GIF Webinar Guide

Atoms for Peace. The Next Generation

This webinar provides a historical perspective on the Atoms for Peace program, which launched the development of nuclear power around the globe, and describes the current outlook for the development and deployment on the next generation of nuclear power (Generation IV).

Introduction to Nuclear Reactor Design

Why is a 4th generation of nuclear reactors needed? And what are the most promising reactor technologies? The GIF initiative has led to reconsider some of the options adopted in the past and stimulated the investigation of new tracks for long term sustainable nuclear energy. To grasp the rationale for selecting Generation IV reactor systems, and their main characteristics, requires some basic knowledge in the fundamentals of nuclear reactor design. What is behind the terms “criticality,” “breeding,” and “fast or thermal neutrons”? How to select the coolant, moderator, neutron spectrum, fuel materials and composition and to choose the ad hoc combinations to design nuclear reactors in line with Generation IV criteria, in particular sustainability? This is the objective of this rather technical webinar targeting civil society stakeholders.

Role of Nuclear Energy in reducing CO2 Emissions

"Nuclear energy is ready to contribute to global net zero". Some people think this sentence as fact, some people doubt and think too early to say due to technical or/and psychological reasons. Sustainable, safe, clean, reliable, flexible and affordable energy source for 24/7, that is we really want in our daily life. Therefore, people make efforts to develop these characteristics in each energy sources to fill gaps to the ideal and desirable energy.
In GIF, not only historical GIF goals (Sustainability, Economics, Safety and Reliability, Proliferation Resistance and Physical Protection), but also have activities in this field. Pursuit flexible characteristics with CEM NICE Future/COP26, release ESG asset report, and launch new TF, whose name is Non-Electric applications of Nuclear Heat: NEaNH, to draw commercialization map depended on market needs.
Off course, these kinds of approaches are taken in various organizations to obtain clear future vision in vague present. So, in this webinar, three worldwide leaders in this field, from GIF(NEaNH), IAEA, and NEA are invited to have each 20-minute presentations and following panel discussion. Let’s join future searching event for better tomorrow.

Guest Speakers:
Dr. Shannon Bragg-Sitton (GIF NEaNH-TF, INL), Accelerating Economy-Wide Decarbonization via Nuclear Energy
Mr. Wei Huang (IAEA NEPIK),Nuclear Energy, An Important Part of Solution to Net Zero World
Ms. Diane Cameron (OECD NEA),Meeting Climate Change Targets: The Role of Nuclear Energy

Moderators:
Dr. Patricia Paviet (GIF ETWG, PNNL)
Dr. Tatjana Jevremovic (IAEA WCRT)

Maximizing Clean Energy Integration: The Role of Nuclear and Renewable Technologies in Integrated Energy Systems

Many cities, states, utilities, and public commissions are setting energy standards that aim to reduce carbon emissions. In order to realize a clean and resilient energy future, new methods of energy production, distribution, and use will be required. The primary focus of the DOE Office of Nuclear Energy (DOE-NE) Program on Integrated Energy Systems, led by researchers at Idaho National Laboratory (INL), has been to assess the potential of integrated energy systems to enhance the flexibility and utilization of nuclear reactors alongside renewable generators and, thereby, to maximize the use of the clean energy provided by these systems. This work begins with the question: “What goals are we trying to achieve, and how will the produced energy be used?” These questions must be addressed within the context of a specific deployment location, which has implications relative to the electricity market structure, supply, and demand; available feedstock for industrial processes; and available product markets. Product streams, ranging from potable water to hydrogen, fertilizer, synthetic fuels, and various chemicals, have been considered. Each product stream has its own market and market drivers and its own geographic location that would maximize profitability. Some of these products would only require electricity to support production, while others require both thermal and electrical energy. This webinar highlights work led by INL, in collaboration with other national laboratories and industry partners, to evaluate integrated energy system options that utilize nuclear energy in new ways. By working with key collaborators in the nuclear industry, these analytical studies are now becoming a reality in demonstration projects.

GIF VHTR Hydrogen Production Project Management Board

The objective of the GIF VHTR Hydrogen Production Project Management Board is to provide a collaborative environment among the signatories for the development, optimization and demonstration of economical large-scale hydrogen production processes that do not emit green-house gases through the use of nuclear energy. The main processes considered by the signatories include Sulphur-Iodine (S-I), High Temperature Steam Electrolysis (HTSE), Copper-Chlorine (Cu-Cl) and Hybrid Sulphur (HyS). The signatories include Canada, EU, France, Japan, Korea and the USA. China has been an observer, waiting to join the group formally, but contributing strongly to the developments. The S-I process has been demonstrated for short term operation by China, Korea and Japan. EU, France and the USA have been very active in HTSE. Canada has been focusing on the Cu-Cl Cycle with plans for demonstration of an integrated lab-scale system in 2021. This webinar will provide an overview of these activities and their relevance to mitigating global warming.

Overview of Small Modular Reactor Technology Development

Nuclear electricity generation started with prototype- and test reactors of a small size and low power. Relatively quickly these were replaced by increasingly larger nuclear power plants due to increased needs, economy of scale and limited available sites. For several years the interest in small modular reactors (SMRs) has increased with over 50 concept designs now under development. The IAEA defines SMRs as advanced nuclear power plants with one or more individual modules that each produce electric power up to 300 MWe. A module may be built in factories and shipped to nuclear sites for installation and added as the need arises. All advanced technologies are included (water cooled, Gen-IV systems and micro-reactors). SMRs claim enhanced passive safety features, simplified design and operations, economy by numbers and the flexibility in hybrid energy systems and non-electric applications. The webinar highlights the attractive features of SMRs, major challenges, the current status of SMR technology and near-term deployment plans.

Global Potential for Small and Micro Reactor Systems to Provide Electricity Access

Small and micro-scale modular reactors have received considerable attention for their potential to reduce costs, load follow and meet electricity needs in places where the size of conventional reactor technologies is unwarranted. This small scale is particularly relevant in the developing world where large centralized grids are uncommon and the need for electricity is considerable. More than 1 billion people globally are currently estimated to live without access to any electricity. The Agenda for Sustainable Development calls for reliable, affordable and clean energy for all people by 2030, creating an additional imperative for rapid low carbon technological deployment. This talk will present a novel market analysis of near-term energy demand. We use state-of-the-art satellite imagery to identify regions with no night-time light as a proxy for electricity poverty, and ambient population to determine the number of persons in these regions. GIS is used to create corresponding maps showing the capacity needed to provide this degree of electricity as a function of location if only micro and mini-grids are available. Additional considerations including resilience to natural hazards, siting considerations and competitive technologies are discussed.

