-sharing arrangement for R&D was needed to address the long-range, policy-driven goals of government and the acceptability and usefulness of derived solutions to industry. The national strategic interests in the long-term operation of existing nuclear power plants included meeting energy-diversity objectives, providing energy security, and minimizing cost impacts (due to plant replacements) to rate payers. The nuclear industry also had an incentive to ensure the continued safe and reliable operation of their operating nuclear power plants.
Therefore, at the nexus of these mutual interests, cost sharing is being employed through cooperative R&D activities. DOE and industry are independently funding specific related projects and sharing information to achieve goals of mutual interest. DOE-funded R&D addresses fundamental scientific questions, where private investment or capabilities are insufficient to make progress on broadly applicable technology issues for public benefit. The U.S. government (i.e., DOE and its national laboratories) holds large theoretical, computational, and experimental expertise in nuclear R&D that is not available within the industry. For this reason, benefits will extend to the next generation of reactor technologies being deployed and those still under development.
While industry is likely to invest in applied research programs that are directed toward enhancing operations or in developing incremental improvements, industry is unlikely to invest significantly in research programs that focus on longer-term or higher-risk gains. Additionally, because research necessary for nuclear power plants long-term operation is of a broad nature that provides benefits to the entire industry, it is unlikely that a single company will make the necessary investment on its own.
Government cost sharing and involvement is required to promote necessary programs that are of crucial long-term, strategic importance. The LWRS Program, by incorporating collaborative industry stakeholder inputs and shared costs, supports the strategic national interest of maintaining nuclear power as an available resource.
Decisions on second license renewal and required investments to support long-term operation are made by plant owners. On January 31, 2018, Florida Power & Light filed to renew its licenses for its Turkey Point Nuclear Units 3 and 4 with the NRC that, upon approval, would allow the utility to operate the units until 2052 and 2053, respectively. This follows previous announcements by Dominion Generation Group of second license-renewal submittal plans for its Surry Power Station and by Exelon Corporation for two operating reactors at the Peach Bottom Atomic Power Station in Southeastern Pennsylvania. The LWRS Program has worked with owner-operators to provide the technical basis for second license renewal specifically, as well as to address current and future issues needed to ensure that a long-term viable source of nuclear-power generation remains available to U.S. electricity markets.
The science-based technical results from the LWRS Program provide data, methods, and technologies that are used by owner-operators to make informed decisions and take actions needed to ensure the continued operation of the existing U.S. LWR fleet. Through the variety of R&D activities carried out together with and used by industry, the LWRS Program reduces key uncertainties and risks that many owner-operators face regarding the long-term performance of vital materials, plant modernization, efficiency improvement, and other issues needed to make the investments required for nuclear power plant operation periods to and beyond 60 years.
1.1 Program Overview Sustainability in the context of this program is the ability to maintain safe and economic operation of the existing fleet of nuclear power plants for as long as possible and practical. It has two facets with respect to long-term operations: 1) to provide science-based solutions to the industry to implement technology to exceed the performance of the current labor-intensive business model; and 2) to manage the 6
aging of plant SSCs so that nuclear power plant lifetimes can be extended and the plants can continue to operate safely, efficiently, and economically.
The LWRS Program carries out its mission through a set of three distinct R&D pathways that are summarized below:
Materials Research: R&D to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants. This work will provide data and methods to assess the performance of SSCs essential to safe and sustained nuclear power plant operations. The R&D products will be used to define operational limits and aging mitigation approaches for materials in nuclear power plant SSCs subject to long-term operating conditions, providing key input to both regulators and industry.
Plant Modernization: R&D to address nuclear power plant economic viability in current and future energy markets through innovation, efficiency gains, and business-model transformation through digital technologies. This includes addressing the long-term aging and modernization or replacement of legacy instrumentation and control (I&C) technologies by research, development, and demonstration of new I&C technologies and advanced condition-monitoring technologies for more automated and reliable plant operation. The R&D products will enable modernization of plant systems and processes and a technology-centered business-model platform, which supports improved performance at lower costs.
Risk-Informed Systems Analysis: R&D to optimize safety margins and minimizing uncertainties to achieve high levels of safety and economic efficiencies. The pathway will: (1) deploy the method and tools of technologies that enable better representation of safety margins and the factors that contribute to cost and safety; and (2) conduct advanced risk-assessment applications with industry to support margin management strategies that enable more cost-effective plant operation. The methods and tools provided by the pathway will support effective safety margin management for both active and passive SSCs.
Technical plans for each of the pathways are discussed in Sections 2 through 4. Measurable milestones have been developed for each of the pathways, including both near-term (i.e., 1 to 5 year) and longer-term (i.e., beyond 5 year) milestones. This Integrated Program Plan is updated yearly; a listing of major accomplishments from previous fiscal years can be found in Appendix A.
1.2 Program Management The entire LWRS Program is organizationally aligned within DOEs Office of Nuclear Energy (DOE-NE). Program management and oversightincluding programmatic direction, project-execution controls, budgetary controls, and Technical Integration Office performance oversightare provided by the DOE-NE Office of Nuclear Technology Demonstration and Deployment in conjunction with DOEs Idaho Operations Office (DOE-ID). The functional organization, reporting relationships, and roles and responsibilities for the Technical Integration Office are explained in the following sections and Figure 5.
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Figure 5. Light Water Reactor Sustainability Program organization.
External program review is realized both at the overall program level, as well as at the pathway level.
This has been done in a variety of ways over the lifetime of the program, including via an external review committee that provides feedback to the Technical Integration Office director, as well as periodic reviews by the Nuclear Energy Advisory Committee subcommittees. These provide advice to DOE-NE. Each pathway has informal review groups that provide feedback to the Pathway lead. In 2015, the LWRS Program Technical Integration Office implemented a tiered approach to external review, beginning with a set of between three and ten experts (i.e., the exact number depends upon the size of the particular pathway under review) for each pathway that will review plans and progress. One to three experts from each of the Pathway Review Teams then participates in a review at the Technical Integration Office (program) level. This process is repeated approximately every 18 months, and the feedback will be used to make changes to the program, as agreed upon by the federal program manager.
1.3 Program Research and Development Interfaces Planning, execution, and implementation of the LWRS Program are done in coordination with the nuclear industry, NRC, universities, and related DOE R&D programs (e.g., Nuclear Energy Advanced Modeling and Simulation [NEAMS], Consortium for Advanced Simulation of LWRs [CASL], Nuclear Energy Enabling Technologies, and the Fuel Cycle R&D Program) to assure relevance, efficiency, and effective management of the work.
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The development of the scientific basis to support service operation extensions beyond 60 years and facilitate high-performance, economic operation of the existing LWR fleet over the extended period is the central focus of the LWRS Program. Therefore, coordination with both industry and the NRC is needed to ensure a uniform approach, shared objectives, and efficient integration of collaborative work for the LWRS Program. This coordination requires that articulated criteria for the work appropriate to each group be defined in memoranda of understanding that are executed among these groups.
1.3.1 Industry The LWRS Program works with industry on nuclear-energy-supply technology R&D needs of common interest. The interactions with industry are broad and include cooperation, coordination, and direct cost-sharing activities. The guiding concepts for working with industry are leveraging limited resources through cost-shared R&D, direct work on issues related to the long-term operation of nuclear power plants, the need to develop state-of-the-art technology to ensure safe and efficient operation, and the need to focus government-sponsored R&D on the higher-risk and/or longer-term projects, incorporating scientific and qualitative solutions that use the unique expertise and facilities at the DOE laboratories. These concepts are included in memoranda of understanding, nondisclosure agreements, and cooperative R&D agreements. Cost-shared activities are planned and executed on a partnership basis and include significant joint management and funding. Periodic coordination meetings are held at the program and technical pathway levels to facilitate communication.
EPRI has established the Long-Term Operations program and a Plant Modernization program, which are complementary to the DOE LWRS Program. EPRI and industrys interests include applications of scientific understanding and tools to achieve safe and economical long-term operation of the current LWR fleet. Therefore, government and private-sector interests are similar and interdependent, leading to strong mutual support for technical collaboration and cost sharing. The interface between DOE-NE and EPRI for R&D work supporting long-term operations of the existing fleet is defined in a memorandum of understanding.f 1.3.2 Nuclear Regulatory Commission NRC has a memorandum of understandingg in place with DOE, which specifically allows for collaboration on research supporting the long-term operation of nuclear power plants. Although the goals of the NRC and DOE research programs may differ, fundamental data and technical information obtained through joint research activities are recognized as potentially of interest and useful to each agency under appropriate circumstances. Accordingly, to conserve resources and avoid duplication of effort, it is in the best interest of both parties to cooperate and share data and technical information and, in some cases, the costs related to such research, whenever such cooperation and cost sharing may be done in a mutually beneficial fashion.
1.3.3 International DOE is coordinating LWRS Program activities with several international organizations with similar interests and R&D programs. The LWRS Program participants continue to develop relationships with international partners, including the following international organizations, to gain awareness of emerging issues and their scientific solutions:
- f. Memorandum of Understanding between United States Department of Energy (DOE) and The Electric Power Research Institute (EPRI) on Light Water Reactor Research Programs, dated December 15, 2015, and signed by John E. Kelly, Deputy Assistant Secretary for Nuclear Reactor Technologies, Office of Nuclear Energy, DOE, and Neil Wilmshurst, Vice President Nuclear, EPRI.
- g. Memorandum of Understanding between U.S. Nuclear Regulatory Commission and U.S. Department of Energy on Cooperative Nuclear Safety Research, dated May 01, 2014, and signed by Brian W. Sheron, Director, Office of Nuclear Regulatory Research, NRC, and John E. Kelly, Deputy Assistant Secretary for Nuclear Reactor Technologies, DOE-NE.
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Organisation for Economic Co-operation and Developments Halden Project: The Halden Project is a jointly financed R&D program under the Nuclear Energy Agency (NEA) of the Organisation for Economic Co-operation and Development (OECD) and comprises national organizations in 18 countries, including licensing and regulatory bodies, vendors, utilities, and research organizations.
The Norwegian Institute for Energy Technology executes the program at its Halden establishment in Norway. DOE is an associate member of the Halden Reactor Project.
The OECD forms several committees and working groups within NEA to assist member countries in maintaining and further developing the scientific and technical knowledge base required to assess the safety of nuclear reactors and fuel-cycle facilities. These committees and working groups focus attention on specific areas requiring attention by the international nuclear industry and regulatory community. Currently, the LWRS Program participates in several working groups sponsored by the OECD Nuclear Energy Agency.
Materials Aging Institute (MAI): The MAI was founded as a partnership between Électricité de France, EPRI, and the Tokyo Electric Power Company and is dedicated to understanding and modeling materials degradation. The collaborative interface with the LWRS Program is coordinated through EPRI, a founding member of MAI.
International Atomic Energy Agency (IAEA) Plant Life Management: IAEA is the worlds center of cooperation in the nuclear field and works with its member states and multiple partners worldwide to promote safe, secure, and peaceful nuclear technologies.
Nuclear GENeration II & III Association (NUGENIA): NUGENIA is an international collaborative R&D network of major stakeholders in nuclear-power generation from industry, utilities, research institutions, and technical safety organizations. It was launched in Brussels, Belgium, in 2012, and hosted by the European Commissions Joint Research Centre. NUGENIA aims to provide a single framework for collaborative R&D on nuclear-fission technologies, with a focus on the current fleet of nuclear reactors (known as Generation II and III). It brings together existing nuclear-power generation R&D under a single umbrella, including several European Networks of Excellence, such as Nuclear Plant Life Prediction, Severe Accident Research Network, and the European Network on Inspection and Qualification.
European Network of Excellence Nuclear Plant Life Prediction (NULIFE): The European Network of Excellence Nuclear Plant Life Prediction has been launched under the Euratom Framework Programme, with a clear focus on integrating safety-oriented research on materials, structures, and systems, and using the results of this integration through the production of consistent lifetime assessment methods.
Bilateral Activities: There are several U.S. bilateral activities either underway (e.g., U.S.-Argentina, U.S.-Japan, U.S.-India) or under discussion (e.g., U.S.-Canada) that include activities specific to the LWRS Program. These bilateral activities provide an opportunity to leverage work ongoing in other countries.
1.3.4 Universities Universities participate in the program in at least two ways: 1) through the Nuclear Energy University Program (NEUP); and 2) via direct contracts with the national laboratories that support the LWRS Program. NEUP also funds nuclear-energy research and equipment upgrades at U.S. colleges and universities and provides scholarships and fellowships to students (see https://neup.inl.gov). In addition to contributing funds to NEUP, the LWRS Program provides descriptions of research activities important to the LWRS Program, while the universities submit proposals that are technically reviewed. The top proposals are selected and those universities work closely with the LWRS Program in support of key LWRS Program activities. Universities also are engaged in the LWRS Program via direct subcontracts under which unique capabilities and/or facilities are funded by the program.
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1.3.5 Advanced Modeling and Simulation Tools A common theme for the pathways is use of computer modeling of physical processes or development of a larger-system computer model. Extensive use of computer modeling is intended to distill the derived information so that it can be used for further research and as the basis for decision making.
Computer modeling occurs in three forms, with many synergistic aspects within the LWRS Program:
- 1. Modeling a physical behavior (such as crack initiation in steel) is an example of direct computer modeling. The resulting model is used to store information for use in other pathways and to use in its own right for further research.
- 2. Development of more-detailed computer modeling tools capable of encoding more-complex behaviors (such as predictive component-aging models).
- 3. Creation of larger integrated databases that roll up lower-level results and allow decision making. The large, system-wide, integrated models allow complex behavior to be understood in novel ways and new conclusions to be drawn. These integrated databases can be used to further guide physical and modeling research, improving the entire program.
Because of their overlapping nature and numerous interfaces, these modeling activities tend to be naturally cross-cutting activities between the LWRS Program pathways.
1.3.5.1 Nuclear Energy Advanced Modeling and Simulation A critical interaction of the LWRS Program is with DOEs NEAMS Program. The LWRS Program is leveraging detailed, multiscale, science-based models developed by the NEAMS program. These advanced computational tools, under development in NEAMS, are being used to create a new set of modeling and simulation capabilities, which will be used to better understand the safety performance of the aging reactor fleet. These capabilities are information sources and tools for advancing the goals of the LWRS Program.
1.3.5.2 Department of Energys Energy Innovation Modeling and Simulation Hub The LWRS Program is coordinating with DOEs Energy Innovation Modeling and Simulation Hub managed by CASL. The hub addresses current operational challenges faced by U.S. nuclear utilities and leverages existing modelsincluding models developed by national laboratories and industryas well as develops new models.
A primary initial product of the hub is a sophisticated integrated model of an LWR (i.e., a virtual reactor with focus on modeling the reactor core). The virtual reactor will be used to address issues for existing LWRs (e.g., long-term operation and power uprates). The hub has a series of challenge problems selected principally to demonstrate the capability of the virtual reactor to enable long-term operation and power uprates. Some of these challenge problems may utilize models under development in the LWRS Program (e.g., systems analysis and component aging models) because the legacy tools have computational limitations that make them unsuitable for some of the challenge problems. The LWRS Program will link with CASL models when detailed core modeling is needed for LWRS Program activities.
1.4 Summary DOE-NE directs the LWRS Program and closely coordinates with other agencies, nuclear industry, and international partners to achieve program goals. The LWRS Program Technical Integration Office supports DOE-NE in accomplishing these goals. Technical integration and program execution are 11
accomplished by using facilities and staff from multiple national laboratories, universities, industry, consulting organizations, and research groups both domestically and internationally.
In summary, the electrical-energy sector is challenged to supply increasing amounts of electricity in a safe, dependable, and economical manner and with reduced carbon dioxide emissions. Nuclear energy is an important part of answering the challenge through the long-term safe and economical operation of current nuclear power plants and by building new nuclear power plants. DOE-NE supports a strong and viable domestic nuclear industry and preserves the ability of that industry to participate in nuclear projects here and abroad. The LWRS Program provides, in collaboration with industry programs, the technical basis for extended safe, reliable, and economical operations of the existing commercial fleet of nuclear power plants.
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- 2. MATERIALS RESEARCH 2.1 Background Nuclear reactors present a variety of challenging service environments to materials that serve as SSCs. Many components in an operating reactor must tolerate high-temperature water, stress, vibration, and an intense neutron field. Degradation of materials in this environment can affect component performance and, without accurate predictive knowledge of component lifetime or if degradation is left unmitigated, can lead to unexpected and costly repairs or failure of these components while in service.
More than 25 different metal alloys can be found within the primary and secondary systems, along with additional materials in concrete, the containment vessel, I&C equipment, cabling, and elsewhere. This diversity of material types, challenging environmental conditions, stress states, and other factors make material degradation in a nuclear power plant a complex phenomenon. In simplified form, Figure 6 shows, that many variables have complex and synergistic interactions that affect materials performance in ways that can lead to impacts on plant operation or reduce the safety performance of a nuclear power plant. Furthermore, unexpected failures or, conversely, the unnecessary repair of components due to overconservative estimates of degradation can lead to higher operational costs.
Impact to Plant Operation (Load)
Precursors Ability of the plant to respond to safety impacts (Capacity)
Potential impacts on economics, reliability and safety Figure 6. Complexity of interactions between materials, environments, and stresses in a nuclear power plant and the impact they have on operations.
The continued operation of the existing nuclear power fleet beyond 60 years will place continued demands on materials and components in their in-service environments. Understanding the performance of these materials during these longer periods of operation entails characterization of the materials as they age under the demands of in-service conditions and relating that knowledge to the performance characteristics of the different SSCs. The research conducted through the activities described here is intended to provide data, models, methods, and techniques to inform industry on long-term materials performance. NRC Technical Report (NUREG)/CR-7153, Expanded Materials Degradation Assessment (EMDA), gives a detailed assessment of many of the key issues in todays reactor fleet and provides a starting point for evaluating those degradation forms particularly important for consideration during continued operation. While extending operation will add additional time and neutron fluence, the primary impact will be increased susceptibility to degradation mechanisms. The application of modern materials science to mechanistic studies, and the development of technology and materials are critical to resolve 13
challenging issues in a timely and practical manner that produces results that can be used by industry to monitor, predict, and plan for the effective management of materials in their in-service environments. The latter may include deployment of mitigation techniques to reduce susceptibility of materials to certain degradation modes, advanced repair methods, and an option for new replacement materials with improved performance over those currently in use.
In the past two decades, great gains were realized in techniques and methodologies that can be applied to the nuclear materials problems of today. Indeed, modern materials-science tools (e.g., advanced characterization tools and computational tools) must be employed. While specific tools and the science-based approach can be described in detail for each particular degradation mode, many of the diverse technical topics and information needs in this area can be organized the following key areas:
Measurements of degradation: High-quality measurement of properties in all components and materials is essential to assess the extent of degradation under extended-service conditions, which supports the development of mechanistic understanding and can be used in model validation. Superior data that follows appropriate standards, codes, and quality assurance guidelines is of value to regulatory and industry interests and to academia.
Mechanisms of degradation: Basic research to understand the underlying mechanisms of selected degradation modes can lead to better prediction and mitigation. For example, research on irradiation-assisted stress corrosion cracking (IASCC) and primary-water stress corrosion cracking will be beneficial for evaluating performance for extended lifetimes and could build on existing aging management programs within EPRI and NRC.
Modeling and simulation: Improved modeling and simulation efforts have great potential in reducing the experimental burden for long-term operational studies. These methods can help interpolate and extrapolate data trends for extended life. Simulations predicting phase transformations, radiation embrittlement, and swelling over component lifetimes would be extremely beneficial to licensing and regulation in extended service. Furthermore, with the development of advanced multiphysics computing codes, complex engineering-scale assessments of structural-component durability and capacity for safety can be made on the basis of these materials scale-degradation models.
Monitoring: While understanding and predicting failures are extremely valuable tools for the management of reactor components, these tools must be supplements to active monitoring. Improved monitoring techniques will help characterize degradation of core components, cabling, and concrete structures. For example, improved detection techniques for identifying crack-development or material-degradation levels will be invaluable. New nondestructive examination (NDE) techniques may eventually lead to more-precise evaluation of component-condition monitoring and lifetime assessment.
Mitigation strategies: Understanding, controlling, and mitigating materials-degradation processes and establishing a sound technical basis for long-range planning of necessary replacements are key priorities for extended nuclear power plant operations and power uprate considerations. While some components can be replaced, decisions to replace others may not be practical or economically favorable. Techniques such as post-irradiation annealing have been demonstrated to be very effective in reducing hardening and susceptibility to IASCC, along with surface modification, based on initial studies. Water-chemistry adjustments can also be effective in reducing some corrosion problems.
2.2 Research and Development Purpose and Goals Materials research provides an important foundation for licensing and managing the long-term, safe, and economical operation of nuclear power plants. Aging mechanisms and their influence on nuclear power plant SSCs are predictable with sufficient confidence to support planning, investment, and licensing for necessary component repair, replacement, and license extension. Understanding, controlling, and mitigating materials degradation processes are key priorities. Proactive management is essential to 14
help ensure that any degradation from long-term operation of nuclear power plants does not affect the publics confidence in the safety and reliability of those nuclear power plants. The strategic goals of the pathway are to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants and to provide data and methods to assess performance and increase longevity of SSCs essential to safe and sustained nuclear power plant operations.