MicroReactors: A Technology Option for Accelerated Innovation

Micro-reactors are very small nuclear reactors capable of operating independently from the electric grid to supply highly resilient power, and are well suited to serve the power needs for remote communities that currently do not have access to reliable, resilient and affordable energy. A typical commercial microreactor is envisioned to be a mobile nuclear power plant in a 2-20 MWe range that is fully factory built, fueled and assembled. It is transportable to the remote site via ground, sea or air with black start, renewable integration and island mode operation capability. They are designed to be self-regulating and walk-away safe with minimal operator intervention. NEI estimates that Microreactors could deliver electricity at rates between $0.09/kWh and $0.33/kWh. This presentation will describe ‘genericized’ microreactor designs being pursued by various vendors, technology gaps and the role of DOE’s Microreactor R&D.

Evaluating Changing Paradigms Across the Nuclear Industry (MicroReactors)

Dr. Lovering’s recent work focuses on microreactors (SMRs <10MWe), trying to understand the pathways to commercialization and economic competitiveness. To understand their potential, a techno-economic evaluation of microreactors for off-grid and community microgrid applications was first performed. The results indicate that microreactors can be cheaper and more reliable compared with 100% renewables systems, and they can also be cost-competitive with diesel where fuel costs are greater than $1/liter and the microreactor capital cost is less than $15,000/kW. However, the levelized cost of electricity (LCOE) for microreactors is most sensitive to the initial capital cost, and whether this technology will ever move beyond niche markets will depend on the learning effects accrued through factory fabrication. Therefore, the hypothetical trade-offs between economies of scale and economies of volume for potential factory-fabricated microreactors are also examined. The break-even volumes necessary for microreactors to become cost-competitive with large reactors and with fossil fuels, using parameters from historic nuclear builds and analogous energy technologies are calculated. Drawing from the literature on learning rates across energy technologies, potential learning rates for various sized microreactors based on historical relations are predicted.

Opportunities for Generation-IV Reactors Designers through Advanced Manufacturing Techniques

The development of critical design criteria for new advanced reactor systems, components, and materials requires an understanding of both fabrication and the irradiation environment during normal operating and accident conditions. Next-generation researchers and designers are therefore challenged not only by demands for improved performance, they must also work to shorten the development and commercialization lifecycle for new nuclear reactors and systems to remain competitive. This provides unique and exciting opportunities for all contributors to this field of study. This presentation will offer a strategic overview of the impact that advanced manufacturing has on the lifecycle of new generation reactors. By evaluating state-of-the-art practices found in other large manufacturing industries, this presentation provides an overview of major innovation areas that are considered to benefit the GEN-IV systems (SFR, GFR, LFR, FSMR...). Synergetic advanced manufacturing approaches beneficial to the collective GEN-IV systems, with some examples of differentiating approaches necessary for specific reactor designs, are discussed. Furthermore, new paradigms in licensing approaches for additively manufactured parts will be discussed.

Performance Assessments for Fuels and Materials for Advanced Nuclear Reactors

A host of novel fuel and material concepts are being investigated as part of the GenIV reactor development initiative. While many of these candidates are rooted in historical programs from previous reactor development campaigns, most of these concepts were never fully evalauted for long-term performance in non-LWR facilities. The performance data that is needed for candidate material downselection, feasibility studies, and eventual qualification is, currently, very costly in terms of monetary cost and human capital. The use of an ‘all of the above’ strategy for performance assessment is needed to reduce the cost of ushering materials through the qualification process. In this presentation, we will discuss the efforts that are currently underway, and those planned for the near future, to advance many of these candidates from concept to deployment.

Artificial Intelligence in Support of NE Sector

This lecture will start with brief introduction to Artificial Intelligence and will fast forward to application of AI in design, construction and operation of nuclear power plants in future. Technical focus will be on several practical examples including use of machine learning and deep learning neural networks for weld inspections, sonar detection, material testing and Natural Language Processing (NLP) for conducting semantic search of documents. Aim is to show how engineers can de-risk and accelerate engineering projects with help of AI.

EPRI Virtual Reality Training

Effective training for employees is an ongoing need for utilities, whether it is for new hires, or personnel changing jobs, new equipment and processes, or changes in the industry. Virtual Reality (VR) training is being used more prominently across many industries and has been shown in certain cases to improve results related to safety, quality, operations, and cost savings. This presentation will cover how we can use VR training in the nuclear industry, and EPRI’s new paradigm for VR training.

The Nuclear Workforce of the Future - Opportunities and Needs for the International Nuclear Sector

Although nuclear energy is by far the largest producer of carbon free electricity on the planet, and nuclear applications in healthcare have been saving lives for decades, most people go through life unaware of the career opportunities within the nuclear sector. The need to attract and retain talented people within the nuclear sector has always been important, but as the sector grows it is now becoming an imperative. Questions and topics addressed in this webinar include:
- What does the nuclear workforce look like today, in terms of volumes, locations and skillsets of people in the sector?
- How does the workforce need to evolve and develop to meet the future ambitions of the nuclear sector?
- Increasing awareness of the sector itself and the exciting and purposeful career opportunities that exist.
- Addressing some of the barriers to attracting, recruiting and retaining talent in the nuclear sector.
- Recognising the differentiators that mean hundreds of thousands of people stay with the nuclear sector for their whole career.
- How to find out about career opportunities and either enter or expand your career within the nuclear sector.

MOOK: The Knowledge Management Method Applied to a Gen IV Project, The Continuation of a Successful Story

The nuclear renaissance we are all living worldwide will need a significant increase of manpower and young generation.
- But how can we guarantee that this new generation will gain the most and the best from the older and from the huge experience already gained from GEN IV reactors?
- How can we use methods being attractive, innovative, efficient and time saving?
- How can we apply these methods of Knowledge Management to all kinds of persons (experts, technicians, scientists, head of project) and with all kinds of systems?
This webinar will present the innovative Knowledge Management methods that were developed and successfully experienced on the ASTRID project: a GEN IV Sodium Fast Reactor project developed from 2009 to 2018 at CEA. In particular, the method and its specific tool of Knowledge Management called MOOK will be presented, and its feedback shared with you.

Revolutionizing Nuclear Engineering Education: Developing Virtual Labs for Neutron Detection, Geiger Counter, and Reactor Experiments

In recent years, the Nuclear Industry has been facing the challenge to bridge the gap in technological fluency between generations, as highlighted in the “NSUF University Research Reactor Fitness Study Report” (2019) conducted by Idaho National Labs in collaboration with research reactors in the United States. This gap poses a significant threat to the industry, especially as the world is striving towards global decbarbonization to combat climate change. This urgency prompted our research group to develop a project with the potential to revolutionize the industry and address this challenge by bringing about a nuclear energy renaissance globally. This research project involves developing Virtual Laboratories (Virtual Labs) for nuclear engineering, which will greatly enhance the training experience of students and engineers in the industry. With the math parameterized on Matlab scripts with respect to the physics behind the neutron detector lab, Geiger counter lab, and reactor lab, the scripts could then be converted and developed in C# using the Unity game engine. The Neutron detection Virtual Lab offers instructor-tailored procedures for setting up and plugging of nuclear electronics and aims to develop modular virtual lab electronics for broader applications. The Geiger counter lab simulates radiation detection using the dose rate equation and gamma attenuation equation, with plans to move from 2D to an augmented reality 3D representation. The virtual reactor lab uses point kinetics (PKE) to simulate the PUR-1 light water open pool reactor located at Purdue University, with actual reactor data for benchmarking. The development of these virtual labs not only has the means of bridging the technological fluency and generational gap thus identified in the nuclear industry and assisting in decarbonization efforts but also the potential to integrate with AI (Artificial Intelligence) and take our industry to that next level.

Overview of Canadian R&D Capabilities to Support Advanced Reactors

Growing concerns regarding energy security and climate change are driving increased interest in nuclear power as a clean energy option, including small modular reactors (SMRs) that employ a range of technologies and open new roles and markets. SMRs are advanced nuclear reactors that have a power capacity of up to 300 MW(e) per unit which extend beyond traditional reactors, offering the opportunity of safer, cheaper, and more efficient generation of emissions-free electricity, as well as heat for industrial processes. Various institutions within Canada such as universities and Canadian Nuclear Laboratories are conducting research, collaborating at the national and international level and are participating in Government of Canada programs to establish capabilities to support the safe design, deployment and future operation of SMRs. An overview snapshot of the SMR deployment landscape in Canada and innovative supporting research within Canada will be discussed.

Safety of Generation IV Reactors

Excellence in safety and reliability is among the goals identified in the technology roadmap for Generation IV nuclear reactors. This webinar will give an overview of the activities of the GIF Risk and Safety Working Group done in support of the six Generation IV nuclear energy systems towards the fulfilment of this goal. Topics include a presentation of the safety philosophy for Generation IV systems, the current safety framework for advanced reactors, and the methodology developed by the group for the safety assessment of Generation IV designs. Other ongoing activities between the group and the designers of Generation IV systems will be also highlighted.

SFR Safety Design Criteria (SDC) and Safety Design Guidelines (SDGs)

This webinar provides the outlines of the safety design criteria (SDC) and safety design guidelines (SDG) established to achieve high development goals of Gen IV reactors including safety and reliability. Reflecting the lessons learned from the Fukushima Daiichi nuclear power plant accident, the SDC describes requirements that must be met by Gen IV Sodium-cooled Fast Reactors (SFRs), and the SDG provides guidelines on how to apply the SDC to the actual design. The Gen IV SFRs are required to adopt advanced devices and systems as a built-in safety feature, combinations of active safety systems with passive mechanisms or inherent features to prevent and mitigate core damage. Taking the charactristics of the SFR as liquid metal cooling fast reactor system into account, the SDG recommends spesific design measures such as inherent / passive reactor shutdown, natural circulation decay heat removal and in-vessel retention of degraded core.

Experience of HTTR Licensing for Japan’s New Nuclear Regulation

On June 3rd, 2020, JAEA obtained the permission of reactor installation change for restarting the High Temperature Engineering Test Reactor (HTTR, 30 MW) from the Nuclear Regulation Authority (NRA) of Japan. This is the first restart permission for a gas-cooled reactor in Japan under the New Regulatory Requirements. By employing a high-temperature-resistant and large-heat-capacity graphite core, HTTR enables inherent safety characteristic of slow and limited temperature transient without fuel damage even in a loss of coolant accident. The safety review by NRA against the New Regulatory Requirements has confirmed that no fuel damage would occur even in the event of a beyond design basis accident such as multiple losses of reactor shutdown functions. With the aim of early restarting, the refurbishment to the HTTR as mandated by the permission of changes to reactor installation is steadily carrying out, including installation of countermeasure systems against internal and external fires and so on. In this Webinar, an experience of HTTR licensing for Japan’s New Nuclear Regulation will be introduced.

ESFR SMART a European Sodium Fast Reactor concept including the European feedback experience and the new safety commitments following Fukushima accident

The ESFR-SMART project is what in the Anglo-Saxon world is called a "working horse" or a "concept car". Its role is to introduce, outside any constructive planning, new ideas for the future, which can be valuable guides for R&D. Unlike in an "industrial" project, which initially had a construction schedule, one can introduce innovative ideas, even if their lower technological-readiness level would require development and time. For these new ideas, research and first calculations were also performed during the project, to check their general feasibility and the absence of major impossibilities. The presentation will show the design proposition for this SFR reactor including the experience feedback from European experience on Phénix, Superphénix, EFR and ASTRID. This design includes all improvements in response to the last safety rules issued after the Fukushima incident. Finally, the first calculations performed for the design validation are also explained with the necessary R&D.

Estimating Costs of Generation IV Systems

This webinar will provide an overview of the Economic Modelling Working Group’s Cost Estimating Guidelines for Generation IV Nuclear Energy Systems (GIF, 2007). Topics include an overview of the Guidelines, a comparison of the Guidelines with other nuclear power plant cost estimating models, and a discussion of benchmarking activities by the EMWG with INPRO.

Graded Approach: Not just Why and When, but How

Standards and regulations in many countries discuss graded approach; some even require it. Criteria or justifications for grading are commonly addressed. Not much, however, is discussed about the methods that can be used to grade a process once the criteria are met. This webinar will remove any mystery associated with graded approach. Mr. Chermak asserts there are only two ways to grade one’s approach to Quality Assurance - and they are very simple. We look forward to your company while we learn about and delve into graded approach.

Closing Nuclear Fuel Cycle

The steps of PWR nuclear fuel cycle along with alternative fuel cycle options are described. The concepts of two methods for closing the fuel cycle, i.e., recovering the residual uranium and plutonium contained in spent fuel for reuse by wet PUREX and dry Pyroprocessing, are explained. The major issues to be considered for closing the fuel cycle are identified to provide an understanding of sustainability and nonproliferation.