The LWRS Program, through the Materials Research Pathway, is involved in this R&D activity for the following reasons:
MR Pathway tasks provide fundamental understanding and mechanistic knowledge via science-based research. Mechanistic studies provide better foundations for predictive-tool development and focused mitigation solutions. These studies are also complementary to industry efforts to gain relevant operational data. The U.S. national laboratory and university systems are uniquely suited to provide this information given their extensive facilities, research experience, and specific expertise.
Selected Materials Research Pathway tasks are focused on development of high-risk, high-reward technologies to understand, mitigate, or overcome materials degradation. This type of alternative-technology research is uniquely suited to government roles and facilities. These pursuits may be outside the area of normal interest for industry sponsors due to the risk of failure.
MR Pathway tasks support collaborative research with industry and regulators (and meet at least one of the above objectives). The focus of these tasks is on supporting and extending industry capability by providing expertise, unique facilities, or fundamental knowledge.
Combined, these thrusts provide high-quality measurements of degradation modes, improved mechanistic understanding of key degradation modes, and predictive-modeling capability with sufficient experimental data to validate these tools. They produce new methods of monitoring degradation and develop mitigation techniques to improve performance, reliability, and plant economics.
This information must be provided in a timely manner to support long-term operation. Early research within the Materials Research Pathway was focused primarily on providing mechanistic understanding, predictive capabilities, and high-quality data to inform decisions and processes of both industry and regulators. The Materials Research Pathway is taking the knowledge learned and is beginning to transition towards predictive models, and on alternative technologies to overcome or mitigate degradation. Task outputs and products are designed to inform operational decisions. Specific products and impacts will be discussed in the following sections.
2.3 Pathway Research and Development Areas The Materials Research Pathway activities have been organized into five principal areas: 1) reactor metals, 2) concrete, 3) cables, 4) buried piping, and 5) mitigation strategies. These research areas address key material issues in SSCs that were designed for service without replacement throughout the original life of the plant. Management of long-term operation of these components can be difficult and expensive.
As nuclear power plant licensees seek approval for extended operation, the way in which material systems and components age is evaluated, and their capabilities are reassessed to ensure they maintain the ability to perform their intended functions in a safe and reliable manner. Methods for reducing susceptibility to various forms of damage and repair methods to prolong operation are also analyzed.
Identifying, formulating, and prioritizing all of these competing needs have been done in a collaborative manner with industry and regulatory organizations beginning with a workshop focusing on materials aging and degradation held in 2008. From this beginning, technical experts representing broad institutional experience identified and prioritized an initial list of research tasks establishing the organization and structure of the Materials Research Pathway. Research since that workshop has identified additional needs, and these new research topics have also been considered, with further 15
collaborative efforts culminating in the 2014 edition of NUREG/CR-7153. Continued dialogue with EPRI, NRC, vendors, utilities, and other institutions around the world has helped prioritize these emerging needs in the Materials Research Pathway research.
2.3.1 Assessment and Integration The objectives of this research task are to provide comprehensive assessments of materials degradation, relate these to consequences for SSCs and economically important components, incorporate results, guide future testing, and integrate with other pathways and programs. This task also works with industry through hosting workshops, meetings, and individual discussions to identify emerging research and analysis needs for improving plant efficiency and long-term operational planning regarding materials performance. This task also works within the program to support the release of information regarding pathway activities, accomplishments, and external reviews to ensure that funded projects are of value to industry stakeholders. Furthermore, this task provides an organized and updated assessment of key materials-research issues with collaborators through road mapping of research plans and contributing to the technical discussions with EPRI to update the Materials Degradation Matrix documents.
2.3.2 Reactor Metals Many metal alloys can be found throughout the primary and secondary systems of nuclear power plants. Some of these materials (in particular, the reactor internals) are exposed to high temperatures, water, and neutron flux, creating degradation mechanisms in the materials that are unique to reactor service. Research projects in this area will provide a technical foundation to establish the ability of those metals to support extended operations. These projects are a combination of experimental and theoretical work that is used in tandem to develop a greater depth of knowledge of the mechanisms of degradation that influence long-term properties of materials. These experimental and simulation models developed from this research are often supported by the analysis of materials collected from nuclear power plants.
2.3.2.1 Performance of reactor pressure vessel steels To ensure that commercial nuclear power plants can be safely and reliably operated for extended service periods, it will be necessary to demonstrate that the reactor pressure vessels (RPVs) for those plants can maintain adequate safety margins against radiation-induced embrittlement. Unfortunately, limited plant surveillance data are available for understanding the effects of irradiation on RPV steels, including from increased neutron fluences (total fast neutron exposure) associated with extended reactor service; therefore, research relies on accelerated, or higher flux (rate of neutron irradiation), studies. The RPV research conducted in the LWRS Program follows the methodology of utilizing experimental test data, models, and the examination of harvested materials to develop an improved understanding and prediction of materials performance out to extended operating lifetimes. Experimental test methods provide knowledge of fluence effects while modeling efforts are employed to address flux-related issues that are difficult to evaluate experimentally. Examination of harvested materials provides data to benchmark models, and to evaluate the variability among materials in current use. All these efforts are directed at developing improved models to measure changes in the steel as it ages and predict behavior more accurately under transient and off-normal operating conditionsspecifically, pressurized thermal shock.
Experimental RPV Research Aging of RPV steels results in radiation-induced hardening, manifested as increases in the ductile-brittle fracture temperature (T) for the duration of a plants service life. A primary objective of the LWRS Programs research is to develop a robust physical model to accurately predict transition temperatures at high fluence (i.e., at least 1020 n/cm2, E >1 MeV) for vessel-relevant fluxes pertinent to extended plant operations. The Advanced Test Reactor (ATR)-2 project involves experimental testing of representative and archival heats of steel used in RPV construction to examine the mechanical properties and microstructural changes associated with extended plant operations. The ATR-2 experimental work 16
(Figure 7) bridges previous test reactor and surveillance data for insight on the effects of flux and fluence through the experimental testing of 172 different RPV alloys with systematic variations in compositions.
This work involves collaborations with numerous laboratories, industry partners, and organizations. The data collected from the ATR-2 project are used with previous database information on RPV performance to develop an experimentally derived reduced order model for T. This work is to be completed in 2019.
Computational Modeling of RPV Properties In addition to the experimentally derived model, two additional approaches for determining radiation-induced hardening of RPV steels have been developed and are being further evaluated as the ATR-2 data examination is completed. The first is a mechanistic approach using cluster-dynamics theory to predict the formation of Cu-rich and Mn-Ni-Si precipitate volume fractions, from which mechanical properties can be determined through well-established correlations. The cluster-dynamics model has been benchmarked against current experimental database information and will be later incorporated into engineering-scale multiphysics modeling efforts through the use of the Grizzly modeling tool (to be discussed in the next section). The cluster-dynamics model is particularly useful in addressing flux effects over long aging periods that are difficult to experimentally evaluate. The second modeling approach uses a machine-learning algorithm to predict RPV performance. In this case, previous RPV database information (Irradiated Variables [IVAR] dataset) was used to train the machine-learning model to predict hardening of the ATR-2 samples (Figure 7). The lessons learned from this technique will be useful in the further application of machine learning to other areas of the program, such as the prediction of stress corrosion cracking. Current RPV modeling efforts are now being directed toward evaluating the stability of Cu-rich and Mn-Ni-Si precipitates, developed during aging, that affect hardening in order to establish an understanding of the mechanisms of their formation and the parameters necessary for annealing, to be used as a method for mitigating embrittlement.
Figure 7. Experimental database sets on RPV alloys relative to surveillance and vessel-service data as a function of flux and fluence (left), and (right) the comparison of experimental ATR-2 hardening (y) data to those predicted using machine learning using IVAR and previous datasets for fitting.
Engineering-scale Assessment of RPV Performance The RPV plays an essential safety role, and its integrity must be ensured during a variety of transient loading conditions. These can include off-normal conditions such as a pressurized thermal shock (PTS),
as well as transients encountered during normal startup, shutdown, and testing of the reactor. As long-term operation scenarios are being considered for LWRs in the U.S., it is important to have a flexible simulation tool that can both be used to perform probabilistic evaluations of RPV integrity under a wide variety of conditions and incorporate improved predictive models of RPV steel embrittlement. The multi-physics Grizzly-based model addresses all aspects of RPV performance to provide a comprehensive 17
simulation tool that accounts for aging effects on materials properties and the global thermomechanical response of the RPV to loading and that provides fracture analysis of pre-existing flaws and their potential for crack propagation in calculating the probability of vessel failure under postulated accident scenarios.
The first version of the Grizzly RPV model (Figure 8) was released in 2018 and benchmarked against the latest version (16) of the Fracture Analysis of Vessels - Oak Ridge (FAVOR) code developed by the NRC. The Grizzly version provides a 3D evaluation of flaws and will be expanded in future years to include more refined analysis of flux calculations, particularly outside the beltline, as well as take advantage of updated models for T (from LWRS research) that does not under predict RPV hardenability at high fluences in the manner that current regulatory modelssuch as the Eason, Odette, Nanstad, and Yamamoto (EONY) and ASTM E900 modelsdo.
Figure 8. Results of 1D axisymmetric, 2D planar, and 3D Grizzly models of the global response of an RPV at a point in time during a PTS event.
Master Curve Development Though RPV surveillance programs provide advanced information on RPV performance, they have two deficiencies. The first is that current surveillance data do not extend to fluences required for end-of-license performance, thus requiring the use of higher-flux experimental reactor irradiations. The second limitation is that while RPV surveillance programs provide information that is useful for evaluating embrittlement, there is a potential for bias in the fracture-toughness data derived from Charpy V-Notch (CVN) tests. The CVN specimens, a common test-specimen geometry in surveillance capsules, provides measurements that are qualitative and must be correlated to fracture-toughness and crack-arrest-toughness properties for structural integrity evaluations. Recent work examined the use of miniature compact tension (MCT) fracture-toughness specimens that can be machined from broken halves of tested Charpy V-notch impact bars (Figure 9). The testing of MCTs from Charpy specimens will allow the determination and monitoring of actual fracture toughness, instead of indirect predictions using Charpy specimens. Multiple MCTs can be fabricated from a single Charpy specimen, providing additional information on test and material variability. Recent work confirmed the validity of the MCT testing in non-irradiated and irradiated RPV weld material. These data, along with confirmatory testing by the Central Research Institute of Electric Power Industry, to be completed in 2019, will be used towards the modification of ASTM E1921 standard to accommodate the use of MCTs in the determination of the fracture toughness transition, or Master Curve approach used in evaluating RPVs.
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Figure 9. Example of the size dimensions of the MCT test specimen to that of a conventional Charpy V-notch test bar (shown as an already broken/tested bar) and the comparison of MCT data to other compact tension-test geometries on the Master Curve for Linde-80 weld metal from the Midland reactor.
The major milestones associated with the RPV related tasks are By 2019, complete experimentally derived, reduced-order model for transition temperature shift in RPV steel By 2019, complete confirmatory testing by external collaborators of miniature compact tension specimens as a technique for RPV Master Curve determination.
In 2020, deliver validated model of the mechanisms of high-fluence precipitation in RPV alloys In 2021, deliver validated model for precipitate stability in annealed RPV steels associated with mitigation of embrittlement effects By 2021, complete the three-dimensional modeling of fracture propagation using the extended finite element method in Grizzly In 2021, integrate an embrittlement model for high-fluence conditions that includes Mn-Ni-Si precipitate-phase development into Grizzly By 2023, complete testing and analysis of the Zion RPV materials, compare with performance models and evaluate the materials with regard to safety margins.
Future milestones and specific subtasks will be based on the results of the previous years testing as well as ongoing, industry-directed research. Furthermore, future research will explore mitigation techniques, detailed in later sections (Section 2.3.6.3). The experimental database generated supports increasingly complex model development, culminating in engineering-scale models that are of value to both industry and regulators. Completion of data acquisition and model development to permit prediction of embrittlement in RPV steels at high fluence is a major step in informing long-term operational decisions, and high-quality data can be used to inform operational decisions for the RPV by industry. For example, data and trends will be essential in determining operating limits. The data will also allow for extension of regulatory limits and guidelines to extend service conditions.
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2.3.2.2 Mechanisms of irradiation-assisted stress corrosion cracking The objective of this research is to evaluate the mechanisms of IASCC in austenitic stainless steels, and to apply that knowledge towards predictive models for component degradation and the development of techniques for mitigating IASCC susceptibility. Experiments of increasing complexity, starting with single-variable testing, will be used to isolate specific effects of material and environment conditions.
Research will include crack-initiation and crack-growth tests in simulated LWR environments and include microstructural and microchemical analysis. New in situ characterization techniques are being applied to examine localized deformation mechanisms in irradiated test materials that result in the generation and propagation of cracks.
The dependence of water chemistry on the corrosion and increased susceptibility to IASCC initiation and crack growth rates are also part of the LWRS set of activities. Techniques are being used to evaluate the differences between LiOH and KOH water chemistry of primary water systems (see Figure 10). These evaluation methods are applied to determine surface-chemistry effects and coupled with microstructural analysis techniques to evaluate changes in corrosion susceptibility of the material at local features, such as grain-to-grain boundaries, inclusions, and the effects induced by localized deformation observed in irradiated materials that have been stressed.
The susceptibility of 304 and 316 stainless steel to IASCC is expected to become more severe with increased exposure to LWR environments. Long-term service will result in very high accumulations of radiation damage that may be manifested in changes in microstructure, such as void formation and different phase fractions than present at lower-damage conditions. Research continues in the program with examining crack initiation as a function of stress for austenitic stainless steels irradiated up to 125 dpa.
Combined, the experimental approaches being taken will provide mechanistic understanding that can be used to identify key operational variables to mitigate or control IASCC, optimize inspection and maintenance schedules to the most susceptible materials and locations, and in the long term, design IASCC-resistant materials.
The major milestones associated with this task are By 2020, deliver probabilistic model for IASCC initiation of austenitic steel to high fluence By 2023, deliver mechanistic-model tool on the critical applied stresses that initiate irradiation assisted stress corrosion cracking.
Completing research to identify the mechanisms of IASCC is an essential step to predicting the extent of this form of degradation under extended service conditions. Understanding the mechanism of IASCC will enable more-focused material inspections and more-accurate decisions on materials replacement. In the long term, mechanistic understanding also enables the development of a predictive model.
Furthermore, the research knowledge gained can be applied to developing new, alternative replacement alloys for those conventionally used in reactor designs.
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Figure 10. Experimental results of atomic-force microscopy and schematic diagrams of the differences in corrosion-layer development between LiOH and KOH water chemistry.h The formation of the passivating oxide layer is inhibited by the presence of hydrated Li+ ions. These ions undergo dehydration on the surface, followed by preferential adsorption of -OH from water contained in the electrical double layer (EDL). This action results in the perturbation of the latter, surface acidification, and formation of a defective oxide film that provides less substrate protection from corrosion.
2.3.2.3 Cracking initiation in nickel-base alloys The objective of this task is the identification of underlying mechanisms of stress corrosion cracking (SCC) in Ni-base alloys. Understanding and modeling the mechanisms of crack initiation is a key step in predicting and mitigating SCC in the primary and secondary water systems. This effort focuses on SCC crack-initiation testing of Ni-base Alloys 600, 690 and 52/152 weld material in simulated LWR water chemistries and is complementary to NRC- and EPRI-sponsored research on Alloy 600 and Weld Alloy 82/182. The objective of the LWRS research is to identify the mechanisms controlling crack nucleation and to investigate the transition from short to long crack growth in Ni-base alloys under realistic LWR conditions to help establish the framework to effectively model and mitigate SCC initiation processes.
The approach is to investigate the important material (e.g., composition, processing, microstructure, strength) and environmental (e.g., temperature, water chemistry, electrochemical potential, stress) effects on the SCC susceptibility of nickel-base alloys. Research to date has identified the three critical stages that promote SCC initiation in alloy 600 and identified the development of microstructural precursors to crack initiation in Alloy 690 (Figure 11).
- h. R.P. Giron, X. Chen, E. Callagon La Plante, M. Gussev, K.J. Leonard, G. N. Sant, Revealing how alkali cations affect stainless steel passivation in alkaline aqueous environments, submitted to American Chemical Society, Applied Materials and Interfaces (May 2018). Also available in LWRS Milestone report number, M2LW-18OR0402025, August 2018.
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Figure 11. Schematic examples of the different stages of crack initiation occurring in Alloy 600 (top) and Alloy 690 (bottom).
The major milestone associated with this task is to deliver, by 2020, a predictive-model capability for Ni-base alloy SCC susceptibility.
Completing research to identify the mechanisms and precursor states to SCC initiation is an essential step to predicting the extent of this form of degradation under extended-service conditions. Understanding underlying causes for crack initiation may allow for more-focused material inspections and maintenance, new SCC-resistant alloys, and development of new mitigation strategies, all of which are of high interest to the nuclear industry. This mechanistic understanding may also drive more-informed regulatory guidelines and aging-management programs.
2.3.2.4 Computational modeling of aging effects The development and phase evolution occurring in structural metal alloys during long-term operation in LWR-relevant conditions can be different from those developed under accelerated-aging conditions experienced in high-flux experimental reactors. Modelling of the development of these phases is critical in determining overall performance, including potential embrittlement, radiation-induced degradation, and susceptibility to corrosion-related damage. This task has provided detailed microstructural-based analysis through modeling and confirmatory experimental evaluations of phase transformations in key samples and components (in both age-accelerated and service-aged materials). The results of the task were used to develop and validate computational models of phase transformations under LWR operating conditions of austenitic stainless steel during 2018.
2.3.2.5 High-fluence swelling of core internal materials The Radiation Induced Microstructural Evolution (RIME) code, completed in 2018, is a modern, mean field, cluster-dynamics approach developed to investigate the material and irradiation parameters that control swelling under the exposure conditions that are representative of the LWR core-operating environment. Validated against high-flux experimental reactor test data, the further comparison to ex-22
service or harvested material of high fluence is required. The development of the RIME code is an important step in estimating the useful life of core-internal components. Current research involves evaluating the feasibility to import this code to larger, engineering-scale models to assess changes to in-core structures and components of commercial reactors as they age, making this a more valuable tool for industry and regulators to use.
2.3.2.6 Environmentally assisted fatigue The objective of this task is to model environmentally assisted mechanisms for fatigue through a mechanistic-based approach supported by experimental studies to develop a finite-element-based fatigue model. This will provide a capability to extrapolate the severity of fatigue degradation under realistic reactor-environment loading cycles and under multi-axial stress states. The experimental data will inform regulatory and operational decisions while the model will provide a capability to extrapolate the severity of this mode of degradation to extended-life conditions. In 2017, a thermomechanical model for Grade 508 low-alloy steel was developed to assess thermal fatigue associated with RPV and pressurizer cycling. In 2018, research was completed on a hybrid-model, data-based framework for deterministic and probabilistic fatigue life estimation of safety-critical nuclear reactor components, specifically examining a pressurized water reactor (PWR) surge line under design-basis and grid-load-following loading conditions. That model accounts for thermal stratification of the fluid flow during transient operations resulting in an induced thermomechanical response of the surge-line pipe. The hybrid modeling strategy shows promise of predicting probabilistic life without conducting hundreds of costly fatigue tests relevant to a given loading and environmental condition (Figure 12). However, this modeling strategy is at an incipient stage and requires further improvement and experimental validation. Research in 2019 continues with experimental testing of the dissimilar-metal weld transition between the stainless steel surge-line and low-alloy steel pressurizer.
Figure 12. Estimated Weibull cumulative distribution function (CDF) for fatigue in air (left) and PWR water (right) conditions, comparing modeled (grey lines) to experimental sample sets (black lines and symbols) at different strain amplitudes (a). Demonstrating that the model provides a good correlation and can be used for estimating CDF when experimental data points are not available.
The major milestones associated with this task are In 2020, complete environmental fatigue assessment of dissimilar metal (Alloy 82/182) weldments in the surge-line pipe transition to the pressurizer and the generation of virtual S-N curves for dissimilar-metal weld using variable amplitude fatigue tests in relevant LWR conditions By 2025, examination of environmentally assisted fatigue degradation of irradiated materials.
Completing research to identify the mechanisms of environmentally assisted fatigue to support model development is an essential step to predicting the extent of this form of degradation under extended-23
service conditions. This knowledge has been identified as a key need by regulators and industry.
Delivering a model for environmentally assisted fatigue will enable more-focused material inspections, material replacements, and more-detailed regulatory guidelines.
2.3.2.7 Thermal aging of cast austenitic stainless steels Cast austenitic stainless steel (CASS) and austenitic stainless steel welds (ASSW) are extensively used for many massive components of primary coolant systems in LWRs, including coolant piping, valve bodies, pump casings, and piping elbows. The performance of these materials beyond 40 years is not well defined and is a concern identified in the EMDA. The objective of this research is to provide conclusive predictions for the behavior of CASS and ASSW components in LWR environments by resolving uncertainties in scientific understanding and performance during extended (60+ years) nuclear plant operations. Mechanical and microstructural data obtained through accelerated aging experiments and computational simulation will be the key input for the prediction of CASS behaviors and for the integrity analyses for various CASS components. In 2015, research was expanded to include ASSW as part of the International Nuclear Energy Research Initiative between the U.S. and Republic of Korea, through collaborative efforts with the Korea Advanced Institute of Science and Technology.