Sustainability a Powerful and Relevant Approach for Defining Future Nuclear Fuel Cycles

Technically, nuclear energy is anticipated to be one of the most efficient energy source to mitigate the global climate change together with the renewables, due to its low green-house-gases emissions, its reliability and its high base-load capacity. However, public opinion survey and phase-out decision regularly reminds us that political decisions are not only driven by technical criteria. Beyond the well-known technical and economic optimization, many other criteria are of growing importance such as environmental and social concerns. This rather recent situation requires changing our rationale technical approach to the wider sustainability approach, which also includes the overall environmental footprint and the more general social acceptability and social impact. This presentation will illustrate how sustainability can help us to identify the most promising trends for future nuclear fuel cycles in order to ensure a long-term future of nuclear energy.

Scientific and Technical Problems of Closed Nuclear Fuel Cycle in Two-Component Nuclear Energetics

The webinar presents the overview of scientific and technical problems of closed nuclear fuel cycle in two-component nuclear energetics. The presentation will highlight the existing problems of the current technological platform of NE (thermal reactors in an open nuclear fuel cycle) and the advantages of the new technological platform (fast reactors with closed nuclear fuel cycle). Latest developments associated with the use of mixed UN fuel & spent nuclear fuel reprocessing are briefly presented as well. The remaining research challenges of the new technological platform being developed within the “Proryv” Project framework are summarized in the light of the present technology understanding.

Comparison of 16 reactors neutronic performance in closed Th-U and U-Pu cycles

Just as in all other industries, sustainability is vital to nuclear energy production. Recycling of nuclear fuel contributes to the environmental and social pillars of that sustainability because it simultaneously improves natural resources utilization and waste minimization. This webinar provides additional insight to the consequences of repetitive fuel recycling and compares selected reactors based on their neutronics performance in the closed Th-U and U-Pu cycles. Because the closed fuel cycle has been discussed in several previous GIF webinars, this presentation focuses on less common perspectives. The closed fuel cycle will be presented as a Bateman equation eigenstate. In several cases, the eigenstate will be achieved by irradiation of subcritical fuels. It will be shown that all reactors in the respective fuel cycle have, by chance, the same average neutron production per fission. Hence, the usual measure η-2 will be replaced by fission probability discussion. Although the Bateman equation eigenstate in this comparative study is reached without fission products, their role in the closed cycle will be addressed.

Proliferation Resistance and Physical Protection of Generation IV Reactor Systems

This webinar will provide an overview of the activities of the Generation IV Proliferation Resistance and Physical Protection Working Group. Topics include a presentation of the methodology developed by the group, an illustration of the methodology to an example nuclear system, and a summary of ongoing interactions between the group and the designers of the six Generation IV nuclear energy systems. Other outreach activities of the group associated with various national and international organizations will be briefly summarized.

Advanced Reactor Safeguards and Material Accountancy Challenges

The worldwide expansion of nuclear energy is likely to be led by the deployment of advanced and small modular reactors. These reactors are enabled by significant private and public support and hope to capitalize on lower upfront costs, modular construction, enhanced safety, and better integration in the power grid. The Generation IV International Forum seeks to enable the safe and secure expansion of nuclear energy through eight goals covering sustainability, economics, safety & reliability, and proliferation resistance & physical protection. Nuclear material safeguards are a key consideration of the proliferation resistance goal. Advanced reactor vendors, depending on deployment location, will need to meet both domestic material control and accounting (MC&A) regulatory requirements as well as international safeguards obligations. Many of the advanced reactor designs have limited regulatory experience and must navigate a regulatory structure that was built up around large light water reactors (LWRs). The use of different coolants, fuels, sealed cores, and variations in fuel handling all present MC&A and safeguards challenges. This webinar will provide an overview of materials accountancy challenges for advanced reactors, the differences between domestic and international safeguards requirements, and proliferation resistance goals as defined by the Generation IV International Forum. The presentation will be built around the six Generation-IV systems describing key challenges for each area and how new safeguards approaches and technologies can enable new nuclear deployment.

Overview of Waste Treatment Plant, Hanford Site

Currently, the U.S. Department of Energy (DOE) stores ~90 million gallons of radioactive and hazardous waste in ~230 underground tanks at Hanford and Savannah River. At Hanford, approximately 20 million gallons of that waste is in a liquid form (supernatant), approximately 10 million gallons is in the form of insoluble sludge materials, and the remainder is in a partially soluble solid form referred to as saltcake. Treatment and immobilization of the tank waste into a glass waste form is planned with the Hanford Waste Treatment and Immobilization Plant (WTP) being the principal plant where this will be accomplished. This webinar focuses on the integrated flowsheet that encompasses storage, retrieval, pretreatment, immobilization, and disposal. The major emphasis or focal point will be the vitrification with respect to: 1) Troublesome waste components and their impact on glass formulation/operations; 2) Critical process and product performance properties (why and how they are measured); 3) Process control strategies and use/impact of glass models/algorithms; 4) Relationship between acceptable glass compositional regions and operational flexibility; 5) Significant advancements in glass formulation and the impact on the flowsheet/operations; 6) Operational lessons learned.

Nuclear Waste Management Strategy for Molten Salt Reactor Systems

Molten salt reactors (MSRs) are being considered as one of the potential nuclear options to meet future energy demands. While the MSR designs differ from the more traditional light water reactor, many of the waste streams are similar between the concepts. Other streams differ drastically. This presentation outlines strategies for the treatment and processing of MSR-type wastes from concepts of reconditioning and recycle of certain components to partitioning and direct immobilization of other waste components. The primary focus will be on the management of fuel salt related wastes while other streams unique to MSRs are also briefly discussed.

Safe Final Disposal of Spent Nuclear Fuel in Finland

This webinar highlights the history, describes the status, and examines the strategy for safe final disposal of spent nuclear fuel in Finland. A privately held company, Posiva Solutions Oy, is responsible for the final disposal of spent nuclear fuel of its owners. Posiva is charged with research & development, planning and design, as well as the construction and operation of the encapsulation plant and final disposal facility. Following the site investigation (1983-1999), Posiva applied for a decision-in-principle from the Finnish government and parliament. It was granted 2001. In 2012, Posiva applied for the construction permit, which was authorized in 2015, and started preparatory construction works in 2016. Construction works of the encapsulation plant and the excavation of the underground facilities started in 2019 and will be ready in 2023. Posiva submitted the operating permit application to the Finnish government in December 2021.A Trial Run of Final Disposal (TRFD) will be carried out in 2024, prior to the estimated start of the operation in 2025. Safeguards are applied for the first time for a spent fuel repository.