The major milestone associated with this task is to complete analysis and simulations of aging of cast austenitic stainless steel components and austenitic stainless steel welds by 2019, with the delivery of a predictive capability for components under extended service conditions.
Completing research to identify potential thermal-aging issues for CASS/ASSW components is an essential step to identifying possibly synergistic effects of thermal aging (corrosion, mechanical, etc.) and predicting the extent of this form of degradation under extended service conditions. Understanding the mechanisms of thermal aging will enable more-focused material inspections and material replacements and more-detailed regulatory guidelines. These data will also help close gaps identified in the EPRI Materials Degradation Matrix and EMDA reports.
2.3.3 Concrete Concrete makes up the largest volume of material used in nuclear power plants and is exposed to varied environmental conditions. Figure 13 provides a schematic view of a typical PWR design to illustrate the amount of concrete used. There are some data on performance through the first 40 years of service, and in general, the performance of reinforced concrete structures in nuclear power plants has been very good. Although the vast majority of these structures will continue to meet their functional or performance requirements during the current and any future licensing periods, it is reasonable to assume that there will be isolated examples where, as a result primarily of environmental effects, the structures may not exhibit the desired durability (e.g., for water-intake structures or as a result of the freezing/thawing damage of containments) without some form of intervention.
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Figure 13. Cut-away of a typical pressurized water reactor, illustrating large volumes of concrete and the key role of concrete performance (source: NRC).
2.3.3.1 Concrete and civil structure degradation Although a number of organizations have sponsored research addressing the aging of nuclear power plant structures (e.g., NRC, the NEA and IAEA), there are still several areas where additional research is needed to demonstrate that the structures will continue to meet functional and performance requirements (i.e., that they can maintain structural margins). The EMDA, NUREG/CR-7153, provided a list of research priorities addressing second license renewal. Along with irradiated concrete, the effects of alkali-silica reaction in nuclear structures are the focus of the Materials Research Pathway.
The objective of this task is to provide data and information in support of continuing the service of safety-related nuclear power plant concrete structures past their initial 40-year design life. In meeting this objective, pathway activities have included the initial release, in 2017, and continued development of the Irradiated Minerals Aggregate and Concrete (IMAC) database, evaluation of long-term effects of elevated temperature and irradiation, improved damage models, development of improved constitutive models and analytical methods for evaluation of non-linear response, investigation of non-intrusive inspection methods for thick reinforced-concrete sections, and formulation of structural reliability methodology to address time-dependent changes in concrete structures and evaluate how aging affects structural reliability.
Part of the recent research on evaluating the impact of radiation damage on concrete has involved the initial development of the microstructure-oriented scientific analysis of irradiated concrete (MOSAIC) tool to assess the susceptibility of plant-specific concrete damage due to radiation-induced structural degradation.i The MOSAIC tool (Figure 14) takes input from structural analysis using a combination of ellipsometry and X-ray fluorescence that, combined, provide identification of mineral make-up of the
- i. A. Giorla, Development of non-linear Fast Fourier Transform (FFT) solvers for the simulation of irradiation induced expansion in concrete, ORNL report, ORNL/SPR-2017/578, LWRS milestone report M2LW-17OR0403013, March 2018.
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aggregate. Processing of the structural information using the IMAC database of irradiation induced changes in properties as a reference, the MOSAIC tool provides an assessment of the sensitivity of concrete to radiation-induced damage.
Diattenuation Intensity Map (2MGEM)
Damage Feldspar Quartz Cement Nonlinear Calcite FFT solver Mineral Phase Simulated Distribution Image Irradiation Damage Using MOSAIC Silicon Map (XRF)
Figure 14. Process for the MOSAIC tool to assess concrete susceptibility to radiation-induced damage starting from the structural inputs of two modulator generalized ellipsometry microscopy and X-ray fluorescence that are developed into a mineral phase-distribution image prior to passing through a non-liner fast Fourier transformation solver to simulate the damage generated in the concrete aggregate structure.
The major milestones associated with this task are By 2020, the validation of radiation-induced volumetric expansion model in the MOSAIC tool By 2021, complete a model tool (at engineering scale) to assess the combined effects of internal expansion and constraint on structural performance for concrete components By 2022, complete development of a tool kit to evaluate the effects of radiation on concrete under various experimental conditions and concrete mixes found in specific commercial nuclear power plants.
Future milestones and specific tasks will be based on the results of the previous years testing as well as ongoing, industry-led research. This effort will work toward the development of a methodology and documentation that provides risk-informed guidelines for evaluation of the performance of aging safety-related concrete SSCs for use in current and future condition assessments taking into account service conditions and environmental factors that might diminish the residual life of these structures during potential future design-basis events.
2.3.3.2 Alkali-silica reactions in concrete structures The goal of this project is to study the development of alkali-silica reaction (ASR) expansion and induced damage of large-scale specimens that are representative of structural concrete elements found in nuclear power plants. This will be done through experimentally validated models that explore the 26
structural capacity of ASR-affected structures, like the biological shield and containment and fuel-handling buildings. Experiments have been conducted in accelerated conditions, employing extensive monitoring and nondestructive techniques, to evaluate structural stresses generated in the large-block test specimens. Final destructive testing has begun on the large test blocks to address the question of the residual shear capacity. This project will benefit from the experience and knowledge gathered from the RILEM international committee on the prognosis of ASR-affected structures.
The major milestone associated with this task is to complete, by 2021, a model tool to assess effects of ASR on structural performance of concrete components.
Future milestones and specific tasks will be based on the results of the previous years testing as well as ongoing, industry-led research. This effort will work toward the development of a methodology and documentation that provides risk-informed guidelines for evaluation of the performance of ASR affected structural concrete elements. Information obtained during this research will also benefit the understanding of condition-monitoring predictions through nondestructive techniques.
2.3.3.3 Nondestructive evaluation of concrete and civil structures Developing new techniques that allow for condition monitoring of concrete structures and components is the objective of this research. This effort includes performing a survey of available samples, developing techniques to perform volumetric imaging on thick reinforced-concrete sections, determining physical and chemical properties as a function of depth, developing techniques to examine interfaces between concrete and other materials, developing acceptance criteriamodel and validation and developing automated scanning systems. This task is collaborative with the Plant Modernization Pathway.
Recent developments have focused data processing techniques of linear array ultrasonic testing, such as the Model-Based Image Reconstruction (MBIR) and synthetic aperture-focusing technique (SAFT). An example of the results from consecutive signal-processing iterations of adjusted signal parameters is shown in Figure 15, where a deeply buried flaw within the concrete is detected.j Over the past year, effort has been applied to decreasing the processing time of the MBIR signals, towards the goal of real-time analysis with the objective of developing an effective concrete NDE prototypes system.
The major milestones associated with this task are By 2021, complete a prototype (specific hardware NDE technique and supporting software system) concrete NDE system By 2022, complete a field-testing evaluation of concrete NDE system.
The development of NDE techniques to permit monitoring of the concrete and civil structures could be revolutionary and allow for an assessment of performance that is not currently available via core drilling in operating plants. This would reduce uncertainty in safety margins and is valuable to both industry and regulators.
- j. D. Clayton and D. Bull-Ezell, unpublished data (2018).
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Figure 15. Linear array ultrasound data collected on a thick specimen containing intentional flaws. Signal processing, using SAFT, of a given flaw is compared to that of the MBIR forward model showing improved imaging of a defect.
2.3.4 Cabling Understanding cable-aging mechanisms resulting in changes to cable performance and improved means to accurately assess these property changes is an important area of study to ensure the safe and efficient operation of nuclear power plants (Figure 16). This effort also provides plant operators the necessary information to conduct more-accurate and cost-effective inspections in determining when mitigation or replacement is required. Degradation of these cables is primarily caused by long-term exposure to high temperatures, though synergistic effects with irradiation and moisture may produce additional concerns for long-term use. While wholesale replacement of cables is economically undesirable, incorporating more-accurate condition-monitoring techniques is a strategic investment in continuing safe and reliable operation.
Figure 16. Diagram of the technical approach to cable-aging studies towards deployable NDE methods for determining remaining useful life.k
- k. R. Duckworth, E. Frame, S. Davis, L. Fifield and S. Glass, Benchmark Accelerated Aging of Harvested Hypalon/EPR and CSPE/XLPE Power and I&C Cables in Nuclear Power Plants, presentation at the 24th International Conference on Nuclear Engineering, July 27th, 2016.
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2.3.4.1 Mechanisms of cable insulation aging and degradation This task provides an understanding of the role of material type, history, and the environment on the degradation of cable insulation; understanding of accelerated testing limitations; and support to partners in modeling activities, surveillance, and testing criteria. This task will provide experimental characterization of key forms of cable and cable insulation in a cooperative effort with NRC and EPRI.
Tests will include evaluations of cable integrity following exposure to elevated temperatures, humidity, and ionizing irradiation. This experimental data will be used to evaluate mechanisms of cable aging and determine the validity or limitations of accelerated-aging protocols. The experimental data and mechanistic studies can be used to help identify key operational variables related to cable aging, optimize inspection and maintenance schedules to the most susceptible materials/locations, and, in the long range, design tolerant materials.
The major milestones associated with this task are By 2020, complete an assessment of potential cable degradation mitigation strategies In 2021, deliver a predictive model for cable degradation By 2023, develop an assessment of aging on reliability of splices and connections.
Future milestones and specific tasks will be based on the results of the previous years testing, as well as ongoing, industry-led research. Completing research to identify and understand the degradation modes of cable insulation is an essential step to predicting the performance of cable insulation under extended-service conditions. These data are critical to develop and deliver a predictive model for cable-insulation degradation. Both will enable more focused inspections and material replacements and better-informed regulations. The development of in situ mitigation strategies may also allow for an alternative to cable replacement and would be of high value to industry by eliminating costly replacements.
2.3.4.2 Nondestructive evaluation of cable insulation The objectives of this task include the development and validation of new NDE technologies for the assessment of cable-insulation integrity. Degradation of the cable jacket, electrical insulation, and other cable components is a key issue that is likely to affect the ability of the currently installed cables to operate safely and reliably for extended periods. The tracking of cable degradation through NDE techniques requires significant development to understand the abilities and limits of various techniques to track cable degradation against established assessment techniques that involve destructive methods (i.e.,
elongation at break) and requires the further development of advanced signal processing to track changes associated with degradation that is benchmarked against known experimental data. In 2017, a significant effort was completed on identifying promising NDE techniques for both local and long-length evaluation of cables. Current research involves developing more promising NDE methods for cable insulation and modelling NDE signals to provide predictions of remaining useful life. Current research techniques include frequency-domain reflectometry and dielectric spectroscopy.
The major milestones associated with this task are By 2020, complete physics-based model research of NDE signal evaluation for the assessment of cable health By 2021, develop acceptance criteria and usage guidance for cable NDE.
Reliable NDE and in situ approaches are needed to objectively determine the suitability of installed cables for continued service. The ultimate goal of this research is to provide guidance for utilities and regulators leading to more-robust cable-aging management programs that can assure in-service cable integrity under the anticipated design-basis events.
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2.3.5 Buried Piping Maintaining the integrity of many miles of buried piping at a nuclear power plant is necessary for extended plant operations. While much of the buried piping is associated with either the secondary side of the plant or other non-safety-related cooling systems, some buried piping serves a direct safety function.
Maintaining the integrity of the buried piping in these systems is necessary to ensure the systems can continue to perform their intended functions under extended plant service periods. Industry and regulators are already performing considerable work in this area, with challenges existing in the area of NDE techniques deployable to assess pipe wall integrity. The Materials Research Pathway is supportive of research in the Plant Modernization Pathway and at EPRI through participation in expert panel meetings on the subject. The Materials Research Pathway continues to evaluate the subject for gaps and research needs.
2.3.6 Mitigation Technologies Mitigation technologies include weld repair, post-irradiation annealing, water-chemistry modifications, and replacement options for the use of new materials with reduced susceptibility to various modes of degradation. The purpose of this research area is to evaluate mitigation technologies that will reduce susceptibility to degradation and will provide beneficial economic impact through less repair or replacement needs.
2.3.6.1 Advanced weld repair The objective of this task is to develop advanced welding technologies that can be used to repair highly irradiated reactor internals without helium-induced cracking in the heat-affected zonea mode of damage controlled by the level of heat input during welding, residual stresses developed, and irradiation level of the material. As shown in Figure 17, radiation-induced transmutation of He in reactor materials presents significant challenges associated with weld repair that makes current technologies unsuitable.
This joint research effort with EPRI began the evaluation of advanced welding techniques (laser and friction stir welding) on the weld repair of alloys with helium contents typical of highly irradiated reactor materials. This work further builds upon the Integrated Computational Welding Engineering (ICWE) tool, completed in 2016 and undergoing patent application, that evaluates residual stresses developed during welding. Applying the ICWE technology to laser welding has the potential of reducing helium induced cracking in irradiated materials. Identifying the limits of different weld techniques, based on heat input and residual stress in the work piece affecting helium-induced cracking, is the basis of the 2019 Nuclear Energy University Program proposal call. Industry stakeholders can use these models and weld process-development studies to further improve the practices for repair of irradiated materials.
The major milestones associated with this task are In 2021, complete process optimization of weld parameters for irradiated stainless steel By 2022, optimize welding of Ni-base Alloy 182 and demonstrate friction stir cladding of irradiated base metals By 2024, complete SCC testing of weld-repaired material In 2026, complete aging (reirradiation) of weld-repaired irradiated materials.
Demonstration of advanced weldment techniques for irradiated materials is a key step in validating this mitigation strategy. Successful deployment may also allow for an alternative to core internal replacement and would be of high value to industry by avoiding costly replacements. Further, these technologies may also have utility in repair or component-replacement applications in other locations within a nuclear power plant, where residual stresses developed on welding can influence stress corrosion crack susceptibility.
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Figure 17. (left) Forecast of helium generation at 75 wppm boron at 60 EFPY. Red Zone: >10 appm He (not weldable with current welding processes); Yellow Zone: 0.1 to 10 appm He (weldable with heat-input control during welding repair); Green Zone: <0.1 appm He (No special process control is needed in welding repair). (Source: EPRI report, EPRI BWRVIP-97A). (right) Inside view of the laser and friction stir weld subsystems installed in the hot-cell cubicle for testing irradiated materials.
2.3.6.2 Advanced replacement alloys Advanced replacements alloys for use in LWR applications may provide greater margins of safety and performance and provide support to industry partners in their programs through more economical operations. Void swelling, IASCC susceptibility, and decreased fracture toughness are the major concern at high levels of radiation damage. However, most in-core structures consist of austenitic stainless steels, which are susceptible to degradation at a relatively low dose. Thus, replacement of these components may become a necessity. The Advanced Radiation Resistant Materials (ARRM) project was created to address these issues and is a collaboration between the LWRS Program and EPRI. The ARRM project is aimed at identifying promising candidate alloys that can replace conventional low-strength 304L/316L stainless grades and high-strength Ni-base Alloy X-750. Phase I, which finished in 2017, examined the microstructural stability of ion- and proton-irradiated alloys, IASCC behavior of proton-irradiated materials under PWR primary and hydrogen water chemistry (HWC) conditions for boiling water reactor (BWR) plants, fracture-toughness and steam-oxidation testing of over a dozen alloys. Further down-selection to five alloys has occurred, with continued Phase II examinations involving neutron-irradiation testing of Alloys 310, optimized Grade 92, 690, 725 and optimized 718. The ultimate objective of the work is to select at least one low-strength and one high-strength alloy for future reactor core internal support components and fasteners, respectively.
The major milestone associated with this task is to complete, by 2025, the development and testing of low- and high-strength alternative alloys with superior degradation resistance compared to 316L (low strength) and X-750 (high strength).
Completing the Phase I joint effort with EPRI on the alloy down-selection and development plan has been an essential first step in this alloy-development task and provides a better understanding of the susceptibility to degradation of alloys alternative to 304/316 stainless steel and Ni-base Alloy X-750.
Phase II materials continue the ARRM candidate alloy validation through neutron-irradiation testing. The alloys that are emerging from this study offer the potential for greater safety margins and resistance to key forms of degradation at high fluences and long component lifetimes than are seen in the current generation of materials.
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2.3.6.3 Thermal annealing This research provides a critical assessment of thermal annealing as a mitigation technology for RPV embrittlement and core internal susceptibility to IASCC. This task will build on the other RPV tasks and extend the mechanistic understanding of irradiation effects on RPV steels to provide potential mitigation strategies. This task will provide experimental and theoretical support to resolution of technical issues required to implement this strategy. Successful completion of this effort will provide the data and theoretical understanding to support the economic evaluation and potential implementation of this mitigation strategy.
Recent studies have been conducted on the impact of post-irradiation annealing (PIA) treatment on the reduction of crack growth rates in neutron-irradiated stainless steel under BWR water environment and different applied loading conditions. The PIA treatment was found to mitigate cracking susceptibility in 304L stainless steel with 5.9 dpa irradiation damage. Trends show that greater degrees of annealing (time and temperature) led to a decrease in all measures of intergranular stress corrosion cracking (IGSCC) susceptibility (maximum stress, uniform strain, total strain, percent of intergranular attack) changing monotonically with heat-treatment severity. Further research using higher-fluence samples is warranted.
The major milestones associated with this task are By 2020, complete an evaluation of annealing on reducing stress corrosion crack growth in low fluence stainless steel By 2020, establish conditions necessary for post-irradiation annealing through modeling precipitate stability of relevant high fluence RPV alloys By 2023, complete the experimental testing of annealing as a mitigating technique for high-fluence RPV steels By 2026, complete an evaluation of the lasting benefits of annealing high-fluence RPV steels susceptible to embrittlement (re-irradiation of annealed materials).
While a long-term effort, demonstration of annealing techniques and subsequent irradiation for RPV sections is a key step in validating this mitigation strategy. Initial work on PIA of stainless steel has been promising, but requires further evaluation of high-fluence materials. This work is intended to provide industry with data on the targeted conditions required for annealing, including the short- and long-term benefits for annealing RPV or core internals, from which informed decisions can be made based on economic benefits.
2.3.6.4 Water chemistry HWC is another effective strategy in reducing crack growth rates in BWR water conditions; similarly, its effectiveness with materials with high accumulated neutron damage requires further evaluation. In 2018, a collaborative effort was completed between the LWRS Program and Nippon Nuclear Fuel Development Corporation (NFD, Japan) to evaluate the effectiveness of HWC as a crack-growth-rate mitigation technique for BWR materials at high doses. Results from this work on 304L stainless steel showed a reduced influence on crack growth rates at higher fluences, as well as when stress-intensity factors were significantly high at the crack tip. Additional work on addressing the mechanisms responsible for the effect of water chemistry on IASCC susceptibility is being examined for LiOH- and KOH-based water chemistry for PWR primary systems. That work is part of the mechanisms of IASCC work detailed in Section 2.3.2.2.
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2.3.7 Integrated Industry Activities Service materials from active or decommissioned nuclear power plants provide invaluable access to materials for which there is limited operational data or experience to inform license-extension decisions.
In coordination with other materials tasks, these materials allow an assessment of current degradation models to further develop the scientific basis for understanding and predicting long-term environmental degradation behavior. Many radiation-induced phenomenaincluding the generation of void swelling in materials and changes in precipitate volume fraction and distributions, all of which induce mechanical-property changes and affect environmental susceptibilityare largely dependent on flux rate. Therefore, being able to compare low-fluence power-generating reactor-exposed materials to those of experimental reactor data for model validity is important.
The LWRS Program is currently engaged in two key activities that support multiple research tasks in the previous sections: the Exelon (formerly Constellation) Pilot and Zion Harvesting Projects. The Exelon Pilot Project is a joint venture between the LWRS Program, EPRI, and the Exelon Energy Nuclear Group.
The project utilizes Exelons nuclear stations, R. E. Ginna and Nine Mile Point 1, for research opportunities to support extended operation of nuclear power plants. Specific areas of joint research have included baffle former bolt retrieval (an activity that involved the removal of bolts from storage containers in the spent-fuel pool in 2016), development of a concrete-inspection guideline, installation of equipment for monitoring containment rebar and concrete strain, and additional analysis of RPV surveillance coupons. Opportunities for additional and continued collaboration will be explored in coming years.
The Zion Harvesting Project, in cooperation with Zion Solutions, has involved the selective procurement of materials, structures, and components of interest to the LWRS Program, ERPI, and NRC from the decommissioned Zion 1 and Zion 2 nuclear power plants. Materials removed from the plant include low-voltage cabling, and through-wall thickness sections of the RPV, including the beltline weld.
Current research involves the testing of those RPV sections through hundreds of test specimens, along with the evaluation and continued aging of cables to determine end-of-life conditions. Research performed on the Zion RPV sections will address gaps in knowledge on attenuation effects and material variability in RPV cross-sections, provide direct comparison to surveillance and high-flux reactor data, and provide material for future evaluating mitigation techniques such as warm prestressing or annealing and the further effects of re-irradiation to higher total fluence (see Section 2.3.6.3).