Sodium Cooled Fast Reactors (SFR)

This webinar will give an overview of distinctive fast reactor characteristics and identify key performance benefits. A brief history of development and international experience with SFRs will be reviewed. Finally, the Generation-IV international collaboration on SFR technology research and development will be described.

European Sodium Fast Reactor: An Introduction

This webinar presents a brief history of the conceptual development of a large-power (3600 MWth) European Sodium Fast Reactor (ESFR), discusses the status of the current R&D activities on Generation-IV ESFR safety enhancements of the Horison-2020 ESFRSMART project, and provides an overview of new safety measures proposed for improvement of the three safety functions: reactivity control, heat removal and radioactivity containment. Also, experimental programs currently on-going in Europe in support of the ESFR R&D are briefly introduced. A summary of the activities to be performed during the next phase of the project concludes the webinar.

Experimental R&D in Russia to Justify Sodium Fast Reactors

Development of new reactor systems calls for justification of safety and operability. Not all the phenomena can be justified by way of calculations. Many systems have been developed for the first time, and some systems are functioning under the totally new conditions. The main requirement to the release of an engineering design is validity of the decisions made. In Russia, a fleet of test facilities has been designed for experimental studies in the area of liquid-metal cooled fast reactors. Most of the test facilities use liquid metal as a coolant, but there are test facilities where water is used as a coolant. In this webinar, the elements and systems of the Generation IV nuclear reactors will be discussed and illustrated by the BN-1200 reactor, and consideration will be given to the experimental R&D carried out in Russia to justify them.

Lead Fast Reactor (LFR)

The Lead-cooled Fast Reactor (LFR) is characterized by a fast neutron spectrum; a liquid coolant with a very high margin to boiling and relatively inert interaction with air or water; and design features that capitalize on these attributes. As with other fast spectrum reactors, the LFR offers fuel cycle options that greatly enhance resource utilization and sustainability. LFR concepts offer great potential in terms of safety, simplification, proliferation resistance and the economic performance. The webinar presents background on fast reactor physics, the historical development and present status of LFR technology and the main characteristics of LFR concepts under current consideration.

Advanced Lead Fast Reactor European Demonstrator - ALFRED Project

The webinar presents the main design features of the ALFRED nuclear reactor demonstrator as developed in the frame of the Collaborative projects funded by the European Community Framework Program. The presentation will provide an overview of specific design solutions, safety approach and safety characteristics of ALFRED, touching the most important aspects of the demonstrator. Latest developments are briefly presented as well. The remaining research challenges are then explained at the light of the present technology understanding to highlight the present status of knowledge and further steps to be pursued.

MYRRHA an Accelerator Driven System Based on LFR Technology

SCK•CEN is actively working on designing and building a new multifunctional research installation: MYRRHA as in Multi-purpose hYbrid Research Reactor for High-tech Applications. This webinar will present the MYRRHA project, an accelerator driven system coupling a sub-critical Pb-Bi cooled reactor and a high power proton accelerator through a spallation target which is the very first prototype of a nuclear reactor driven by a particle accelerator in the world. As an external source of neutrons, this particle accelerator maintains the nuclear fission chain reaction. It is referred to as a subcritical reactor: the core does not contain enough fissile material to spontaneously maintain the chain reaction. This innovative nuclear technology is safe and easy to control. When the particle accelerator is stopped, the chain reaction also stops automatically within a fraction of a second.

Gas Cooled Fast Reactor (GFR)

The Gas Cooled Fast Reactor (GFR) is one of the six promising technologies selected by the GIF. The presentation summarizes the main advantages and drawbacks of GFRs and the key design and safety issues as well as the related research and development programs.

The ALLEGRO Experimental Gas-Cooled Fast Reactor Project

The webinar presents the main design features of the ALLEGRO nuclear reactor demonstrator as developed in the frame of the European V4G4 Consortium “V4G4 Centre of Excellence” associating nuclear research organizations from the Czech Republic, Hungary, Poland, Slovakia and France. The presentation provides an overview of the existing concepts of ALLEGRO, goals of the development, specific design solutions, and the safety approach and safety characteristics of ALLEGRO, touching the most important aspects of the demonstrator. Latest developments associated with both the use of UOX fuel and the new safety features are briefly presented as well. The remaining research challenges are summarized in the light of the present technology understanding to highlight the present status of knowledge and further steps to be pursued.

Very High Temperature Reactors (VHTR)

Among the six Generation IV concepts eventually selected for international cooperative development, the Very High Temperature Reactor (VHTR) was seen as an early favorite among many of the members. Indeed, among the seven original members of the VHTR System Arrangement (SA), three had already operated or tested high temperature gas-cooled reactors. The accession of the People’s Republic of China to the VHTR SA in 2008 brought that number to five. This presentation will describe how the continued cooperative development of the VHTR concept as a Generation IV system will deliver on nuclear energy’s promises of sustainable, economic, safe, reliable and proliferation resistant power and energy supply.

Design, Safety Features and Progress of the HTR-PM

The high-temperature gas-cooled reactor pebble-bed module (HTR-PM) is aimed to extend nuclear energy application beyond the grid, including cogeneration, high-temperature heat utilization, and hydrogen production. The first concrete of HTR-PM demonstration power plant, which has been approved as part of the National Science and Technology Major Projects, was poured five years ago, in Rongcheng, Shandong Province, China. The thermal power of a single HTR-PM reactor module is 250 MWth, the helium temperatures at the reactor core inlet/outlet are 250/750 °C, and a steam of 13.25 MPa/567 °C is produced at the steam generator outlet. Two HTR-PM reactor modules are connected to a steam turbine to form a 210 MWe nuclear power plant. The progress of HTR-PM project in China has drawn considerable attention worldwide. In this webinar, the design basis, design principles, general design features and safety characteristics of HTR-PM will be presented. Main engineering verification experiments of components and systems for the HTR-PM, such as helium blower, steam generator, will be introduced. Progress of the HTR-PM demonstration power plant, including civil engineering, first-of-a-kind equipment manufacturing, licensing, installation of the main equipment, will be described. In addition, the irradiation test results of pebble fuel samples and the status of commercial fuel production plant will be explained.

Supercritical Water Cooled Reactors (SCWR)

Supercritical Water-Cooled Reactors (SCWRs) are a class of high temperature, high pressure water-cooled reactors that operate above the thermodynamic critical point of water (374°C, 22.1 MPa). These concepts combine the design and operation experience gained from hundreds of water-cooled reactors with the experience from hundreds of fossil-fired power plants operated with supercritical water. The main goals of using supercritical water in nuclear reactors are to increase the efficiency of modern nuclear power plants, decrease capital and operational costs, and finally decrease electrical energy costs. This presentation describes SCWR concepts being pursued in the international community and highlights the technical advancements and challenges in the development.

Molten Salt Reactors Taxonomy and Fuel Cycle Performance

As the title says, Molten Salt Reactors (MSR) utilize molten salt in their core. Especially, when it is used as a fuel carrier, it provides enormous designing freedom. Consequently, there are many concepts available in the literature; some more matured than others. Recently there was a common effort at IAEA to classify these concepts and create MSR taxonomy. This taxonomy will be briefly introduced in the seminar together with the reactor physics characterization of the major MSR classes and families. Particular MSR features will be discussed and listed together with the involved families. The fuel cycle performance will be evaluated from neutronics perspective. It will include discussion about cross-section of applicable material for MSR construction. Several MSR families will be classified according to the breeding gain and applicable fuel cycles. Last but not least, salt composition and distribution in the MSR system during its operation will be presented together with an initial assessment of possible radionuclides release during accidental conditions.

Overview of FHR Technology

Fluoride Salt Cooled High Temperature Reactors (FHRs) use solid, ceramic fuel with a molten salt coolant, and deliver heat in the temperature range from 600°C to 700°C. This presentation will review key design features of FHRs and recent work to develop the technical basis for safety analysis and licensing.

Concept of European Molten Salt Fast Reactor (MSFR)

Liquid-fueled reactors exhibit unusual and interesting properties in terms of operation and safety compared to solid-fueled reactors, requesting a revision of some well-known conception and safety rules. In this webinar, such characteristics of the Molten Salt Reactors (MSRs) will be presented, together with the past and current R&D activities. The concepts studied in the frame of the Generation-IV international collaboration will be briefly described, and the presentation will then focus on the concept of Molten Salt Fast Reactor (MSFR), reactor based on a fast spectrum and studied since almost a decade mainly by calculations and determination of basic physical and chemical properties, initially at CNRS in France and now more largely in the European Union. The main design choices and characteristics of this MSFR concept will be explained and discussed including transient simulations, chemistry and material issues, safety analysis, research roadmap and laboratory scale experiments.

Molten Salt Actinide Recycler and Transforming System with and without Th-U support: MOSART

The Molten Salt Reactor designs, where fissile material is dissolved in the molten salt fluorides, under consideration in the frame work of the GIF are briefly described. The presentation mainly focuses on the MOlten Salt Actinide Recycler & Transforming (MOSART) system without and with U-Th support fueled with different compositions of transuranic elements trifluorides from spent LWR fuel. New design options with homogeneous core and fuel salt with high enough solubility for transuranic elements trifluorides are being examined at NRC “Kurchatov Institute” because of new goals. The webinar has the main objective of presenting the fuel cycle flexibility of the MOSART system while accounting technical constrains and experimental data received in this study. A description is given of the experimental results on key physical and chemical properties of fuel salt and combined materials compatibility to satisfy MOSART system requirements. In the webinar the main design choices and characteristics of MOSART concept are explained and discussed including safety, transient simulations, laboratory scale experiments and program plan for the development of the small power Demo MOSART unit.

Czech Experimental Program on MSR Technology Development

The webinar will give an overview of the existing experimental development of Molten Salt Reactor (MSR) technology in the Czech Republic. A technology of nuclear reactor systems with liquid molten salt fuel has been investigated in the Czech Republic since 1999. After 2005, the studies cover also thorium – uranium fuel cycle technology, material research and development of selected components of the MSR technology. Today a new, four-year (2017 – 2020) project of MSR technology development is the key component of the Czech MSR R&D program on fluoride salt-cooled nuclear reactor systems. The project is a follow-up and broadening of existing Czech activities in MSR. The aim of the project is to contribute to the development of MSR and FHR reactor technology in the area of reactor physics, nuclear – chemical engineering and material research. The project is solved by a consortium of Czech research institutions and industrial companies leaded by the Research Centre Řež.

Overview and status update on Molten Salt Reactor Technology development in the US

Reactor safety is evaluated to demonstrate that a plant’s operation does not present significant additional risk to the life and health of the public. Reactor safety evaluation historically focused on maintaining adequate containment of radionuclides during the maximum credible accident. However, as progressively larger light water-cooled reactors (LWRs) were developed in the 1960s, the increased potential for catastrophic accidents necessitated expanding the safety adequacy from the containment of radionuclides under all conditions to the prevention of accidents and the mitigation of their consequences. Either a deterministic or probabilistic pathway could be taken to demonstrate the safety adequacy for US molten salt reactors (MSRs). The deterministic pathway relies on adapting accepted minimum design criteria for LWRs to MSRs, whereas the probabilistic pathway relies on adequately modeling the risks of MSR accidents to discern what can occur, how likely it is to occur, and the consequences of its occurrence. MSR designs as envisioned have a readily apparent high degree of passive safety. Their combination of low pressure, low stored energy within containment, negative reactivity feedback, and effective passive decay heat removal substantially reduces the potential for cascading and escalating events. This MSR resiliency opens a third demonstration pathway that refocuses safety adequacy on containment of credible accidents, precluding the need for complete probability information. This approach would be especially useful for early prototype plants which lack sufficient performance data to take advantage of higher fidelity, data-driven risk modeling. This webinar will describe the current status and comparative advantages of the three alternative MSR safety adequacy demonstration pathways.

China's Multi-purpose SMR-ACP100 Design and Project Progress

The Nuclear Power Institute of China has developed many small and medium (SMR) reactors including those used for nuclear seawater desalination, nuclear district heating, and nuclear-powered commercial ships, multi-purpose small modular pressurized reactors. Developing nuclear energy is a measure of meeting national economic development and satisfying the need for energy conservation and reduction of emissions. Large size nuclear power units are less suitable to regional networks and non-electric fields, highlighting the importance of studying small and medium reactors. Small reactors are also in high demand for developing countries and the vast Midwest of China. Domestic cities and regions are also in urgent need of nuclear energy for heating, desalination, and district heating.