The major milestones associated with this task are By 2020, complete microstructural and mechanical evaluation of baffle former bolts By 2023, complete testing and analysis of the Zion RPV materials, comparing them with performance models and evaluating them with regard to safety margins By 2026, complete the study of re-irradiation of Zion material to higher fluence, comparing test data with predictive models.
Discussions regarding continued harvesting of material (including cables, concrete, and RPV samples) are underway. Additional milestones will be identified as samples become available. Samples from the Zion RPV will serve as feedstock for thermal annealing and attenuation studies.
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2.4 Research and Development Collaborations Effective and efficient coordination will require contributions from many institutions, including input from EPRIs parallel activities in the Long-Term Operations Programs strategic action planl and NRCs second license renewal activities (referred to as subsequent license renewal by NRC). In addition to contributions from EPRI and NRC, participation from utilities and vendors will be required. Given the breadth of the research need and direction, all technical expertise and research facilities must be employed to establish the technical basis in this R&D area for extended operations of the current nuclear power plant fleet.
The activities and results of other research efforts in the past and present must be considered on a continuous basis. Collaborations with other research efforts may provide a significant increase in cost sharing for research and could speed research for both partners. This approach also reduces unnecessary overlap and duplicate work. Many possible avenues for collaboration exist, including the following:
EPRI: Considerable research efforts on a broad spectrum of nuclear reactor materials to provide a solid foundation of data, experiences, and knowledge. Cooperation on selected material research activities is reflected in the LWRS Program and EPRIs Long-term Operation Program Joint R&D Plan.m Research plans include joint roadmap development, typically with the NRC on activities that include structural concrete, cable systems, and weld repair.
NRC: Broad research efforts of NRC are considered carefully during task selection and implementation of LWRS research. Cooperative efforts through conduct of the EMDA have identified the research gaps to be addressed for commercial power plant life extension. The NRC is an active collaborator on structural-concrete and cable-systems testing, with further close collaborations being developed for core internals and RPV.
Nuclear companies, utilities, and vendors: The Materials Research Pathway has been engaged with a number of companies at the start and throughout the process of materials test campaigns and work activities towards the programmatic objectives highlighted in this report. Industry has supplied materials, consulted on test plans, contributed time and equipment, and been involved in analyzing test results. Industry-focused workshops and short technical meetings have also been provided.
Central Research Institute of Electric Power Industry: Through the framework of the DOE Civil Nuclear Working Group, coordinated and collaborative efforts have been applied to the examination of materials degradation in RPVs, core internals, and structural concrete. This has also included comparative testing on small-scale sample test techniques.
International Nuclear Energy Research Initiative projects: The Korean Advanced Institute of Science and Technology (Republic of Korea) performs collaborative work with the LWRS-Materials Research Pathway on the testing and evaluation of austenitic stainless steel weld materials.
Membership in technical committees and organizations: Research on irradiated concrete and correlated reactor-aging issues are part of the International Committee on Irradiated Concreten Technical Committee 259-ISR Prognosis of deterioration and loss of serviceability in structures affected by alkali-silica reactions, within RILEMo, the International Union of Laboratories and Experts in Construction Materials, Systems and Structures. Involvement in the International Group on Radiation Damage Mechanisms (IGRDM) in Pressure Vessel Steels, and the International
- l. This document is an internal EPRI document and is not publicly available
- m. DOE-NE Light Water Reactor Sustainability Program and EPRI Long-Term Operations Program - Joint Research and Development Plan, INL-EXT-12-24562 Rev. 4, April 2015
- n. Information on the first general meeting of the International Committee on Irradiated Concrete, held November 2015, Knoxville, TN (http://web.ornl.gov/sci/psd/mst/ICICFGM/index.shtml)
- o. RILEM (http://www.rilem.org/gene/main.php) 34
Cooperative Group on Environmentally Assisted Cracking (ICG-EAC). This also includes LWRS support of researchers in technical code committees of the American Society for Testing and Materials.
2.5 Summary of Research and Development Products and Schedule The strategic goals of the Materials Research Pathway are to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear power plants and to provide industry with cost-effective mitigation, repair, or replacement options for the safe and economical extended operation of commercial nuclear plants. Near-term research is focused primarily on providing mechanistic understanding, predictive capabilities, and high-quality data to inform decisions and processes by both industry and regulators. Longer-term research is focused on alternative technologies to overcome or mitigate degradation. A chronological listing of the major milestones in the pathway can be found in Appendix A.
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- 3. PLANT MODERNIZATION 3.1 Background The U.S. operating nuclear fleet is an important national asset, providing approximately 20% of the nations electric supply as well as supporting grid stability, providing carbon-free energy, and ensuring generation-fuel diversity. However, the economic viability of the fleet is challenged by the abundance of low-cost shale, gas-fired generation, and heavily subsidized renewable generation. Several nuclear plants have, and others may, shut down permanently due to unprofitable operations, foregoing their remaining licensed operating years.
Nuclear power plants have significant opportunity to reduce their operating costs while improving operational performance through comprehensive plant modernization. Most sectors of the industrial economy renew and modernize their infrastructure on a regular basis, adjusting to new market conditions and applying new beneficial technologies, particularly those that are digital-based. The operating nuclear fleet, by contrast, operates largely based on a state of technology and a related model that is over 40 years old. The model is characterized by analog technology and a large operating staff performing manual activities for most plant functions.
Indeed, the focus of the industry over the life of the operating LWR fleet has been to incrementally improve the original infrastructure, rather than modernize it. Nuclear utilities have dealt with a number of non-discretionary capital investments to address safety and regulatory issues. This has resulted in deferral of much-needed reinvestment in the plants to address their aging systems and improve their operational efficiency. This reinvestment is now urgent.
It is therefore critical that proven solutions be available to nuclear utilities for wide-scale plant modernization that provides near-term cost reductions while producing a future state that is operationally and financially sound for decades to come. The Plant Modernization Pathway addresses many of the challenges that must be addressed in order to modernize operating nuclear power plants by conducting research to develop and deploy advanced digital technologies that enable widespread cost reduction and operational improvements.
Plant modernization begins with the Instrumentation and Control (I&C) systems. I&C systems are essential to ensuring safe and efficient operation of the LWR fleet. These technologies affect every aspect of nuclear power plants. They are varied and dispersed, encompassing systems from the main control room to primary systems and throughout the balance of the plant. They interact with every active component in the plant and serve as a kind of central nervous system.
Current instrumentation and human-machine interfaces in the nuclear-power sector employ analog technologies, such as those shown in Figure 18. In other power-generation sectors, analog technologies have largely been replaced with digital technologies. This is in part due to the manufacturing and product-support base transitioning to these newer technologies. It also accompanies a transition of education curricula for I&C engineers to digital technologies. Consequently, product manufacturers refer to analog I&C as having reached the end of its useful service life. Although considered obsolete by other industries, analog instrumentation and control continues to function reliably in the nuclear industry, though spare and replacement parts are becoming increasingly scarce. The same is true of the workforce that is familiar with and able to maintain analog I&C. In 1997, the National Research Council conducted a study concerning the challenges involved in modernizing existing analog instrumentation and controls with digital I&C systems in nuclear power plants. Their findings identified the need for new I&C technology integration. This need has only grown in the subsequent 20 years.
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Figure 18. Current instrumentation in a nuclear power plant control room is dominated by analog technology.
Replacing existing analog devices with digital technologies has largely not been undertaken within the nuclear-power industry worldwide. Those efforts that have been carried out are broadly perceived as involving significant technical and regulatory uncertainty. This translates into delays and substantially higher costs for these types of refurbishments. Such experiences have slowed the pace of analog I&C replacement and further contribute to a lack of experience with such initiatives. In the longer term, this may delay progress on the numerous I&C refurbishment activities needed to establish plants that are cost competitive in future energy markets. Such delays could lead to an additional dilemma; delays in reinvestment needed to replace existing I&C systems could create a bow wave for needed future reinvestments. Because the return period on such reinvestments becomes shorter, they become less viable the longer they are delayed. This adds to the risk that I&C may become a limiting or contributing factor that weighs against the decision to operate nuclear power plants for longer periods. I&C replacement represents potentially high-cost or high-risk activities if they are undertaken without the needed technical bases and experience to facilitate their design and implementation.
Most I&C upgrade projects today result in islands of digital control distributed throughout the plant.
They are physically and functionally isolated from one another in much the same way as their analog predecessors. Each may employ a different interface and use approach, thereby introducing potentially contradictory interaction styles that must be learned and recalled by licensed operators to be used correctly. Digital technologies are often implemented as point solutions to performance concerns, such as aging, in individual I&C components. This approach is characterized by planning horizons that are short and typically only allow for like-for-like replacements. It is reactive to incipient failures of analog devices and uses replacement digital devices to perform the same functions as analog devices. Consequently, many capabilities of the replacement digital devices are not used. This results in a fragmented approach to refurbishment that is driven by immediate needs. This approach to instrumentation, information, and control (II&C) aging management minimizes technical and regulatory uncertainty although, ironically, it reinforces the current technology base.
To displace the piecemeal approach to digital-technology deployment, a new vision for plant modernization that leverages the benefits of digital technologies is needed. This vision must extend beyond I&C systems to every aspect of nuclear plant operations and support. The current labor-intensive approach to plant work activities is no longer viable in todays energy market. Rather, these activities must be automated where possible, deploying a variety of existing and emerging technologies that can 37
more efficiently conduct these activities on a continuous basis. For example, much of the plant workload consists of periodically verifying the condition of plant components and systems. Online monitoring technologies can assume these functions on an ever-expanding basis as new sensors and analytical capabilities are developed. Where human skill and expertise are needed, work activities can be assisted by technology in ways that dramatically increase efficiency and accuracy in conducting them.
Altogether, these technologies will enable a new operating model that is significantly more cost competitive and sustainable than the current approach. This will enable the formulation of a new, more cost-effective basis for operating nuclear plants. The model includes consideration of goals for nuclear power plant staff numbers and types of specialized resources, targeting operation and management costs and the plant capacity factor to ensure commercial viability of proposed long-term operations, improved methods for achieving plant safety margins and reductions in unnecessary conservatisms, and leveraging expertise from across the nuclear enterprise.
3.2 Research and Development Purpose and Goals The Plant Modernization Pathway conducts targeted R&D to address aging and reliability concerns with the legacy I&C and related information systems, as well as the related operational processes, of the U.S. operating LWR fleet. This work involves two major goals:
- 1. To develop transformative digital technologies for nuclear plant modernization that renew the technology base for extended operating life beyond 60 years.
- 2. To enable implementation of these technologies in a manner that results in broad innovation and business improvement in the nuclear plant operating model, thereby lowering operating costs.
The objective of these efforts is to develop, demonstrate, and support the deployment of new digital I&C technologies for nuclear process control, enhanced worker performance, and enhanced monitoring capabilities to ensure the continued safe, reliable, and economic operation of the nations nuclear power plants. The technology R&D and collaborations through the Plant Modernization Pathway are intended to overcome the inertia that sustains the current status quo I&C systems technology and to motivate transformational change and a shift in strategyinformed by business objectivesto a long-term approach to plant modernization that will be more sustainable. Accordingly, DOE (through the LWRS Program Plant Modernization Pathway) is involved in this activity for the following reasons:
Instrumentation and control modernization is critical to the sustainability of the operating nuclear fleet. It is the only viable means of plant modernization that enables sufficient business improvement to ensure that the operating LWR fleet is viable in current and future energy markets.
Because of its short-term operational focus, the U.S. commercial nuclear industry could modernize its legacy I&C systems and still miss the opportunity to transform its operating model, thereby missing out on efficiencies available through advanced technologies that could reduce the costs of plant operations and outages.
A coordinated national research program is needed to develop transformative technologies and an implementation roadmap for meaningful plant modernization.
DOEs national laboratories maintain unique capabilities to develop and deliver a strategy for plant modernization that can be successfully deployed by the private sector:
- A federally funded program with industry cost sharing is technologically and organizationally neutral 38
- Utilities must own the solution to successfully producing a plant-specific licensing case for modernized instrumentation, control, and monitoring technologies
- National laboratories will collaborate with utilities to overcome barriers to technology deployment.
An effective R&D initiative must engage stakeholders (i.e., plant owners, regulators, vendors, and R&D organizations) to initiate relevant R&D activities. This requires the development and execution of a long-term strategy for nuclear power plant I&C-technology modernization based on the unique characteristics of the U.S. nuclear industry and its regulatory environment. In the near term, this strategy should lead to the ability to transition to a business model for nuclear power plant operation that employs a new technology base that will be less labor intensive, will facilitate greater digital-application deployment, and can be deployed seamlessly across the operational enterprise. The execution of this R&D approach will lay the foundation for a technology base that is more stable and sustainable over the long-term and assures the continued safety of power generation from nuclear energy systems.
The Pathway addresses the R&D needed to achieve a set of strategic plant modernization outcomes in the current LWR fleet by developing and demonstrating new technologies and operational concepts in actual nuclear power plants. This approach to associated R&D efforts focuses on two important outcomes:
Reducing the technical, financial, and regulatory risks of upgrading aging I&C systems to support extended plant life to and beyond 60 years of operation.
Providing the technological foundations for a transformed nuclear power plant operating model that improves plant performance and addresses the challenges of future business environments.
The research program is conducted in close cooperation with the nuclear utility industry to ensure that it is responsive to the challenges and opportunities in the present operating environment. The scope of the research program is to develop a seamless, integrated digital environment as the basis of the new operating model. The program is informed by a utility working group (UWG) comprised of leading nuclear utilities across the industry (representing approximately 70% of the existing LWR fleet) and EPRI. The UWG developed a consensus vision of how more integrated modernized plant I&C systems and modernized work practices and processes could address a number of challenges to the long-term sustainability of the LWR fleet.p Utility partners serve as participants in many R&D activities in which new technologies are eventually demonstrated and validated for production usage. This arrangement has a number of advantages:
It assures that a significant portion of LWR fleet decision-makers share the end-state vision for plant modernization It assures that near-term technologies are immediately beneficial while they comprise the long-term building blocks of a more comprehensively modernized plant It greatly reduces the risk of implementation for any one utility It allows the utilities to move forward together in transforming their operating models to fully exploit these technologies, providing a transparent process for coordinated assistance from the major industry support organizations of EPRI, the Institute of Nuclear Power Operations, and NEI.
- p. Long-Term Instrumentation, Information, and Control Systems (I&C) Modernization Future Vision and Strategy, INL/EXT-11-24154, Revision 3, November 2013.
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New value from II&C technologies is possible if they integrate with work processes, directly support plant staff, and create new efficiencies and ways of achieving safety enhancements. For example, data from digital II&C in plant systems can be provided directly to work-process applications and then, in turn, to plant workers carrying out their work using mobile technologies. This saves time, creates significant work efficiencies, and reduces errors.
Therefore, the new technologies of the Plant Modernization Pathway are founded on a seamless digital environment (see Figure 19) for plant operations and support by integrating information from plant systems with plant processes for workers through an array of interconnected technologies:
Plant systems: beyond centralized monitoring and awareness of plant conditions, deliver plant information to digitally based systems that support plant work and directly to workers performing these work activities in all of their work locations Plant processes: integrate plant information into digital field-work devices, automate many manually performed surveillance tasks, and manage risk through real-time centralized oversight and awareness of field work Plant workers: provide plant workers with immediate, accurate plant information that allows them to conduct work at plant locations using assistive devices that minimize radiation exposure, enhance procedural compliance and accurate work execution, and enable collaborative oversight and support even in remote locations.
Figure 19. The Plant Modernization Pathway is developing an architecture that encompasses all aspects of plant operations and support, integrating plant systems, and immersing plant workers in a seamless information architecture.
A strategy was developed to transform the nuclear power plant operating model by first defining a future state of plant operations, enabled by advanced technologies, and then developing and demonstrating the needed technologies to individually transform plant work activities. The collective work activities are grouped into the following major enabling capabilities:
- 1. Instrumentation and control architecture
- 2. Monitoring and plant automation
- 3. Advanced applications and process automation.
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In each of these areas, a series of pilot projects are being conducted that enable the development and deployment of new I&C technologies in existing nuclear power plants (see Figure 20). A pilot project is an individual R&D project that is part of a larger strategy needed to achieve modernization according to a plan. Note that pilot projects have value on their own, as well as collectively. A pilot project is small enough to be undertaken by a single utility, demonstrating a key technology or outcome required to achieve success in the higher strategy and supporting scaling that can be replicated and used by other plants. Through the LWRS Program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants.
The pilot projects conducted through this pathway serve as stepping stones to achieve longer-term nuclear plant modernization. They are designed to emphasize success in some crucial aspect of plant-technology refurbishment and sustainable modernization. They provide the opportunity to develop and demonstrate methods of technology development and deployment that can be broadly standardized and leveraged by the commercial nuclear-power fleet. Each of the R&D activities in this pathway achieves a part of the longer-term goals of safe and cost-effective sustainability. They are limited in scope, so they can be undertaken and implemented in a manner that minimizes technical and regulatory risk. In keeping with best industry practices, prudent change management dictates that new technologies are introduced slowly so that they can be validated within the nuclear safety culture model.
Figure 20. Pilot projects for the Plant Modernization Pathway.
The LWRS Program provides a structured research program and expertise in plant systems and processes, instrumentation and control, digital technologies, and human-factors science as it applies to nuclear power plant operations performance. The utilities provide cost-sharing in time, expenses, expertise in plant functions, plant documentation, and access to plant facilities, often including the plant training simulator. The products of the pilot projects are technology developments, demonstrations, and technical-basis reports that can be cited in regulatory filings, vendor specifications, utility feasibility studies, industry standards and guides, and lessons-learned reports.
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The transformation of the nuclear power plant operating model into the future vision will take more than a decade. This is necessary to fully assimilate pilot project technologies into plant operations and business processes. The rate of transformation is a function of how pilot projects are defined and sequenced, such that later combinations of these technologies create new capabilities that address the requirements of more-complex nuclear power plant work activities. The stages of transformation are depicted in Figure 21.
Figure 21. Stages of transformation in the Plant Modernization Pathway.
The first stage involves development of enabling capabilities that are needed to motivate the first movers in industry to adopt new digital technologies. The pilot projects serve to introduce new technologies to nuclear power plant work activities and validate them as meeting the special requirements of the nuclear operating environment. They must be verified not only to perform the intended functions with the required quality and productivity improvements, but also to fit seamlessly into the established cultural norms and practices that define the safety culture of the nuclear power industry. This stage is characterized as new digital technologies improving the quality and productivity of work functions as they are now defined.
The outcomes of the first stage are control-room upgrades employing new digital technologies that afford improved plant communication and coordination of critical activities, and on-line monitoring technologies to improve awareness of plant component performance and aging phenomena. The Human Systems Simulation Laboratory (HSSL) is a key development focus of this stage to enable studies and validations of main control room simulation as well as distributed command and control center (e.g., outage control center) simulation. This is also the stage where foundational technologies are developed for nuclear power plant field workers that are designed to improve efficiency, reduce error, and enable greater oversight and the integration of planning, execution, and operation of plant activities at power and during scheduled outages.
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The second stage begins when the enabling capabilities are combined and integrated to create new functionality. This is an aggregation stage: individual capabilities are combined to create new functions in the plant, such as visualization and communication for outage management or predictive capabilities with online monitoring to better estimate the remaining useful life of a major plant component, thereby improving spare-parts management and improving plant capacity. It includes the introduction of more enabling capabilities as further digital technology advancements are introduced and integrated. Pilot project technologies are formulated in anticipation of this integration stage, so they work together to support future integrated functions that will leverage capabilities across the nuclear enterprise. This stage is characterized as the reformulation of major organizational functions based on an array of integrated technologies.
The outcomes of the second stage are the early hybrid control room, automated work activities, advanced outage management, and centralized online monitoring facilities.
The third stage occurs when there is substantial transformation in how the nuclear power plant is operated and supported, based on the embedding of major plant functions in a seamless digital environment. This is enabled by adopting newly developed technologies to achieve process efficiencies and the continued creation of new capabilities through technology integration. This stage is characterized as a transformation of the nuclear power plant organization and plant operating model, based on advanced digital technologies that redefine and focus the roles of plant workers and support organizations on value-added tasks, rather than on organizational and informational interfaces.
The outcomes of this third stage are the highly integrated control room and integrated operations with their attendant functions and capabilities.
3.3 Plant Modernization Pilot Project Descriptions and Deliverables DOE and collaborating utilities will conduct R&D activities to achieve nuclear plant modernization.
The objective is to develop and demonstrate technologies needed to support the transition from analog to digital technologies, enable broad business innovation and cost performance improvement, test and demonstrate solutions to achieve viable end-state visions for plant modernization and provide the technical bases that support these transformations to ensure a sustainable operating life to 60 years and beyond.
For each of the areas, the current performance issues and needs are described, followed by a description of how technology developments can improve performance. Pilot projects that address needs in these areas are being carried out and are described in the following sections.
3.3.1 Instrumentation and Control Architecture Currently the LWR fleet has a mixture of traditional, analog I&C technology and newer, digital technology. Virtually all U.S. nuclear plants have undertaken some digital upgrading over the lifetime of the stations. In some cases, digital systems were the only practical replacement for legacy analog components. In other cases, digital systems were the preferred technology in that they could provide more precise control and greater reliability. The cumulative effect on the LWR fleet has been an ever-increasing presence of digital systems in LWR control rooms.
Developing and demonstrating an effective and efficient path forward for licensing and deployment of modernizing the LWR fleet through digital I&C has been elusive thus far. This has resulted in digital I&C upgrade projects at commercial nuclear power plants costing substantially more than expected, taking longer to perform, and producing a chilling effect on modernization and investments of this type.
Several challenging issues remain unresolved and require significant R&D for nuclear utilities to move forward with modernization. These key issues include defining the end-state digital architecture, developing a business case for implementation, addressing licensing-process burden, and developing implementation schedules compatible with short refueling outages.
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In addition to addressing the challenges associated with a modern I&C infrastructure, no large scale changes have been made to the layout or function of LWR control rooms. Utilities constantly strive to improve operator performance and, in particular, to address performance weaknesses identified as contributors to plant safety challenges. This usually is seen in enhancements to operator-performance protocols and expected behaviors.
Introducing digital systems into control rooms creates opportunities for improvements in control room function that are not possible with analog technology. These can be undertaken in measured ways, such that the proven features of the control room configuration and functions are preserved, while gaps in human performance that have been difficult to eliminate are addressed. By applying human-centered design principles in these enhancements, recognized human error traps can be eliminated and the introduction of new human error traps can be avoided.
Digital technology introduction provides an opportunity to enhance human performance in the control room. The process of designing and implementing digital control room technologies to replace analog systems serves as an opportunity to implement human-centered design activities throughout the various stages of design, acquisition, and implementation. These design activities and their technical bases (human-factors design standards and cognitive-science research) were not available at the time of the original design of main control rooms. Considerable progress has been made in these fields since the completion of the industrys response to the Three Mile Island 2 Action Plan, which requires a human factors approach to control room changes. Replacement digital technologies that have more-powerful and flexible graphical and informatics capabilities, together with a substantially improved understanding of how to leverage these capabilities to support effective human performance, afford the opportunity to realize a more human-centered main control room. This does not require a full-scope approach to control-room modernization, such as refurbishing or replacing an entire main control room as a single engineering project. Rather, it can be accomplished through gradual and step-wise related projects that are carried out when digital I&C systems are implemented to replace analog I&C systems addressing near-term reliability and operational needs. These types of enhancements can be performed anytime in the life cycle of the main control room and can add to the business case for implementing digital I&C.
Pilot projects have been defined to develop the needed technologies and methodologies to achieve performance improvement through incremental control-room enhancements as nuclear-plant I&C systems are replaced with digital upgrades. These pilot projects are targeted at realistic opportunities to improve control-room performance with the types of digital technologies most commonly being implemented, notably distributed control systems and plant computer upgrades.
This work employs the HSSL as a test bed, providing a realistic hybrid control-room simulation (refer to Section 3.2) for development and validation studies as part of pilot projects. In addition, the Plant Modernization Pathway has an agreement in place for access to control-room upgrade technologies developed by the Halden Reactor Project, which has played a key role in several European control-room upgrades. The Plant Modernization Pathways research program is well positioned to provide enabling science for U.S. hybrid and single step control room enhancements.
Major outcomes:
The industry will have a comprehensive nuclear-plant-modernization strategy that is adopted by a number of first-mover nuclear utilities.
A generalized end-state I&C architecture will be provided to the industry that will deliver significant operations and maintenance (O&M) cost savings, as well as serve as a focal point for I&C supplier-developed systems and components and generic regulatory approvals. It is recognized that individual plants have design-basis requirements that necessitate some customization of the end-state I&C architecture.
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Nuclear plant control rooms will be substantially modernized, with many nuclear plants moving to fully digital control rooms and others developing advanced hybrid control rooms.
Regulatory barriers to efficient I&C architectures and other new I&C beneficial features will be resolved through coordinated industry licensing and regulatory activities based on the consensus I&C end-state architecture.
A comprehensive business case for I&C and control-room modernization will be developed, comparing expected cost-savings for I&C end-state architecture with expected long-term costs to maintain legacy I&C systems now in use.
Detailed implementation plans for both single-step and hybrid approaches to plant modernization will be developed, reflecting the industrys best practices for project and risk management.
Major milestones associated with the pilot projects supporting I&C architecture are By 2019, develop strategy for full nuclear-plant modernization with a participating utility based upon digital technologies that demonstrates a sustainable plan and approach for efficient long-term plant operation and risk-informed asset management By 2019, develop a digital I&C architecture design and approach for managing the transition of legacy analog I&C to new advanced digital I&C that effectively addresses human-factors concerns, cost, and regulatory considerations.
In 2019, validate the human-factors engineering technical bases that support approaches to plant modernization of the main control room and plant systems.
In 2020-2021, demonstrate integration of advanced I&C infrastructure with a modernized control room using the Human Systems Simulation Laboratory.
In 2020, demonstrate newly developed methods that can meet all regulatory requirements as an accepted qualification method for I&C.
From 2020 to 2021, demonstrate integration of advanced alarm and task displays with data analytics using the DOE Human Systems Simulation Laboratory.
In 2020-2021, develop and deliver research results that support safe interaction between man, technology, and organizations to support the full nuclear-plant modernization and control-room modernization activities in the Human Systems Simulation Laboratory.
3.3.2 Online Monitoring and Plant Automation As nuclear power plant systems begin to be operated during periods longer than originally anticipated, the need arises for more and better types of monitoring of material and component performance. This includes the need to move from periodic manual assessments and surveillances of physical components and structures to centralized online condition monitoring. This is an important transformational step in the management of nuclear power plants. It enables real-time assessment and monitoring of physical systems and better management of active electromechanical components based on their performance. It also provides the ability to gather substantially more data through automated means and to analyze and trend performance using new methods in order to make more-informed decisions regarding maintenance strategies. Of particular importance will be the capability to determine the remaining useful life (RUL) of a component to justify its continued operation over an extended plant life.
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The current technology base for monitoring in the U.S. nuclear industry consists of signal processing techniques and advanced pattern-recognition (APR) programs that are technically mature and commercially supported. The application of advanced analytics is in the early stages of implementation by leading nuclear utilities. The implementation rate has been slow due to the required funding and infrastructure development for integrating monitoring programs within the operating and business environment.
APR provides highly sensitive anomaly detection of current condition or behavior for targeted components. Much of the value of online monitoring comes through early warning of imminent component failures. Commercial APR products rely on the continuous input of well correlated plant data to provide this early warning. While APR systems are effective at identifying equipment operating in conditions that may shorten the equipments RUL, they are limited to identifying operating data values that are abnormal in comparison to a historical baseline. Commercially available APR products cannot perform the next essential step of diagnosing the underlying cause for the abnormal data value. This diagnosis step relies entirely on a staff of highly trained specialists to troubleshoot and diagnose the underlying problem and to recommend a corrective action response. Furthermore, the RUL of a monitored asset cannot be determined by APR technology, and there are long-term failure modes that are not detectable with APR technology. Current APR products are not suitable for long-term monitoring and management of nuclear assets and, in particular, for passive assets evaluated on an intermittent basis using NDE measurement techniques.
The development and advancement in diagnostic and prognostic capabilities is required to achieve an automated ability to directly identify equipment condition from initial warning signatures. This will support analysis of long-term component behavior, related risk, and RUL. It will further provide verification of asset condition as evidence of design qualification and economic viability. The work of this project is being coordinated with two other LWRS Program Pathways. The Plant Modernization Pathway is working with the Risk Informed Systems Analysis Pathway to appropriately incorporate advanced risk-modeling analytical methods and tools. Also, the Plant Modernization Pathway is developing new technologies to automate monitoring, as well as evaluating methods to do that, while the Materials Research Pathway is evaluating materials to better understand what would be considered unacceptable erosion or corrosion, requiring replacement.
Advanced, digital monitoring technologies will enable early detection of degraded conditions that can be addressed before they significantly contribute to preventable consequences or damage. The early detection of degradation is one of the more significant factors in extending a components lifetime. A timely response to the causes of degradation also can significantly improve nuclear safety and prevent damage to other nearby components and structures. Finally, these new capabilities will reduce the cost of manual diagnostic work.
A gap exists between the state of technology needed and the effective application of diagnostics and prognostics to nuclear plant assets. To address this gap, research tasks have been developed to demonstrate a plant operational support model that the LWR nuclear industry can implement, which will substantially reduce labor requirements to an affordable and sustainable level.
Major outcomes:
Research will identify new sensors and analytical capabilities to support advancements to existing technology. These new instruments will support an optimized centralized online monitoring approach enabling higher levels of automation of operations and maintenance activities.
An efficient, secure monitoring end-state architecture will be developed that provides real-time knowledge of component degradation and plant thermal performance, as well as enables innovative support arrangements for fleet centralization and third-party remote services.
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Research will develop methods to support transition to inline technologies that are integrated into all plant work and risk-management processes so that any required decisions and actions are seamlessly and automatically enabled.
Regulatory barriers to greater use of monitoring and automation will be addressed through coordinated industry efforts to demonstrate and validate their technical and regulatory acceptability through appropriate licensing and regulatory-guidance efforts.
Specific business cases will be developed for monitoring and plant-automation technology implementation, both incrementally for new technologies and comprehensively in determining the cost savings for the entire nuclear plant O&M costs.
The major milestones associated with online monitoring and plant automation are By 2019, complete a technology roadmap for comprehensive nuclear power plant monitoring using advanced I&C technology In 2019, develop a data-driven method for monitoring plant-operations performance that integrates process data with anomaly detection both to reduce cost of operations and to improve plant performance By 2019, develop a risk-informed model to optimize equipment-maintenance frequency and diagnostic models of equipment performance to enable transition to a condition-based maintenance strategy at nuclear power plants In 2019, develop a common information model requirements developed for nuclear power plant databases to support greater work process and plant automation In 2020, pilot a multimodal online piping-monitoring system.
In 2020, collaborate with industry to demonstrate a common information model example During 2020 and 2021, demonstrate asset-risk models that enable predictive maintenance implementation strategy and calibration monitoring During 2020 and 2021, demonstrate technology, automation, and data-driven methods for monitoring plant operations, inspections, and surveillances From 2020 through 2021, pilot multimodal online-diagnostic reinforced-concrete monitoring system During 2020 and 2021, research automated data integration methods based on a common information model From 2021 through 2022, develop and demonstrate automated data-integration methods.
3.3.3 Advanced Applications and Process Automation Although the LWR fleet has made significant improvements in human performance over the past decade, it continues to be impacted by human error, resulting in plant transients and other significant outcomes. While consequential error rates are relatively low (typically measured in the range of 104 consequential errors on a base of 10,000 hours0 days <br />0 hours <br />0 weeks <br />0 months <br /> worked), the sheer number of work hours accumulated by plant staff over time means that errors impacting plant safety and reliability still occur.
The traditional approach to improving plant worker human performance was to focus on correcting worker behaviors. This approach produced substantial improvement since the time this emphasis began in the mid 1990s. Up to that time, there were periodic plant trips and transients due to human error (such as working on the wrong component or even the wrong operating unit). These types of errors have gradually been reduced until they are relatively rare. However, other types of errors continue to cause or complicate nuclear safety challenges. Between 2008 and 2010, a series of incidents occurred at various nuclear power 47
plants, many of which were considered to be among the industrys best performers. These incidents were documented in INPO SOER 10-2, Engaged, Thinking Organizations,q which assigned a significant portion of the causes to human error and lack of operator fundamental knowledge.
The focus on correct worker behaviors typically involves an analysis of worker actions and implementation of corrective actions in the form of additional training, procedure upgrades, job and memory aids (i.e., acronyms and neck-strap cards), additional peer checking, management job observations, and similar techniques. While some improvement is usually obtained from these corrective actions, a cumulative impact arises by adding complexity to work activities that make work tasks slow and cumbersome. Some industry observers believe that a saturation point has been reached for these techniques and that the industry has reached the practical limits of human reliability at the present error rates.
To further improve human performance for NPP field workers, a fundamental shift in approach is needed. Digital technology can transform tedious error-prone manual tasks in nuclear power plant field activities into technology-based structured activities and functions with inherent error prevention, detection, and correction features. This has the potential to eliminate or reduce human variability in performing routine actions, such as identifying the correct components to be worked on. In short, technology can perform tasks at much higher reliability rates while maintaining desired worker roles of task direction, decision making, and work-quality oversight.
Nuclear power plants are perhaps the only remaining safety-critical operations that rely to a large degree on human skill to conduct routine and emergency activities. Adoption of digital technologies has enabled other high-consequence industries (e.g., aviation, medical procedures, and high-precision manufacturing) to transfer tedious control functions to automation while retaining qualified operators in a supervisory role.
This situation is largely due to technology limitations during the 1970s and early 1980s when the currently operating nuclear power plants were designed. While main processes pertaining to reactor operations are automated (e.g., core power level with automatic rod control), the vast majority of plant controls for configuration changes or placing equipment in and out of service are manual. This overreliance on manual control on this large a scale challenges operators and results in current human-error rates.
A highly automated plant is one in which the most frequent and high-risk control activities are performed automatically under the direction of an operator. Because of higher reliability in well designed automatic control systems, improvements will be realized in nuclear safety, operator efficiency, and production. The chief impediment to the widespread implementation of this concept is the cost of retrofitting new sensors, actuators, and automatic-control technology to existing manual controls. The goal of this research will be to demonstrate that the resulting improvement in safety and operating efficiencies will offset the cost of making these upgrades.
Major outcomes:
An industry-consensus end-state vision for maximum nuclear plant worker and process efficiency will be in place to substantially reduce O&M costs, minimize errors and work-quality issues, and improve overall plant performance.
An end-state architecture will be developed to seamlessly integrate work activities and data-transfer requirements using wireless networks to enable mobile-worker applications and real-time collaborations among work groups, however remote.
- q. Institute of Nuclear Power Operations (INPO) Significant Operating Experience Event Report (SOER) 10 2, Engaged, Thinking Organizations, survey available at https://www.surveymonkey.com/r/2KY7LRY 48
Mobile plant workers will have multiple task-related applications and information resources available to them at the work site, connecting them seamlessly to related work activities, supervisory and work oversight functions, work-management centers, and remote technical support.
Many process requirements that are conducted in the course of a field work activity, such as status reporting, human-error prevention measures, obtaining real-time authorizations, verifying materials and test equipment, etc., will be performed automatically and transparently to the degree possible, maximizing job efficiency, reducing the number of workers per activity to the minimum possible and eliminating many categories of work process errors (such as correct component identification).
Regulatory issues, such as cybersecurity concerns, regarding technology-enabled worker and process innovations, will be adequately addressed and resolved through coordinated industry engagement activities with the regulator.
Appropriate business cases will be developed for monitoring and plant-automation technology implementation, both incrementally for new technologies and comprehensively in determining the cost savings for the entire nuclear plant O&M costs.
A logical sequence of implementation for online monitoring and plant automation technologies will be developed, considering the underlying architecture and the optimum implementation order to gain the highest leveraged benefits through individual technologies. It is recognized that individual plants will have different priorities depending on O&M needs and existing monitoring capabilities.
The major milestones associated with the pilot projects supporting integrated operations are to:
By 2019, identify opportunities and benefits for use of image-processing technologies in operations and maintenance activities and demonstrate an example image-processing technology in a laboratory environment.
By 2019, integrate drone-compatible technology to support operations and maintenance activities at a nuclear power plant site.
During 2020 and 2021, research, develop, and demonstrate methods of data mining in the work-management systems of nuclear power plants.
By 2022, develop concepts for advanced control automation for control-room operators based on human technology-function allocation developed in the pilot project for automating manually performed plant activities.
By 2023, develop a concept for integrating I&C architecture into a seamless digital environment that enables widespread automation and process efficiencies across the plant-support functions. This integration strategy will be consistent with all design and regulatory bases.
By 2024, develop an end-state vision and implementation strategy for an advanced computerized operator support system, based on an operator-advisory system that provides real-time situational awareness, prediction of the future plant state based on current conditions and trends, and recommended operator interventions to achieve nuclear safety goals.
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3.4 Research Facilities and Capabilities The Human Systems Simulation Laboratory (HSSL), located at INL, is used to conduct research in the design and evaluation of hybrid control rooms, the integration of control-room systems, the development and piloting of human-centered design activities with operating crews, and visualization of different end-state operational concepts. This advanced facility consists of a reconfigurable simulator that supports human-factors research, including human-in-the-loop performance and human-system interfaces.
It can incorporate mixtures of analog and digital hybrid displays and controls. It is applicable to the development and evaluation of control systems and displays of nuclear power plant control rooms, as well as other command and control systems.
The HSSL consists of a full-scope, full-scale reconfigurable control-room simulator that provides a high-fidelity representation of an LWR analog-based control room (see Figure 22) with 15 linked bays that feature touch panels that respond to touch gestures in ways similar to the control devices in an actual control room. The simulator is able to run actual LWR-plant simulation software used for operator training and other purposes. It is reconfigurable in the sense that the simulator can be easily switched to the software and control board images of different LWR plants, thus making it a universal test bed for the LWR fleet.
Figure 22. HSSL: a reconfigurable hybrid control room simulator.
For this research program, the HSSL is mostly used to study human performance in a realistic operational context with new control-room technologies and designs. New digital systems and operator interfaces are developed as software prototypes and depicted in the context of the current-state control room, enabling studies of the effects of proposed upgrade systems on operator performance (see Figure 23). Prior to full-scale deployment of technologies (such as control-room upgrades), it is essential to test and evaluate the performance of the system and the human operators use of the system in a realistic setting. In control-room research simulators, upgraded systems can be integrated into a realistic representation of the actual system and validated against defined performance criteria.
The key advantage of mimicking current control rooms is an ability to implement prototypes of new digital function displays into the existing analog-control environment.
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Figure 23. HSSL provides critical support as researchers hold operator workshops with nuclear utilities to evaluate new control-room display designs.
The HSSL uses advanced performance-measurement technologies to objectively assess whether the pilot-project technologies achieve the intended performance goals. This includes eye-tracking technology, which enables researchers to determine where an operators attention is focused while performing their control-room tasks in the HSSL. These tools are being continually improved through research activities.
For example, Halden Reactor Project is developing new tools to enhance the modeling, collection, and analyses of eye-tracking data. These new tools will be validated through testing and incorporated into the measurement technologies for human-performance engineering. Combining new control-room technologies with these advanced performance-measurement tools, the HSSL is used to validate new operational concepts, human-centered design methods, and many first-of-a-kind technologies for the LWR fleet, thereby ensuring safe and efficient nuclear power plant modernization of I&C systems based on demonstrated and validated scientific principles. Validation through operator-in-the-loop studies helps ensure the safety, reliability, and usability of the novel technologies, as well as in meeting regulatory standards for the adoption of new operational systems.
3.5 Cybersecurity Cyber security is recognized as a major concern in implementing advanced digital instrumentation information and control technologies in nuclear power plants in view of the considerable security requirements necessary to protect these facilities from potential adversaries, as well as protecting company-proprietary information. Industry stakeholders of this research have expressed the need to ensure that cybersecurity vulnerabilities are not introduced through the adoption of these advanced digital technologies. Furthermore, these utilities have internal cybersecurity policies and regulatory obligations that must be upheld during implementation of the project technologies.
To this end, a project task has been completed to address cybersecurity issues arising from the technology developments in the pilot projects. A cybersecurity plan assessment has been conducted to identify possible threat vectors introduced by the new technologies. Individual assessments will 51
periodically be conducted for pilot projects to identify threats specific to new technologies, characterize the degree of cybersecurity risk, and recommend effective mitigation measures. These assessments will be discussed with the host utility for the pilot projects, and the information will be provided to the UWG.
Responsibility for cybersecurity ultimately lies with the utilities that implement technologies from this research program. They must ensure their own policies and regulatory commitments are adequately addressed.
It is recognized that these technologies represent a proof-of-concept state; therefore, these technologies are not as prescriptive in terms of underlying technologies as might normally be required in an actual cybersecurity evaluation for a nuclear plant. For example, a technology might refer to the use of wireless transmission of information to mobile field workers without specifying the type of wireless protocols. Therefore, in future utility evaluations of actual implementations of the pilot-project technologies, assessment outcomes might be different according to the implementation options.
The research pathway will continue to apply cybersecurity resources, expertise, and the experience of DOE, as well as the nuclear industry, to provide a sound information basis for utilities in prudent technology-implementation practices and mitigation measures.
3.6 Research and Development Collaborations The Plant Modernization Pathway of the LWRS Program implement a vigorous engagement strategy with the U.S. nuclear-power industry, including operating companies, major support organizations, the NRC and suppliers. This engagement strategy supports industrial and regulatory engagement. The goal of this engagement strategy is to develop a shared vision and common understanding across the nuclear industry of the need for nuclear power plant modernization, the performance improvement that can be attained, and the opportunities for collaboration to enact this vision.
To ensure that the research activities of the Plant Modernization Pathway are relevant and have the maximum impact for sustainability to the LWR fleet, the industry stakeholders are specifically engaged on identifying the:
Key issues and challenges in operating aging systems and managing obsolescence Priorities for mitigating aging and obsolescence through a common set of priorities suitable for federal research Issues that impact ability to implement technical solutions, needed research to enable needed regulation, and other areas for joint research.