Phenix and Superphenix Feedback Experience

France energy situation is specific : no fossil energy available ( oil, coal, gas, etc..), a large fleet of PWR in operation providing about 80% of electricity , and a successful reprocessing activity providing each year about 10 tons of plutonium. In this situation, sodium fast breeder reactors would be very useful for the country, and have been developed with the Rapsodie, Phenix and Superphenix reactors. The feedback experience of these reactors has been analyzed and collected in two books “Phenix: the feedback experience” / EDP sciences 2012, and “Superphenix: Technical and Scientific achievements” / Springer 2016. This thematic analysis was performed on materials, fuel, neutronic, thermal hydraulic, components, water sodium reaction, sodium leaks, safety, and more generally on all the specific technical matters related to this type of reactor. The presentation gives, for each theme cited above, the main results obtained and the main conclusions or recommendations for the future of sodium fast breeder reactors.

Astrid - Lessons Learned

This presentation will first place the context of the choice of Sodium Fast Reactor in the French Nuclear Policy and its rationale for a closed fuel cycle. It will then present the position of the French Sodium Fast Reactor program in the context of Generation IV. The presentation will then focus on the ASTRID (Advanced Sodium Technological Reactor for Industrial Demonstration) project. The technical achievements, major innovation progress and management challenges will be presented. The ASTRID project description will highlight the major use of digital tools (numerical simulation, use of virtual reality, multiscale and multi-physics modeling, PLM: Product Lifecycle Management) used to perform efficiently such a complex project.

BN-600 and BN-800 Operating Experience

Russia occupies a leading position in the field of sodium-cooled fast reactors (SFR) in the world. The following SFRs have been constructed in the USSR and in the Russian Federation: BR-5/BR-10, BOR-60, BN-350, BN-600, BN-800. Total operation experience of the USSR/Russian SFRs exceeds 160 reactor-years, i.e. more than one-third of the total world SFR operational experience. The test reactor BOR-60 and industrial NPPs with the BN-600 and BN-800 reactors are in operation now. Currently the BN-1200 reactor design of a commercial SFR of the 4th generation is under development in Russia and a new MBIR multifunctional research fast reactor is under construction. A 38-year successful operation of the BN-600 reactor has demonstrated a feasibility and maturity of the SFR technology. The main purpose of the BN-800 reactor is to demonstrate closing the nuclear fuel cycle. The BN-1200 is aimed to achieve competitiveness with LWRs and to provide a high level of safety based mainly on passive safety systems and inherent SFR safety features. The main characteristics of the BN-600 and BN-800 reactors are described, and operational parameters achieved (burn-up, load factor, etc.) are discussed. The main stages of the BN-600 and BN-800 operations are indicated, and the operational experience gained is analyzed, including events with sodium leaks outside and leaks in steam generators. The presentation summarizes main lessons learned from the BN-600 and BN-800 operational experience and relevant recommendations for future SFRs.

Introducing New Plant Systems Design (PSD) Code

The nuclear sector is facing two major challenges. The first is to reduce cost of decommissioning old and building new nuclear power plants. In the UK, the Nuclear Sector Deal issued by the UK Government has called for 20% reduction in decommissioning costs and 30% reduction in the new build cost by 2030. The second challenge is to increase safety. The safety requirements have been toughened by the IAEA’s Design Extension Conditions that require plants to withstand multiple hazards and extreme hazards. The challenge is to reduce cost whilst increasing safety and that calls for a different design approach. The nuclear industry is responding to this challenge of reducing cost without compromising safety by taking part in the development of new Plant Systems Design (PSD) code that will change the way design and construction is done. This presentation will explain the new initiative that is being taken by committee of international experts under the aegis of ASME to develop the PSD code which is a technology neutral standard that provides a framework, including requirements and guidance, for design organisations. In traditional nuclear industry approach the design process goes through concept, preliminary design, detail design, construction, commissioning, and operation. The emphasis is mostly on component design not on system design and the whole design process is sequential. The PSD standard aims to bring in three main changes: (a) integrate process hazard analysis in the early stages of design; (b) incorporate and integrate existing systems engineering design processes, practices and tools with traditional architect engineering design processes, practices and tools; and (c) to integrate risk informed probabilistic design methodologies with traditional deterministic design. Main features and advantages of systems-based approach to integrate design and safety in the PSD code will be described. In brief, this presentation will show how the nuclear plants need to be designed in future to meet the ever-increasing cost and safety challenges.

Development of In-Service Inspection Rules for Sodium-Cooled Fast Reactors Using the System Based Code Concept

Effective and efficient in-service inspection (ISI) is crucial to maintain safety of nuclear power plants and to suppress operation costs which affect the power generation costs directly. Hence, ISI rules need to be developed rationally by considering relevant features of the reactor type and design of an individual nuclear power plant. Sodium-cooled fast reactors (SFRs) have several desirable features such as excellent compatibility between purity-controlled sodium and structural materials while traditional volumetric and surface tests are not as easily performed as in light water reactors due to the limited accessibility to components containing the sodium coolant. In this Webinar, development of ISI rules for SFRs using the System Based Code (SBC) concept is introduced. The SBC concept consists of three parts: 1) design to reliability target that must be met throughout the service life, 2) margin exchange among the various technical areas of concern such as design, inspection, fabrication, and fitness for service, and 3) expand technical options by the timely adoption of newly developed technologies that are not in current codes and standards. Such a flexible and consistent concept is suitable to develop ISI rules by taking account of individual features of SFR plants. A unique logic flow to determine ISI requirements by using reliability targets for components is discussed, and the procedure to derive the reliability targets from plant safety requirements and to evaluate corresponding structural reliability is presented.

In Service Inspection and Repair Developments for SFRs and Extension to Other Gen4 Systems

In Service Inspection is a major challenge to consider for future Generation IV Reactors safety. Therefore, a large focus of R&D work has been performed since 2010 in France for the Sodium Fast Reactor systems (SFR), mainly dedicated to the inspection of reactor block structures, primary components and circuits, and Power Conversion System main components (Heat Exchangers). In Service Inspection requirements have to be taken into account since the early pre-conceptual design phase, then consolidated through the basic design phase with more detailed specifications leading to increase the inspection tools ability for immersed sodium structures of SFRs, at about 200°C (shut down conditions). Inspection within the main vessel should be performed either with transducers immersed in sodium (with associated in sodium robotics) and with transducers located out of sodium medium. Nondestructive Examination, Telemetry and Imaging are qualified with experimental in-water and then in-sodium testing, using ultrasonic transducers. Experimental results are then compared to simulations using French CIVA software platform results. Repair was also part of this program, with laser system development. This webinar provides a technical overview of this ISI&R program that involves specific international collaborations done through GENIV mainly. Of course, it also benefits to other Gen4 systems.