Key collaborations for the plant Modernization Pathway include:
Utility Working Group: The Plant Modernization Pathway engages nuclear utilities through the UWG, the set of nuclear utilities that the Pathway has worked with over the life of the LWRS Program. At present, the UWG consists of 14 leading U.S. nuclear utilities. Additional membership for the UWG is constantly pursued with the intent to involve every U.S. nuclear operating fleet in the program. The UWG is directly involved in defining the objectives and research projects of this pathway. Pilot-project collaborators make the results of R&D available and accessible to other commercial nuclear utilities and participate in efforts to support deployment of systems, technologies, and lessons learned by other nuclear power plant owners.
Electric Power Research Institute: EPRI is both a member of the UWG and serves in a direct role in collaborative research with the Plant Modernization Pathway. EPRI technical experts directly participate in the formulation of the project technical plans and in the review of the pilot-project results, bringing to bear the accumulated knowledge from their own research projects and collaborations with nuclear utilities. EPRI will assist in the transfer of technology to nuclear utilities by publishing formal guideline documents for each of the major areas of development.
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Halden Reactor Project: The programs at the Halden Reactor Project extend to many aspects of nuclear power plant operations; however, the area of interest in this R&D program is the human-machine interface technology research program in the areas of computerized surveillance systems, human factors, and man-machine interaction in support of control-room modernization. Halden has assisted a number of European nuclear power plants in implementing I&C modernization projects, including control room upgrades. The Plant Modernization Pathway will work closely with the Halden Reactor Project to evaluate their advanced I&C technologies to take advantage of the applicable developments. In addition to the technologies, the validation and human-factors studies conducted during the development of the technologies will be carefully evaluated to ensure similar considerations are incorporated into the pilot projects. Bilateral agreements may be employed in areas of research where collaborative efforts with the Halden Reactor Project will accelerate development of the technologies associated with the pilot projects.
Major Industry Support Organizations: The LWR fleet is actively supported by major industry support groupsnamely, the NEI and the Institute of Nuclear Power Operations (INPO). These organizations have active efforts in the I&C area, including technical developments, regulatory issues, and standards of excellence in conducting related activities. It is important that these organizations be informed of the purpose and scope of this research program, and that activities be coordinated to the degree possible. It is a task of this research program to engage these organizations to enable a shared vision of the future operating model based on an integrated digital environment and to cooperate in complementary activities to achieve this vision across the industry with maximum efficiency and effectiveness. There are additional industry support groups (such as the PWR and BWR owners groups) that need similar engagement for more focused purposes.
Nuclear Regulatory Commission: Periodic informational meetings are held between DOE and members of the NRC to communicate the aims and activities of individual LWRS Program pathways.
Briefings and informal meetings will continue to be held to inform staff from NRCs Office of Nuclear Regulatory Research about the technical scope and objectives of the LWRS Program.
Suppliers: Ultimately, it will be the role of nuclear industry I&C suppliers to provide commercial products based on technologies developed under this research program. Engagement activities with nuclear industry II&C suppliers are being conducted to facilitate communication and to make the technologies that are produced through research (such as the reports of research, insights, and available lessons learned) to suppliers so that advancements made through this program benefit the LWR fleet through available commercial products based on best practices.
Bilateral International Collaborations: Bilateral international collaborations provide important information to the Plant Modernization Pathway. Collaboration with Korea is ongoing and discussions are underway with India to collaborate on online monitoring and probabilistic risk assessment of failures in digital instrumentation and control.
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- 4. RISK-INFORMED SYSTEMS ANALYSIS 4.1 Background The U.S. nuclear industry faces challenges to remain economically competitive in electricity markets for a number of reasons. Safety is key to all aspects of LWR operation, including the cost of operations.
In the context of extending the operating periods of nuclear power plants, traditional deterministic safety concepts may not guarantee the most economic results. One of the approaches to achieve greater cost efficiencies is to identify and optimize plant safety margins. In the LWRS Program, the RISA Pathway focuses on optimizing safety margins and minimizing uncertainties to achieve high levels of safety and economic efficiencies. The RISA Pathway will provide enhanced capabilities for analyzing and characterizing LWR systems performance by developing and demonstrating methods, tools, and data to enable risk-informed margins management (RIMM).
The RISA Pathway aims to improve economics and reliability and sustain safety of operating nuclear power plants over periods of extended operation. The goals of the RISA Pathway are twofold: 1) deploy the RISA toolkit of technologies that enables better representation of safety margins and the factors that contribute to cost and safety, and 2) conduct advanced risk-assessment applications with industry to support margin management strategies that enable more cost-effective plant operation. The methods and tools provided by the RISA Pathway support effective margin management for both active and passive SSCs. Figure 24 shows the strategy of the RISA Pathway.
The RISA toolkit will be applied in industry-application pilot projects. These pilot projects were developed through discussions with U.S. nuclear utilities. The projects affect the following focus areas that correspond to key industry challenges: (1) enhanced resilient nuclear power plant concepts, (2) cost and risk categorization applications, and (3) margin recovery and operating cost reduction. The pilot projects represent studies using selected applications of the RISA toolkit. The research will also address needed verification and validation (V&V) of the tools and methods that are used in pilot projects. The RISA Pathway will continue communicating with stakeholders to identify emerging issues and challenges faced by the operating fleet and to identify opportunities to conduct applied research with risk-informed methods and tools that improve both margin management and plant economics.
High Value RISA Pilot Project
+ Safety Impact
+ Economic Impact Figure 24. Safety and economic benefit from the RISA Pathway.
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4.2 Research and Development Purpose and Goals The RISA Pathway provides enhanced capabilities for analyzing and characterizing LWR-systems performance by developing and demonstrating methods, tools, and data to enable RIMM.
The purpose of the RISA Pathway R&D is to support plant owner-operator decisions with an aim to improve the economics, reliability, and safety sustainability of current nuclear power plants over periods of extended plant operations. The goals of the RISA Pathway are twofold:
- 1. Deploy the RISA toolkit of technologies that enable better representation of safety margins and the factors that contribute to cost and safety
- 2. Conduct advanced risk-assessment applications with industry to support margin management strategies that enable more cost-effective plant operation.
A strategy to accomplish the above RISA Pathway goals includes the following:
- 1. Conduct research to develop and demonstrate industry applications in pilot projects that employ the RISA methodology collaboratively with organizations from the U.S. commercial nuclear power industry
- 2. Leverage industry pilot demonstration projects to address needs of the entire industry, demonstrating how the use of risk-informed techniques can improve plant efficiency and increase confidence in their use.
Figure 25 shows the RISA Pathway program structure. The main R&D program focuses on methods, tools, and data areas. The RISA R&D results will be applied to challenging industry-focused research:
industry application pilot projects.
Figure 25. The RISA Pathway programmatic structure.
To better understand and characterize safety margins in risk-informed engineering, two types of analyses used in this pathway are integrated: probabilistic and mechanistic analyses, as shown in Figure 26. Note that in actual applications, a combined approach is essential where both types of analysis are used to support any one particular decision.
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Figure 26. Types of analysis that are used in the RISA Pathway.
4.3 Research, Development, and Demonstration Activities and Milestones The RISA Pathway research is performed within the framework of specific focus areas that represent key challenges identified by nuclear power plant owner-operators. The focus areas represent groups of applications of risk-informed technology to assist operating nuclear power plants reduce costs and otherwise adapt to the changing economic and generating-mix environment. The focus area demonstrations are in the areas of (1) enhanced resilient nuclear power plant concepts, (2) cost and risk categorization applications, and (3) margin recovery and operating cost reduction. Scalable pilot projects for RISA methodologies and technologies are planned in each of these focus areas, coordinated through an interest group of other utilities and industry stakeholders. Upon successful demonstration, the technology may then be scaled up to support applications by a larger community of users.
In 2018, the LWRS Program organized a meeting with participation from U.S. nuclear utilities to discuss and develop the pilot projects. Projects were identified based on a discussion of critical issues arising in the U.S. nuclear industry. Table 1 shows how each pilot project relates to a RISA R&D focus area. These are high priority industry issues, topics that can significantly impact plant operations in the near future, making them valuable and relevant applications for the RISA toolkit. The RISA Pathway will continue to communicate with various U.S. nuclear stakeholders to obtain feedback on current research, identify new issues, and develop a long-term plan for research and development that is responsive to the challenges of sustaining the existing LWR fleet.
While simulation methods in risk and reliability applications have been previously proposed, the availability of advanced mechanistic and probabilistic simulation tools were limited; however, with advanced tools and modern computational resources, simulation is now a viable approach to represent complex scenarios. Consequently, the RISA Pathway approach will use a set of existing and advanced simulation tools to model plant behavior, determine safety margins, and evaluate cost-saving strategies to plant performance. The simulation software used within the RISA Pathway includes both mature codes and advanced simulation codes being developed by NEAMS, CASL and other DOE programs.
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Table 1. Pilot projects related to RISA R&D focus areas.
R&D focus areas Pilot Projects Enhanced resilient nuclear RISA Enhanced Resilient Plant Systems power plant concepts Enhanced Operation Strategies for System Components.
Cost and risk categorization Risk-Informed Asset Management applications Plant Health Management.
Dynamic Fire Probabilistic Risk Assessment (PRA)
Modernization of Design Basis Accidents Analysis with Application on Margin recovery and Fuel Burnup Extension operation cost reduction Digital I&C Risk Assessment Plant Reload Process Optimization 4.3.1 Enhanced Resilient Nuclear Power Plant Concepts Enhanced resilient nuclear power plant concepts consists of pilot projects that aim to enhance both the safety and economics of existing nuclear power plants through the use of advanced, near-term technologies that provide substantial improvements to plant safety margins. The value of enhanced, resilient plant concepts is in providing greater safety margins to operating plants; this, in turn, allows plants greater flexibility in managing operations within their current safety margins. This may result, for example, in greater time to cope with design basis accidents that can be used as a basis for requalifying plant SSCs with significant cost savings. Today, the industry is developing accident-tolerant fuels (ATFs) and implementing FLEX and an industry-wide initiative entitled, Delivering the Nuclear Promise:
Advancing Safety, Reliability, and Economic Performance. The enhanced resilient plant systems concept may incorporate combinations of ATF, optimal use of FLEX, enhancements to plant components and systems, and the incorporation of augmented or new passive-cooling systems, as well as improved fuel cycle efficiency to establish improvements in plant safety margins that can be used to requalify or reclassify plant SSCs or otherwise obtain greater flexibility in plant operation with attendant cost reductions.
The key metrics that will be used to evaluate the resiliency enhancements for a nuclear power plant include:
Increased coping time as compared to the current state of fuel and plant systems Decreased Core Damage Frequency and Large Early Release Frequency, as compared to the current state of both fuel and plant systems Increased safety margins, such as more margins on fuel/clad temperature or reduced hydrogen gas generation, as compared to the current state of fuel and plant systems Improved plant economics during normal operations.
The objective of this research effort is to use the RISA methods and toolkit in industry applications, including methods development and early demonstration of technologies, in order to enhance existing reactors safety features (both active and passive) and to substantially reduce operating costs through risk-informed approaches to plant design modifications to the plant and their characterization. High-value evaluations of proposed ATFs, together with enhanced resilient-plant-system concepts, will be performed to identify both the technical and the economic elements associated with industry adoption of the technologies. Two industry application pilot demonstrations are proposed in this focus area.
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4.3.1.1 RISA Enhanced Resilient Plant Systems This research includes a pilot project that will focus on the combined effects and evaluation of different plant-resilience-enhancing technologies to obtain synergy and benefits from near-term technologies and existing resources. For instance, the research will support an evaluation of strategies to improve both safety and economics in the current environment of high operating costs. The pilot project will evaluate safety and economic benefits from the deployment of strategic investments, either stand-alone or in combination with ATF and FLEX concepts. The analyses will use both deterministic (e.g.,
Reactor Excursion and Leak Analysis Program (RELAP)5-3D, TRACE, MELCOR, MAAP) and probabilistic (e.g., SAPHIRE, HUNTER) tools, as well as the LOTUS and Risk Analysis and Virtual Control Environment (RAVEN) controller software. The pilot project will include FLEX equipment in analyses to optimize resiliency during beyond design-basis accidents (BDBAs). The major milestones associated with this task are In 2019, demonstrate incremental improvements in safety margins available to a candidate plant by integrating individual design and operational enhancements, such as near-term ATF concepts, FLEX equipment, and passive cooling systems, through risk analysis for an enhanced resilient plant model By 2020, complete plant-level scenario-based risk analysis of an enhanced resilient-plant model based on an existing BWR By 2021, complete human-reliability analysis to credit FLEX in accident management and perform risk-informed analysis of a passive-cooling design.
The successful completion of these milestones will enable an industry-wide adoption of technologies that enhance plant resilience and allow nuclear power plants to better cope with both internal and external events and operate with high efficiency and fewer disruptions.
4.3.1.2 Enhanced Operation Strategies for System Components The Terry turbine is widely used in nuclear power plants around the world, including Fukushima Daiichi Units 2 and 3. Extended knowledge of the maximum operating limit of this type of equipment will enhance current assessments credited for emergency core cooling, thus enabling a clear identification of turbine operation margin. This knowledge may be used as part of a technical basis to reduce current assessments of plant risk during accidents by providing additional coping time and time to transition to other core cooling equipment, such as FLEX equipment, in order to prevent core damage. The research will be conducted using advanced modeling methods and full-scale experimental testing to develop an overall Terry turbine expanded operating band (TTEXOB). This research will address a specific area for BDBA Terry-turbine performance evaluations, such as pump-function modeling, oil and bearing characteristics, and turbine operating behavior under two-phase flow conditions. The major milestones associated with this task are In 2019, complete and document reactor-core-isolation cooling full-scale component and Terry-turbine nozzle experiments, modeling, and initial validation By 2020, complete test planning for integral full-scale experiments for long-term low-pressure operations By 2021, complete integral full-scale experiments for long-term low-pressure operations By 2021, complete test planning for integrated, scaled experiments replicating Fukushima Daiichi Unit 2 self-regulating feedback (if necessary).
The outcome of this research will provide a technical basis of extended operation of TTEXOB during accident conditions, as well as the turbine maintenance plan.
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4.3.2 Cost and Risk Categorization Applications The objective of this focus area is to conduct research with risk-informed approaches to develop and test methods to decrease operational costs of nuclear power plants. Two cost-sensitive areas have been identified as initial targets: component reclassification-repurpose (see 10 CFR 50.69) and component-testing maintenance.
The first area of interest is component recategorization based on 10 CFR 50.69. In current deterministic regulations, the SSCs are categorized as safety-related or non-safety-related. Safety-related SSCs need special treatment. The safety-related SSCs, under deterministic method, will cause increasing cost of SSC design, licensing, and operation. By using a probabilistic risk-informed method under 10 CFR 50.69, both safety and non-safety related SSCs could be recategorized, following risk-informed safety categorization (RISC):
RISC-1: Safety-related SSCs that perform high safety significance RISC-2: Non-safety-related SSCs that perform high safety significance RISC-3: Safety-related SSCs that perform low safety significance RISC-4: Non-safety-related SSCs that perform low safety significance.
Under the guidance of 10 CFR 50.69 risk-informed categorization, SSCs in the safety-related category in Figure 27 could be re-categorized into the high (RISC-1) or low (RISC-3) safety-significance categories. The SSCs in category RISC-3 could next avoid special treatment, which can enhance plant economics. By using RISA tools and methods, the technical basis of the SSC categorization will be enhanced and could be linked to observable engineering margin metrics.
High safety RISC 1 RISC 2 Probabilistic significance RiskInformed Low safety significance RISC 3 RISC 4 Safety related Nonsafety related Deterministic Figure 27. RISC methodology (courtesy of 10 CFR 50.69).
The second area of interest is to optimize component testing and maintenance costs, while maintaining plant safety and performance. A large portion of the cost in U.S. nuclear power plants comes from maintenance and testing, which is driven by regulatory and reliability requirements to ensure safe and continuous operation. Cost reduction could be achieved by optimizing plant safety, incorporating plant dynamics, physical aging, and degradation processes into the safety analysis in a single consistent analysis framework.
Given these two areas of interest, the objective of this focus area is to develop an innovative framework on risk categorization to enhance economics. Figure 28 provides a schematic diagram of this framework structure. The approach combines physics, risk, and cost information to enable a risk-and-cost-based decision-making process for optimizing maintenance activities and achieving the greatest cost efficiency. Two pilot demonstrations are identified in this focus area.
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Classical
- Plant reliability data
- Plant simulation models Dynamic Cost
- Plant cost data
- Costbased measures
- Testing & maintenance Operation Data
- Time dependent procedures and data Data Analysis costrisk analysis Figure 28. Integrated framework for optimizing maintenance activities.
4.3.2.1 Risk Informed Asset Management The goal of this research is to enhance the long-term safety and economics of nuclear power plants approaching, preparing for, or during a subsequent license renewal (SLR) period of operation by providing a structured risk-informed approach to evaluate and prioritize plant capital investments made in preparation for and during this period of extended plant operation. The risk-informed analysis for capital improvements during SLR includes assessment of the expected useful life of SSCs and the likelihood and impact of unanticipated occurrences on performance during this operating period. The scope of research addresses the optimal conditions and timing for equipment or SSC replacement or refurbishment, based on anticipated conditions with appropriate assessments performed to address uncertainties. This research will produce several near-term key milestones:
In 2019, complete an application of RISA risk-analysis tools for capital SSC refurbishment or replacement for an industry collaborators SLR case to optimize allocation of plant capital By 2020, extend the use of RISAs risk analysis optimization toolkit for SSC maintenance and testing of an existing plant to reduce costs related to plant operation By 2021, expand the risk analysis toolkit from an individual plant to the existing fleet for candidate SSC refurbishment or replacement.
The outcome of this research will provide a set of risk-informed methods and data to optimize plant economic resources and reduce plant operational costs.
4.3.2.2 Plant Health Management Industry equipment reliability (ER) programs are essential to support safe and economic plant operation and are well addressed in several industry-wide and regulatory programs. However, these programs are both labor intensive and expensive. Leveraging advanced monitoring technology can significantly reduce costs, enhance engineering effectiveness, and improve the performance of vital SSCs.
The main goal of this research is to develop risk-informed tools that can be used in monitoring and evaluating data arising from structural health monitoring and deploying structural health-monitoring technologies as an element of plant equipment-reliability programs. Three milestones are planned in this research.
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In 2019, integrate a risk-informed framework into proposed system-health programs to reduce costs and improve engineering effectiveness of the systems through coordination with advanced monitoring technology development By 2020, integrate equipment failure data and models with existing plant system-health programs to optimize testing, maintenance, and replacement of components By 2021, integrate real-time equipment monitoring data to system health.
The outcome of this research will provide a set of analytic tools to better assess plant health condition and use data to optimize SSC testing and replacement costs.
4.3.3 Margin Recovery and Operation Cost Reduction Existing U.S. nuclear power plants are designed and constructed based on defense-in-depth safety principles for critical systems needed to fulfill safety functions. Design-basis safety analyses are performed using deterministic approaches which normally employ conservative models and assumptions to provide tolerances to account for uncertainties. However, the accumulation and aggregation of uncertainties arising from these conservatisms in current deterministic design approaches may result in overly conservative operating requirements that limit the operating flexibility of the current fleet and can result in inflated plant-operating costs.
Research in margin recovery is focused on developing high-fidelity, risk-informed, multiscale and multiphysics tools and methods to conduct a comprehensive investigation of plant safety requirements and their method of implementation to assess approaches to recover margins associated with conservatisms of legacy licensing, design, and analysis. A goal or outcome from this research is to identify and develop the technical bases that may permit existing nuclear power plants to operate more efficiently and with more operational flexibility less cost due to unnecessarily restrictive requirements.
The general objective of this focus area is to develop an integrated evaluation approach that combines plant PRA methods with either a multiphysics best estimate plus uncertainty (MP-BEPU) or a risk-informed multiphysics best estimate plus uncertainty (RI-MP-BEPU) approach. The RI-MP-BEPU framework will employ modern high-fidelity probabilistic and best-estimate modeling and simulation tools with consistent uncertainty propagation, quantification, and sensitivity analysis. This approach will integrate various simulation tools across a full spectrum of plant analysis activities, including core design, fuels performance, component aging and degradation, systems analysis, containment response, radionuclide transport, and release and risk assessment. This will enable complex multi-physics and risk-informed approaches to be implemented so that important nuclear power plant system problems can be solved with efficiency and speed. This approach will be used to estimate and quantify the safety margins that are available in licensing-case scenarios so that decisionmakers, both plant owners and regulators, can consider approaches to the allocation of available safety margins. This could provide the potential for nuclear power plants to reallocate available margins to other activities or applications and provide commensurate operational cost savings.
Four pilot projects are planned in this focus area.
4.3.3.1 Dynamic Fire PRA This research will develop methods and tools to help improve efficiency in creating and employing fire PRA models and reducing conservatism in analyses. The research will focus on two main areas:
- 1) development of data mining methods and a graphical interface tool to couple existing nuclear power plant fire accident PRA analysis data and spatial information to identify and reduce unrealistic fire safety margins in daily plant operation, and 2) combining tools with dynamic capabilities to identify key relationships and timing to improve current fire PRA models. The research will target reduction in the costs for current fire PRA activities. This will be done with the modification and development of a visualization tool to enable users to manage spatial relationships of components in fire zones and their 61
failure properties, execute fire simulation (using the Consolidated Fire and Smoke Transport or Fire Dynamics Simulator software) to determine component and cable failures and easily analyze subsequent component failures or basic events due to cable failures. The PRA modeling capabilities will be coupled with fire-simulation and visualization tools. This will also include evaluating the margins gained by using dynamic fire PRA results used to modify an initial static fire PRA model. The research will deliver the following milestones.
In 2019, demonstrate improvements in plant analysis and reduction of manual efforts with routine fire analysis using improved models for an existing plant By 2020, apply physics-based fire simulations with fire PRA data to demonstrate methods to reduce risk using high-consequence fire scenarios for an existing plant By 2021, develop the strategy to extend the implementation of fire PRA tools for the existing fleet.
4.3.3.2 Modernization of Design-basis Accident Analysis with Application of Fuel Burnup Extension Increasing fuel enrichment and discharge burnup promises significant cost reduction through the nuclear fuel cycle for any type of operating plant around the world. This research will use 6wt% enriched fuel with a 24-month fuel cycle for a four-loop Westinghouse-designed PWR. The RI-MP-BEPU framework will be used to conduct a licensing and deployment study. The VERA-CS code will be used to study pin-resolved power distributions for the core design, followed by detailed fuel-performance calculations for individual fuel rods in the core using the FRAPCON and BISON fuel performance-analysis codes. The analyses will apply both deterministic (e.g., VERA-CS, FRAPCON/FRAPTRAN, BISON, RELAP5-3D, MELCOR) and probabilistic (e.g., SAPHIRE, RAVEN) methods using the LOTUS controller. Uncertainty-quantification and sensitivity-analysis technologies will be integrated into the RI-MP-BEPU framework to address issues of extrapolating experimental data. The major deliverables are In 2019, complete fuel-rod non-burst-potential evaluation under loss-of-coolant accident (LOCA) conditions for an existing plant with extended burnup exceeding the current burnup limit By 2020, complete comprehensive uncertainty quantification and sensitivity analysis to inform the experiments that will be conducted at the TREAT reactor for high-burnup fuel to study fuel fragmentation and relocation issues (this is in collaboration with an NSUF project)
By 2021, demonstrate the feasibility of extended fuel-burnup designs for all prescribed plant transient scenarios and cost-benefits analyses of optimal combinations of enrichment, burnup, and cycle length for an existing PWR.
4.3.3.3 Digital Instrumentation and Controls Risk Assessment Deployment of digital I&C systems for safety-related applications will enhance the long-term reliability of existing and availability of replacement technologies, improve familiarity with a future workforce, and reduce uncertainties associated with aging-system performance. The main objective of this research is to develop and apply methods to assist in the reliability and risk assessment of safety-related digital I&C technologies to support licensing and qualification. The work scope includes technology-gap identification to develop PRA methods for digital I&C qualification methods; conceptual design of a proposed I&C system, modeled at the channel logic level with a detailed design and functional specification, followed by a risk and reliability evaluation to support the analysis of an integrated reactor protection system (RPS); and engineered safety feature actuation system (ESFAS) replacement technology, using a risk-informed, graded approach to safety significance. Several key milestones are expected from this research:
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In 2019, conduct research to develop a risk-assessment strategy for digital I&C upgrades using current digital technology information By 2020, in collaboration with the LWRS Program Plant Modernization Pathway and a participating plant, apply risk-informed tools to address common-cause failure modes that may be induced by a candidate digital I&C technology.
By 2021, conduct plant-level scenario-based simulations of unanalyzed transients concurrent with common-cause failures to demonstrate plant safety with digital I&C upgrades.
4.3.3.4 Plant Reload Process Optimization Optimization of plant reloading reactor core thermal limit is one of the main requests from U.S.
nuclear utilities to reduce operating costs. The optimization of safety margins could be proposed by developing independent methods for analysis of design basis accidentsincluding LOCA and non-LOCA eventsthat will be compliant with the new 10 CFR 50.46c regulations and thermal-conductivity degradation (TCD) evaluations. The research will also include assessments of peak cladding temperature (PCT) during LOCA analysis and departure of nucleate boiling (DNB) analysis associated with non-LOCA events. The research scope includes an analysis of core design, safety margins, fuel performance, and modern data management. This research will focus on developing an approach to customizing the thermal limits of nuclear power plants, based on the core physics at certain reload cycles, by applying risk-informed methods to optimizing safety margin. Three major deliverables are planned:
In 2019, develop the technical basis for a risk-informed framework to optimize the existing reload licensing process for an operating PWR By 2020, perform risk-informed reload licensing calculations to demonstrate the margin and costs benefits for the pilot plant By 2021, extend the reload licensing risk-informed framework to the existing fleet.
4.3.4 RISA Toolkit Deployment Plan The RISA toolkit is a set of computer software that is used in RISA industry-application pilot-demonstration projects in coordination with industry users and stakeholders to support their use in RIMM analyses. Figure 29 shows various computer codes that are included in the RISA toolkit. The RISA Pathway will perform an appropriate degree of software verification and validation (V&V) in order to provide the needed level of quality to support users in the U.S. nuclear industry. It is noted that not all tools are developed under RISA Pathway.
Deployment of the RISA toolkit consists of the following four steps:
Select tools and methods.
Confirm verification and validation status.
Pilot demonstrations using selected tools and methods.
Deploy tools and receive industry feedback.
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RiskInformed Systems Analysis Tools Core Design VERA-CS Fuel Performance FRAPCON/FRAPTRAN BISON Domain Knowledge (plant System Analysis RELAP5-3D RELAP-7 characteristics, operational data, building Containment MELCOR structures, site layout, core Severe Accident MELCOR design, procedures, failure Material Aging GRIZZLY information, etc.)
to Create Plant- Natural Hazards MASTODON NEUTRINO CFAST FDS Specific Models PRA SAPHIRE RAVEN EMRALD HUNTER Code Integration LOTUS RAVEN Market Economics CRAFT RAVEN Figure 29. Current software modules used to perform RISA-specific analyses.
Figure 30 shows an example of a notional timeline of the RISA Pathway five-year pilot-demonstration project. In FY 2019, the RISA Pathway will focus on the initiation of selected pilot demonstrations and deliver preliminary results on each selected pilot project according to industry needs.
Feedback on the selected pilot-project research will be communicated with industry through a RISA Pathway industry working group. Based on initial demonstration studies, each pilot project may be extended to full-scale analysis during FY 2020 and 2021. The V&V of the associated RISA Toolkit will be done during this period. For the last two years of this research, the RISA Pathway will develop optimized methods to implement R&D results to industry, as well as a long-term plan for support by research institutes to the U.S. nuclear industry that will provide sustainable benefit from RIMM. It is noted that the research-project timeline could vary according to the technical maturity and development status of RISA tools and methods.
4.3.4.1 Selection of a RISA toolkit The tools and methods used in the RISA Pathway should have high confidence and sufficient technical maturity for an implementation to industry at its current setting. Advanced technologies multiphysics and multiscale analysis, cutting-edge computational proficiency, and capability for uncertainty controlshould be applied to the RISA toolkits. The toolkit should also have a capability to support risk-informed decision making for both probabilistic and deterministic elements of safety. The RISA toolkit includes various computer-simulation tools that can cover a wide range of work scope.
Many of the tools are currently employed by industry and well-validated, with mature technology levels.
However, there are also tools under development that need to be verified and/or validated for industry use.
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- Preliminary study on selected pilot demonstrations FY2019 *Create RISA-industry working group
- Perform full scale pilot demonstrations FY2020 *RISA Toolkit validation and verification
- Continue full scale pilot demonstrations and validations FY2021 *Initiation of RISA Pathway industry deployment
- Finalize full scale pilot demonstration and validations FY2022
- RISA Toolkit deployment FY2023 *RISA Pathway technology transfer to industry Figure 30. Notional 5-year plan for a pilot demonstration of industry application of the RISA Pathway.
4.3.5 Verification and Validation (V&V) of the RISA Toolkit In order to provide confidence during industry deployment, the selected tools and methods should address quality-assurance levels appropriate for industry use. Well-known methods, such as V&V and uncertainty quantification of the produced results, will both enhance credibility of the selected tools and enable industry to use them with confidence. The selected RISA toolkit will be examined to confirm V&V status to show its technology readiness level (TRL) and to assure quality of the outcome. The RISA Pathway will provide V&V examination and confirmation data for the selected RISA toolkit and application, as appropriate. The V&V data will include specific information for the tool, such as capability and features, quality-assurance program, developer/independent V&V records, separation/integral tests history, user documents, and feedback.
4.3.6 Pilot demonstration using the RISA toolkit Currently, a total of nine RISA pilot projects are proposed during discussions with U.S. nuclear utilities. Each project has its own selected tools and methods. The RISA Pathway will maintain strong engagement with the U.S. nuclear industry to perform each pilot project. Industry will join the pilot demonstration project to identify emerging issues and facilitate innovative solutions and smooth deployment of the RISA toolkit industry-wide.
4.3.7 RISA Toolkit Industry Deployment and Feedback Deployment of the RISA toolkit to the U.S. nuclear industry to help reduce operating costs through improved margins management is a goal of the RISA Pathway. The industry has been engaged with the RISA Pathway from its initiation and supports pilot projects by using a selected RISA toolkit. During the pilot projects, industry will have sufficient time to gain experience with the selected tools and methods.
The RISA Pathway will organize periodic meetings with industry stakeholders to obtain feedback on ongoing research, to identify priorities and challenges to the U.S. LWR industry, and to discuss an effective RISA-toolkit implementation strategy. Licensing and regulation issues will be also addressed.
The result of these meetings are published and shared for long-term direction of the RISA R&D Pathway.
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4.4 Research and Development Partnerships The RISA Pathway relies on a strong partnership with industry to ensure that the RISA Toolkit is useful and is targeting relevant and important challenges. Coordination with the LWRS Program Pathways, other DOE programs, government agencies, and universities are also important, and international activities are pursued as warranted.
EPRI: As an R&D organization for the nuclear industry, EPRI may be a valuable contributor to the technical evaluation and cost-benefit assessment associated with the industry pilot demonstrations.
Through its access to its members, EPRI can also sponsor relevant analyses at operational nuclear power plants to assess and demonstrate the benefits of advanced technology deployment.
NEAMS: The RISA Pathway will leverage models developed (and under development) by NEAMS.
Development of RELAP-7 and RAVEN has been a joint effort between the NEAMS and LWRS Programs.
CASL: CASL is developing a detailed model of the LWR core. If investigations in the LWRS Program warrant it, the LWRS Program-developed models can couple with the CASL-developed models. CASL has an interest in using RELAP-7 for one or more of their challenge problems.
NRC: NRC program outreach includes updates, information exchanges, and collaborative efforts with the staff where identified.
Owners Groups: Interactions will continue with groups such as the BWR and PWR owners groups through information exchange and evaluations of specific topics via case studies.
Bilateral International Collaborations: Bilateral international collaborations provide important information to the RISA Pathway. Discussions are underway with India, Korea, and Japan to partner and develop RISA methodologies.
Multilateral International Collaboration: A variety of international researcher interactions are of potential interest to the RISA Pathway, including the Organisation for Economic Co-operation and Development Nuclear Energy Agency and the European Commission activities.
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Appendix A LWRS Program Accomplishments 67
Appendix A LWRS Program Accomplishments Appendix A includes a summary of the LWRS Programs previous years major accomplishments.
More detail on accomplishments is provided for recent years, with higher-level summaries of the preceding years. Reports on these topics can be found on the LWRS website at https://lwrs.inl.gov.
Fiscal Year 2018 Materials Research (formerly known as Materials Aging and Degradation)
Validated the RIME code for evaluating radiation-induced swelling in stainless steel.
Completely evaluated and validated of mini compact-tension specimen design for use in fracture-toughness determinations of RPV steel for irradiated materials.
Released first version of the Grizzly code for engineering-scale assessment of RPV structural integrity.
Delivered experimentally validated, physically based thermodynamic and kinetic model of precipitate-phase stability and formation in 316L stainless steel under anticipated extended lifetime operation of LWRs.
Procured high-fluence (up to 125 dpa) materials for incorporation into IASCC testing.
Completed fundamental experimental examinations on the mechanisms of water-chemistry (LiOH vs.
KOH) influence on corrosion of stainless steel.
Completed study on the influence of radiation-induced void swelling on crack growth rate under PWR, primary water conditions.
Developed the foundation for the MOSAIC tool to evaluate concrete-mix sensitivity to irradiation damage.
Completed experimental fatigue testing of stainless steel under PWR primary conditions and development of a basic model for environmentally assisted fatigue in a surge line pipe.
Produced preliminary methodology evaluation and technique development for nondestructive examination of concrete sections.
Down-selected candidate alloys for the Advanced Radiation Resistant Materials project following extensive prescreening tests involving ion and proton irradiation. Testing involved IASCC crack-susceptibility screening using proton irradiated samples, as well mechanical testing and corrosion testing of fabricated alloys.
Assessed the efficiency of HWC on irradiation-assisted stress corrosion crack growth rate for high-fluence BWR materials Machined harvested Zion RPV test specimens Demonstrated laser welding and friction stir welding techniques for initial welding of irradiated materials, establishing the capabilities of the hot-cell welding cubicle.
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Plant Modernization (formerly known as Advanced Instrumentation, Information, and Control Systems Technologies)
Conducted research in the Human Systems Simulation Laboratory, documenting the design philosophy and technical basis for near term advanced hybrid control room end-state concepts.
Developed a strategy for implementation of safety-related digital instrumentation and control systems.
Developed an end-state vision and human-factors engineering program plan to support development of a hybrid control-room design in an existing nuclear power plant control room.
Completed the development and evaluated the conceptual design for a liquid radiological waste system resulting in an advanced hybrid control room.
Developed uncertainty-quantification methods to support development and deployment of structural health monitoring of concrete.
Initiated the development of a strategy to carry out full nuclear plant modernization using digital technology as a replacement of existing analog technologies.
Risk-informed Systems Analysis (formerly known as Risk-informed Safety Margin Characterization)
Completed development of a risk-informed approach for a flooding tool based upon smoothed particle hydrodynamics.
Developed a science and technology roadmap for enhanced accident-tolerant nuclear power plant systems, metrics, scenarios, risk analysis, and modeling.
Conducted RISA industry-use case analysis to identify and prioritize applications of risk-informed technology to assist operating nuclear plants reduce costs.
Conducted a plant-level scenario-based risk analysis for a proposed enhanced resilient plant systems design.
Developed a strategy for employing risk-informed methods to enable margin recovery of NPP operating margins to reduce operating costs and improve operational efficiencies.
Reactor Safety Technologies Continued Terry-turbine testing and model development, focused on expanding the operating band of the reactor core isolation cooling system for BWRs and the steam generator auxiliary feedwater system for PWRs.
Completed melt spreading (MELTSPREAD) and core debris coolabilty (CORQUENCH) models for industry use in the development of boiling water reactor severe accident water management strategies including modeling, validation, and users manuals.
Improved a proof-of-concept computational tool to support a BWR technical support center during an emergency.
Continued development of models for assessing BWR containment passive cooling capabilities.
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Fiscal Year 2017 Materials Research Developed a mechanistic model for transition temperature shifts in RPV steels.
Performed evaluation of hydrogenated water chemistry and normal water chemistry on IASCC crack growth rates of high-fluence 304L stainless steel, through a collaboration with NFD in Japan.
Completed the Radiation Induced Microstructural Evolution (RIME) code to predict radiation-induced swelling model (RIME code) for austenitic steel.
Identified the stages of stress corrosion crack initiation and growth in alloy 600 and 690.
Completed installation of a welding cubicle to test advanced welding techniques on irradiated materials.
Developed a process monitoring technique for friction stir welding of irradiated materials.
Completed work on optimizing parameters for auxiliary beam stress improved laser welding conditions.
Demonstrated initial solid-state welding capabilities using the weld cubicle on irradiated materials.
Completed initial work on the influence of post-irradiation annealing as an IASCC mitigation technique.
Completed validation of mini-Compact Tension test development on non-irradiated RPV material and initiate round-robin testing of irradiated specimens.
Developed experimental techniques for in situ mechanical testing of irradiated materials under electron microscopy and neutron scattering characterization to evaluate conditions leading to strain localization associated with crack initiation.
Completed a model for environmentally assisted fatigue for RPV.
Completed 10,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> accelerated aging and analysis of cast austenitic stainless steel materials.
Completed a sample machining of the harvested baffle former bolts from a commercial PWR plant.
Continued machining of RPV samples form the Zion Unit-1 RPV.
Completed the down-selection of advanced candidate alloys for further neutron irradiation testing, toward the ultimate goal of providing industry with superior materials options beyond 316L and X-750 alloys.
Completed the analysis of key degradation modes of cable insulation.
Risk-Informed Systems Analysis Completed a report documenting status of the additional model data sets and results to accelerate the verification and validation of RELAP-7.
Completed a report documenting the progress made in smooth particle hydrodynamics-based wind representation.
Completed a report on results for modeling representative dynamic risk scenarios for a multi-unit power plant site using RAVEN.
Completed a report for the Integrated Cladding/Emergency Core Cooling System Performance:
Demonstration of LOTUS-Baseline coupled analysis of the South Texas Plant model.
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Completed a report documenting the coupled margins analysis for enhanced external hazard analysis including seismic and flooding phenomena.
Completed a report documenting experimental results related to wall penetrations and door failures for flooding fragility experiments.
Completed a report documenting the Beta 1.0 version of seismic probabilistic risk assessment model.
Completed a report to update lower length scale modeling of embrittlement in reactor pressure vessel steels.
Completed a report of a full-scope margins analysis of a commercial reactor to analyze an industry-important issue such as application to 10 CFR 50.69.
Completed a report describing the Reactor Metals beta version 1.5 of Grizzly.
Plant Modernization Completed a report documenting a user study to evaluate automated work package capabilities.
Interrogated ASR degraded concrete samples using acoustic and thermal techniques to support the development of a structural health monitoring framework.
Conducted research to quantify potential cost savings for a nuclear power plant through full modernization of the control room and instrumentation and control systems.
Completed a report that documented the industry and regulatory engagement activities that were conducted by the Plant Modernization Pathway during FY 2017.
Published INL/EXT-16-39808 report, Design Guidance for Computer-Based Procedures for Field Workers.
Completed a report summarizing the Potential to Extend the Range of Established Online Monitoring Technologies, Such as Guided Waves in Nuclear Power Plant Systems.
Created three-dimensional models of commercial nuclear power plant control rooms to support analyses for planned control room upgrades.
More effectively integrated regulatory considerations when performing human factors engineering for control room modernization.
Established a framework for a business case study on control room modernization in a fleet-based context.
Conducted research in the HSSL that evaluated advanced end-state concepts for commercial nuclear power plants.
Developed information-rich displays for a radiological waste control room at a commercial nuclear power plant.
Conducted research to evaluate technologies for detecting interactions between current plant configuration states and component manipulations directed by in-use procedures to support outage risk management improvement.
Reactor Safety Technologies Updated Fukushima Daiichi forensics inspection plan with prioritized activities, timeline, and expected costs.
Initiated Terry turbine testing focused on expanding the operating band of the reactor core isolation cooling system.
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Completed ex-vessel coolability and water management analysis and experiments and delivered initial versions of the analysis tools MELTSPREAD and CORQUENCH to EPRI for evaluation.
Provided updated insights from severe accident analysis modeling for severe accident management guidelines.
Provided improved version of a proof-of-concept computational tool to support a BWR technical support center during an emergency.
Fiscal Year 2016 Materials Research Comparative analysis performed of flux effects on high-fluence RPV alloys between experimental reactor irradiated (e.g., high flux) and commercial reactor surveillance samples (e.g., LWR, lower flux).
Progress reported in the development of computational models for predicting the hardenability of RPV alloys under irradiation and the stability of precipitate phases.
Demonstrated correlation of mini-Compact Tension test specimen data to established standard fracture toughness sample geometries for non-irradiated RPV steels.
Developed separate computational models for solute segregation in austenitic stainless steel under thermal and radiation-induced aging.
Completed 10,000 hour0 days <br />0 hours <br />0 weeks <br />0 months <br /> accelerated thermal aging milestone of CASS alloys.
Completed a report on the role of grain orientation in materials to applied load in the SCC initiation and crack growth in irradiated stainless steel.
Completed a report on the identified stages of crack nucleation and growth in Ni-base alloy 600.
Identified the dependence of grain boundary microstructure on the crack initiation in Ni-base alloy 690 and the variables that influence stress corrosion crack resistance.
Performed round robin testing of alloy 600 to determine laboratory and heat-to-heat variability.
Expanded test capabilities to evaluate SCC initiation in materials.
Completed the harvesting of Ginna baffle former bolts.
Developed component level cyclic plasticity model for thermal fatigue of 508 low alloy pressure vessel steel under load following conditions.
Completed the harvesting of RPV panels from the Zion nuclear power plant and started the process of test sample fabrication.
Completed the design, construction, and pouring of the large-scale alkali-silica test blocks and beginning of non-destructive monitoring of changes in the blocks under accelerated aging conditions.
Delivered unified parameter to assess irradiation-induced damage in concrete structures.
Developed numerical mesoscale radiation induced volumetric expansion (RIVE) model for concrete.
Completed coupling of the concrete RIVE model with temperature, creep, and concrete restraint towards the application of a two-dimensional-representation of a concrete barrier shield through reduced order modeling.
Reported on the implementation of ASR macroscale model in Grizzly.
Completed the synergistic effects of thermal and radiation damage in cross-linked polyethylene.
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Completed thermal aging and analysis of harvested 30-year service life ethylene propylene rubber and chlorosulfonated polyethylene cables, with remaining useful life estimated to be compatible with second license renewal conditions.
Evaluated frequency domain reflectometry as a potential system for cable condition monitoring.
Installed the welding cubicle for irradiated materials at Oak Ridge National Laboratory (ORNL).
Developed the Integrated Computational Welding Engineering (ICWE) tool to proactively manage stresses during laser repair welding of highly irradiated materials.
Risk-Informed Systems Analysis Completed a technical report describing the system reliability analysis capability and surrogate model applications in RAVEN, where the system response is evaluated by sampling the input space using various built-in sampling schemes.
Completed a report documenting the Beta 1.0 release of RAVEN and associated enhancements, including the implementation of ensemble modeling for time-series modeling, the implementation of model validation for surrogate models, and an advanced visualization capability for topology-based data analysis.
Completed a report documenting the 1.0 release of Grizzly, including an engineering fracture capability for reactor pressure vessels, an engineering model for embrittlement, and a modular architecture for modeling aging mechanisms.
Completed a report documenting the Beta 1.5 release of RELAP-7, including improved closure relationships and steam/water properties.
Completed a report on the RISMC analysis, including the effects of higher burnup on cladding performance as part of the LOCA/ECCS evaluation of risk-informed margins management strategies for a representative pressurized water reactor as part of ongoing collaboration with the South Texas Project.
Completed a report documenting RELAP-7 verification and validation activities for the current version of the software.
Completed a report on multi-hazard evaluating including both seismic and flooding for advanced PRA scenarios, where seismically-induced flooding models are coupled with a thermal-hydraulics code in order to better understand plant behavior.
Completed a report on the application of data analysis approaches for the RISMC Toolkit where we implemented and applied several methods and algorithms to analyze large amounts of time-dependent data that are produced during scenario simulation.
Completed a report describing the insights of the flooding fragility experiments for an initial set of mechanical components, including fragility prediction and uncertainty modeling related to water inundation.
Plant Modernization Conducted research to enable development of a concrete health monitoring framework for online monitoring of concrete for degradation due to ASR.
Conducted research that led to the development of an overview display to allow advanced outage control center management to quickly evaluate outage status.
Conducted research to develop a recommended end-state concept for the Palo Verde Nuclear Generating Station control room modernization design project.
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Completed a report documenting online monitoring of induction motors.
Produced a business case study on outage management research that documents the quantitative and qualitative performance improvement potential.
Completed a report summarizing digital features required to integrate work order, procedures, mobile communication, and smart devices to achieve higher worker efficiency.
Completed research that included detailed design guidance for computer-based procedures based on result from all research activities conducted in the project.
Conducted research that led to a proposed structural health framework for online monitoring of aging and degradation of secondary piping systems.
Completed a report that documents the business case for the Control Room Modernization project.
Reactor Safety Technologies Updated the Fukushima Daiichi forensics inspection plan with prioritized activities, timeline, and expected costs.
Performed a Fukushima Daiichi accident uncertainty analysis that provided: (1) information for planning decommissioning activities; and (2) areas of interest from a data sampling standpoint to be used in improving and validating severe accident codes.
Developed initial scope, cost estimates, and experimental plan for expanding the operating band of the reactor core isolation cooling system.
Completed a seismic margins evaluation, which was observed both physically and numerically.
Completed ex-vessel coolability and water management analysis and experiments.
Provided insights from severe accident analysis modeling for severe accident management guidelines.
Provided proof-of-concept computational tool to support a BWR technical support center during an emergency.
Fiscal Year 2015 Materials Research Completed a fracture toughness test for the round robin test program using mini-disc compact specimens.
Conducted a comparative analysis of results from High Flux Isotope Reactor and National Institute of Standards small-angle neutron scattering experiments on RPV steels.
Completed the disassembly and initiation of post-irradiation examinations of the University of California Santa Barbara Advanced Test Reactor-2 experiment.
Established the capability to determine surface strain in four-point bend tests conducted in a BWR normal water chemistry environment.
Reported on the results of post-irradiation examination and localized deformation studies on key specimens to evaluated mechanisms of IASCC.
Documented current results and progress on determining SCC initiation responses for alloy 600 and alloy 690 materials.
Completed the analysis of detailed predictions of swelling under LWR irradiation conditions.
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Completed the development of cluster dynamics framework for modeling precipitate formation and assessing the impact of flux and fluence effects on micro structural evolution in low alloy steel and austenitic stainless steel.
Completed the application of cluster dynamics model to establish radiation enhanced diffusion parameters for understanding phase development and segregation in irradiated metallic alloys.
Completed the development of a component level cyclic plasticity model for environmental fatigue of Type 316 stainless steel.
Completed thermodynamic predictive modeling of the effect of thermal aging on the phase development in cast austenitic stainless steels.
Documented the test plan for analysis of mechanical properties and microstructural analysis of the Ginna baffle bolts.
Documented model development and experimental results on analysis, monitoring, and establishment of large scale testing of ASR-affected concrete structures.
Completed ultrasonic linear array testing of thick concrete test slab and advanced signal processing.
Completed a simplified model and statistical analysis of structural significance of irradiation on the biological shield.
Completed post irradiation evaluation of the effects of fluence and temperature on swelling of select mineral analogues of aggregates.
Completed the assessment of cable aging equipment and the establishment of an experimental test matrix.
Completed the documentation of assessing key indicators of aging cable insulation and current state of the art NDE techniques for cable aging.
Risk-Informed Systems Analysis Released beta version 1.0 of RELAP-7.
Gathered data from three major seismic events (North Anna - August 2011, Fukushima Daichii and Daini - March 2011, and Kaswazaki-Kariwa - 2007) for the validation of seismic models.
Completed the Users Manual for RAVEN (a probabilistic based scenario simulation code).
Completed deterministic reactor pressure vessel fracture mechanics capability in Grizzly (a component aging model).
Conducted a preliminary analysis on the emergency core cooling system cladding acceptance rule.
Evaluated several flooding simulation tools and selected three for future use.
Implemented the extended finite element method technique (for studying reactor pressure vessel flaws) in Grizzly.
Demonstrated human reliability simulation for flooding scenarios.
Completed the preliminary comparison of nonlinear soil-structure interaction analysis with traditional (linear) seismic probabilistic risk assessment.
Plant Modernization Developed a probabilistic health monitoring framework and demonstrated this application to aging concrete structures.
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Completed the requirements for nuclear power plant control room computer-based procedures.
Conducted a field evaluation for the added functionality and new design concepts of the prototype computer-based procedure system.
Implemented software tools in the HSSL that enable fully functional hybrid control room systems.
Developed a process for simulator studies in support of control room upgrades.
Improved the graphical displays for an Advanced Outage Control Center, employing human factors principles for effective real-time collaboration and collective situational awareness.
Conducted evaluations/demonstrations of the automated work package prototype system and plant surveillance and communication framework requirements at host utilities.
Developed a cyber security program evaluation exercise for the pilot project technologies.
Conducted a gap analysis of the current state of digital architecture at nuclear power plants as compared to what is needed to support future digital technology environment.
Demonstrated the importance of verification and validation of systems used by operators across the design lifecycle rather than just in the late stages of the design process.
Reactor Safety Technologies Conducted the uncertainty analysis on the Fukushima Daiichi unit (1F1) accident progression with the MELCOR code.
Conducted a preliminary model of reactor core isolation cooling steam-turbine-driven pump with the MELCOR code.
Conducted a technology gap analysis on accident tolerant components and severe accident analysis methodologies.
Completed a forensics inspection plan with prioritized activities, timeline, and expected costs.
Completed an analysis of environmental conditions experienced from a core melt accident for key sensor parameters in a PWR and a BWR.
Gathered lessons learned from seismic events documenting the gathered data, identified margins, and recommended R&D that can provide more realistic seismic analysis and seismic PRA approaches.
Fiscal Year 2014 Materials Research Completed a post-irradiation examination plan for the assessment of ATR-2 capsules from ORNL and the University of California-Santa Barbara.
Completed the comprehensive and comparative analysis of atom probe tomography and small-angle neutron scattering experiments on available high-fluence RPV steel specimens.
Completed an assessment of embrittlement effects in an RPV nozzle.
Completed the examination of the microstructural and mechanical properties to determine possible root cause of failures in alloy 718 material.
Completed the development of refined microstructural model for radiation-induced swelling in high-fluence core internals.
Measured SCC initiation response in alloy 690, including effects of cold work, surface damage, and dynamic strain.
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Completed phase transformation studies in solute addition alloys.
Completed crack growth rate studies of solute addition alloys and comprehensive analysis of crack growth rate as a function of solute addition and commercial microstructure-controlled alloys.
Developed a test matrix and irradiation test plan to address the potential loss of efficiency of hydrogen water chemistry to mitigate IASCC in a BWR at high-fluence.
Developed an initial correlation between localized deformation and IASCC response using bend tests.
Completed mechanical testing and microstructural analysis for pristine cast stainless steel materials.
Completed the integration plan for joint cable research with EPRI and other stakeholders.
Completed the assessment of experimental work for determining key indicators in aged cables for correlation to NDE techniques.
Completed the design of a large-scale concrete mockup to study the effects of ASR on shear fracture propagation in stress-confined safety-related structures.
Completed the assessment of radiation induced aggregate swelling as a degradation mode in irradiated concrete structures.
Completed the preliminary conceptual design of a thick concrete NDE specimen.
Developed a unified parameter for characterization of ionizing radiation intended for the evaluation of radiation-induced degradation of concrete.
Completed initial investigation of improved volumetric imaging of concrete using an advanced processing technique.
Completed the construction of the enclosure for the dedicated welding hot cell.
Completed the first batch of irradiation experiments to produce helium-containing SS304 samples for use in development of weld repair techniques.
Completed an analysis of the microstructure and basic properties of the procured advanced alloys for the advanced radiation resistant materials program.
Completed the Final Expanded Materials Degradation Assessment.
Identified concrete cores for acquisition from Zion Unit 2.
Risk-Informed Systems Analysis Completed the RELAP-7 Theory Manual.
Tested the RISMC methodology using an LWR case study for enhanced accident tolerance design changes.
Completed a detailed demonstration case study for an emergent issue using RAVEN and RELAP-7.
Completed a report of demonstration of nonlinear seismic soil structure interaction and applicability to new system fragility curves.
Completed a preliminary design plan covering the requirements, development, and important physics for severe accident analysis.
Documented the approach and results obtained from the modeling and simulation for accident tolerant fuel under accident conditions.
Completed a RISMC case study for external event tolerant design changes.
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Completed the RELAP-7 subchannel flow capability development.
Completed the RELAP-7 verification and validation plan.
Completed a report on a more detailed BWR station blackout simulations.
Developed an approach for models that will be used to represent concrete degradation in Grizzly.
Plant Modernization Developed diagnostic and prognostic models for generator step-up transformers.
Developed a probabilistic health monitoring framework and demonstrated the problem of aging concrete structures.
Completed an interim report on the results of the concrete degradation mechanisms and an online monitoring techniques survey.
Completed a computer-based procedures validation study with nuclear power plant personnel.
Developed requirements for control room computer-based procedures.
Developed operator performance metrics for use in control room modernization projects.
Completed a human factors engineering design phase report for control room mode.
Developed a methodology for conducting baseline human factors and ergonomics review using a host nuclear power plant control room.
Completed a Control Room Upgrades Benefit Study Plan describing the study methodologies, industry partners, cost, schedule, facilities, and resources.
Identified advanced outage functions, including the results of the real-time support task involving coordination and automated work status updating.
Developed the requirements for automated work package technologies for a sample of nuclear power plant work processes.
Implemented software tools in the HSSL, which enable fully functional hybrid control room systems.
Completed a cyber security program evaluation exercise for the pilot project technologies.
Fiscal Year 2013 Materials Research Examined reactor surveillance materials from the Ringhals and Ginna nuclear power plants.
Executed small-angle neutron scattering experiments of irradiated RPV materials.
Obtained high-strength Ni-base alloys from service and began post-irradiation examination.
Completed the first stage of SCC initiation testing on alloy 600.
Measured SCC initiation response in alloy 690, including the effects of cold work.
Analyzed the recent characterization of irradiated specimens and irradiation-induced phase transformations.
Assessed thermodynamic and kinetic properties for the model development of phase transformations.
Developed plans for the acquisition and testing of baffle bolts from the Ginna nuclear power plant.
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Completed tensile tests of 316 SS base metal specimens and 316 SS - 316 SS similar metal weld specimens under room and elevated temperature and fatigue testing of 316 SS base metal specimens under room temperature and continued activities on mechanistic modeling.
Completed a study on mechanisms and mitigation strategies for IASCC of austenitic steels.
Executed constant extension rate tests in 320°C water to determine the effect of dose, alloy, and environment on SCC susceptibility.
Analyzed deformation mode changes in irradiated materials using bend tests and finite element modeling.
Completed a preliminary listing of aging conditions and measurement methods for physical properties to be examined for key indicators of cable aging.
Completed an aging assessment of field returned cables from the High Flux Isotope Reactor, the Zion Nuclear Power Station, and the Comision National Energia Atomica (Argentina)
Completed measurements of the physical properties on cables subjected to a range of accelerated aging conditions and assessed the results for key early indicators of cable aging.
Completed initial rejuvenation and tensile tests on cable specimens.
Completed risk-informed guidelines for evaluating the performance of aging safety-related concrete SSCs.
Completed the validation of data contained in the concrete performance database and placed the database into the public domain.
Identified the state-of-the-art on non-destructive testing methods for the assessment of nuclear power plant concrete materials and structures and available concrete samples for NDE testing and evaluated ultrasonic techniques.
Defined the envelope of the radiation at the biological shield wall for U.S. commercial nuclear power plants through 80 years.
Following tests conducted on base alloys, initiated fatigue tests on welded specimens in air.
Completed proactive welding stress control model development.
Risk-Informed Systems Analysis Completed a technical basis report describing how to perform safety margin configuration risk management.
Upgraded RELAP-7 capabilities through the implementation of a seven-equation two-phase flow model, including selected major physical components for BWR primary and safety systems.
Performed a RELAP-7 and RAVEN simulation of a station blackout scenario on the simplified geometry of a BWR.
Demonstrated the proof of concept of the Grizzly component aging model as applied to the RPV.
Demonstrated the modeling of late blooming phases and precipitation kinetics in aging RPV steels.
Plant Modernization Completed a technical report on the measures, sensors, algorithms, and methods for monitoring active aging and degradation phenomena for large power transformers and emergency diesel generators.
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Developed a system for the generic implementation of wireless technologies for equipment condition monitoring and its application in a commercial nuclear power plant.
Completed the evaluation for a final computer-based procedure prototype for field workers.
Developed Digital Control Room Upgrades reference guide in a human factors engineering plan for an optimized, human-factored control board layout.
Developed technologies for an advanced Outage Control Center that improves outage coordination, problem resolution, and outage risk management.
Completed the assembly of the HSSL and demonstrated its capability to model a hybrid (analog and digital) nuclear power plant control room.
Advanced Light Water Reactor Nuclear Fuels Documented SiC joining and irradiation studies and irradiation test preparation activities.
Completed SiC ceramic matrix composite failure mode analysis.
Completed fabrication of SiC ceramic matrix composite - zirconium alloy hybrid cladding prototype samples.
Completed the LWRS Program Fuel Development Material Inventory Database.
Completed the plan for transitioning LWRS Program Fuels activities to the Fuel Cycle Technologies Program Advanced Fuels Campaign and established a path forward for communication/coordination between the RISMC Pathway and the Advanced Fuels Campaign.
Fiscal Year 2012 Materials Research Completed the Expanded Proactive Materials Degradation Assessment Report.
Completed a planning document on concrete measurements to be performed at the Barsebck nuclear power plant.
Completed a report on metallurgical examination of the high-fluence RPV specimens from the Ringhals nuclear power plants in Sweden.
Completed an examination of reactor surveillance specimens from the Ginna, Ringhals, and Palisades nuclear power plants.
Completed the initial assessment for the feasibility of obtaining concrete core samples from identified candidate sites.
Completed the plan for the collection of materials from the Nine Mile Point 1 nuclear power plant during their 2013 outage.
Documented results of examinations of the surveillance specimens from the Ginna and Palisades nuclear power plant reactors.
Developed guidelines for risk-informed condition assessment and evaluation of aging concrete.
Completed NDE roadmaps for:
- Concrete R&D)
- Cables R&D
- Fatigue damage R&D 80
- RPV R&D.
Completed upgrades to test equipment for the evaluation of advanced weldments on irradiated materials.
Completed a plan for the modeling of high-fluence phase transformations in core internals.
Completed a report on high-fluence effects on microstructural evolution of irradiated materials.
Completed a plan for the modeling of high-fluence swelling effects in core internals.
Completed a report on evaluating the influence of bulk and surface microstructures on alloy 600 SCC initiation behavior.
Completed a report on the high-fluence effects on IASCC of stainless steels.
Completed a research plan for surrogate materials and attenuation studies, building on RPV results and findings in Fiscal Years 2009 to 2012.
Completed review of potential replacement alloys for LWRs.
Risk-Informed Systems Analysis Completed a verification and validation strategy for LWRS Program modeling and simulation activities.
Demonstrated the RISMC methodology using a test case based on INLs Advanced Test Reactor (ATR).
Completed the RELAP-7 development plan (funded by DOEs Nuclear Energy Advanced Modeling and Simulation Program).
Demonstrated a single-phase, steady-state version of RELAP-7 (funded by DOEs Nuclear Energy Advanced Modeling and Simulation Program).
Completed the RELAP-7 quality assurance plan (funded by DOEs Nuclear Energy Advanced Modeling and Simulation Program).
Completed an initial demonstration of the Grizzly model for pressurized thermal shock effects on an aged section of a pressurized water reactor RPV and assessed the through-wall attenuation effects of embrittlement.
Completed the plan for RELAP-7 support of a boiling water reactor major plant uprate analysis using RISMC.
Plant Modernization Completed the Advanced Instrumentation, Information, and Control Systems Technologies Pathway vision document.
Developed prototype technologies for nuclear power plant status control and field work processes, with associated study of field trials at a nuclear power plant.
Developed outage work status capabilities, providing a means for communicating work progress and completion status directly from the field activities to the nuclear power plant outage control centers.
Completed a digital full-scale mockup of a conventional nuclear power plant control room.
Developed guidelines and demonstration technologies for nuclear power plant operations and maintenance work processes.
Completed a report on outage emergent issue resolution capabilities.
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Completed a report on strategy and technical plans for online monitoring technologies in support of NDE deployment.
Completed a report on the online monitoring technical basis and analysis framework for large power transformers.
Completed a report on the demonstration of and data collection for prototype computer-based procedures.
Advanced Light Water Reactor Nuclear Fuels Completed the development plan for silicon carbide ceramic matrix composite (SiC CMC) nuclear fuel cladding.
Completed failure mode and performance analysis for SiC CMC.
Documented a plan to codify American Society for Testing and Materials standards for ceramic composites for nuclear applications.
Documented the required analyses to support irradiation readiness for SiC CMC rodlets in INLs ATR.
Completed the fuel clad trade-off study.
Documented the status of irradiation test preparation activities for the joining and irradiation studies.
Completed the design and installation of a nuclear fuel cladding test system that simulates nuclear fuel heating and provides a steam atmosphere.
Selected two industry proposals for SiC CMC joining technology development.
Fiscal Year 2011 Completed a report documenting information gaps on concrete performance and cable-aging degradation from an examination of the R. E. Ginna nuclear power plant during the calendar year 2011 refueling outage.
Published an implementation plan for the development of a nuclear concrete materials database.
Developed a baseline computational model for proactive welding stress management to suppress helium-induced cracking during weld repair.
Completed an initial assessment of thermal annealing needs and challenges; an assessment of needs for environmental fatigue under extended service conditions; an assessment of alloy options and performed alloy downselect evaluation of in-situ cable repair; and an assessment of further irradiation experiment needs.
Published the II&C industry working group vision document for II&C technologies.
Began pilot projects on real-time configuration management and control to overcome limitations with existing permanent instrumentation and real-time awareness of plant configurations.
Completed fueled irradiation safety case documentation to support irradiations in INLs ATR.
Completed initial evaluations of prototype fuel rodlets.
Completed a report documenting the results of collaborations with EPRI/industry to identify, define, and prioritize power uprate challenges and develop power uprate R&D strategies.
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Fiscal Year 2010 Completed a literature review on concrete durability and aging.
Completed the planning document for harvesting material from the R. E. Ginna and Nine Mile Point nuclear power plants.
Completed a report on architectural and algorithmic requirements for a next-generation system analysis code for that can be used to support the safety case of the LWR life extension.
Fiscal Year 2009 Fiscal year 2009 was primarily a planning year. The LWRS Program Plan was issued, a workshop on advanced fuel design was held, and a report on testing and analysis of reactor degradation was completed.
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