Generation IV Coolants Quality Control

The quality of coolant in Fast Neutron Reactors must be controlled due to the potential impact of impurities on the structural material, on the dosimetry and subsequently on the operation. Liquid metals (sodium, lead-bismuth eutectic, pure lead) and gas (He) need to be purified in order to avoid deleterious effects and satisfy several safety requirements. Several purification systems and dedicated instrumentation have been developed for this purpose, taking into account the specific properties of each coolant.

Off-gas Xenon Detection and Management in Support of Molten Salt Reactors

Many molten salt reactor designs involve the circulation of a cover gas to remove evolved fission products and maintain an inert atmosphere. The cover gas leaving the reactor core is expected to contain noble and non-noble gases, aerosols, volatile species, tritium, and other radionuclides. To remove these radioactive gases, developing a robust off-gas system is necessary, along with novel sensors to monitor the gas stream and the treatment system performance. In recent years, a new class of functional materials known as metal–organic frameworks (MOFs) have been discovered using machine learning tools to identify key candidates for specific uses. In this study, a MOF was engineered for to capture of xenon, which is a major contributor to the off-gas source term. The engineered MOF column was tested with a residual gas analyzer (RGA) and laser-induced breakdown spectroscopy (LIBS) sensor for noble gas capture, separation, and monitoring. The RGA and LIBS sensor was used to monitor breakthrough tests with various Xe, Kr, and Ar mixtures to determine the Xe capacity, as well as the selectivity of the MOF. This study offers an initial demonstration of the feasibility of xenon separation (MOF) and monitoring using a LIBS sensor to aid in the development of new capture and sensor systems for molten salt reactors.

Sodium Integral Effect Test Loop for Safety Simulation and Assessment (STELLA)

The STELLA (Sodium Integral Effect Test Loop for Safety Simulation and Assessment) program was launched with the PGSFR (Prototype Gen-IV Sodium-cooled Fast Reactor) development plan in 2012; and, in the 1st stage, the design of two types of heat exchangers in passive DHRS (Decay Heat Removal System) was verified and validated. Specifically, the in-house design codes for the sodium-to-sodium heat exchanger, DHX, and the helical-type sodium-to-air heat exchanger, AHX, were evaluated by the experiment data with the large-scale facility, STELLA-1. The STELLA-1 is a separate effect test facility for component design V&V including the heat exchangers, the mechanical pump, and other miscellaneous components such as sodium valve, electro-magnetic pump, cold trap, plugging meter, etc. For the 2nd stage of the STELLA program, the STELLA-2 facility was designed to investigate the integral effect of safety systems including the comprehensive interaction among PHTS (Primary Heat Transfer System), IHTS (Intermediate Heat Transfer System) and DHRS. Especially, the focus of STELLA-2 is various combinations of interaction between passive and active DHRS with the PHTS and the long-term transient behavior needs to be observed to evaluate the overall safety aspect of PGSFR. The database of STELLA-2 is expected to be used for V&V of the safety analysis code as well as the demonstration of the overall safety system performance of PGSFR. Currently, the STELLA-2 construction has been completed, and it is actively operating to produce valuable data.

The Mechanisms Engineering Test Loop (METL) Facility at Argonne National Laboratory

The Mechanisms Engineering Test Loop (METL) facility, established in 2010, is an intermediate-scale liquid metal experimental facility that provides purified R-grade sodium to various experimental test vessels to evaluate components required to operate in a prototypical Sodium Fast Reactor (SFR) environment. The METL facility has the capability to test small to intermediate-scale components and systems in order to develop advanced liquid metal technologies. METL is a flexible facility as multiple configurations are available, enabling the hosting of experimental investigations from a range of disciplines with cross-cutting results. The presentation will provide an overview of METLs entire ecosystem; beginning with the flagship facility’s design methodology, construction and virtual tour. Supporting equipment such as the qualifying stations, test article removal/insertion flexible-cask and alkali metal passivation techniques/hardware will be covered as well. METLs demonstrated capabilities and infrastructure available to facilitate various research initiatives are summarized and the webinar will conclude with a highlight reel of current experimental programs as well as, what future technologies/demonstrations to expect from the METL team.

Metallic Fuels for Fast Reactors

This webinar will provide an overview of metallic fuels used in sodium-cooled fast reactors. Topics to be briefly surveyed will include: a history of metallic fuel development and use; benefits of metallic fuel technology for fuel reliability and safety; and current development directions in the areas of actinide transmutation and ultra-high burnup.

MOX Fuel for Advanced Reactors

Today, knowledge on MOX fuel behavior in fast neutron reactors comes mainly from feedback on SFRs that have operated in the past in Europe, USA, Japan and are still in service in Russia, India and China. The GENERATION-IV systems (SFR, GFR, LFR, FSMR...) with the associated fuel cycle strategy have been chosen to face the requirements of safety, non-proliferation, sustainability and waste minimization. This completion is possible thanks to the flexibility of fast neutron systems: they offer the possibility of using plutonium and uranium coming from spent fuel, making the best use of resources while reducing waste. Thus (U,Pu)O2 has proved to be the most ready candidate to achieve these performances in reactor and during the fuel cycle. Mox fuel is suitable for exemple for multirecycling, isogeneration, burning or breeding plutonium through adjustment of Pu concentration. Taking into account a wide range of fuel composition (Pu content: 20 to 45%), irradiation conditions and applying the safety criteria, we will present the state of the art on MOX fuel for GENIV systems with respect to knowledge and qualification. The knowledge on (U,Pu)O2 will be presented under the aspects of material properties and fuel behavior under irradiation with post irradiation examinations and modelling. The methodology of MOX qualification will be will be detailed with TRL (Technological Readiness Level) scale evaluation and the need to extend the qualification area in order to cover all design, composition and situations described above. The support of the international organizations (GIF, OECD/NEA, IAEA, EURATOM) to scientific and technological issues will be assessed.

Metal Fuel for Prototype Gen-IV Sodium Cooled Fast Reactor (PGSFR)

Fuel for the Prototype Generation-IV sodium-cooled fast reactor (PGSFR) is metallic fuel which has high thermal conductivity and excellent compatibility with sodium. U-Zr fuel is a driver for the initial core of PGSFR, and U-TRU–Zr fuel will replace U-Zr fuel later. U-TRU-Zr fuel will use TRU recovered through the pyro-electrochemical processing of spent PWR fuels, which contains highly radioactive minor actinides and chemically active rare earth elements as the carryover impurities. This webinar will explain the results and experiences of metal fuel development for PGSFR such as fuel design, fuel fabrication, development and fabrication of fuel components, and fuel verification tests. Fuel design is to determine the configuration, the dimensions and the materials of fuel components and fuel assembly to make sure fuel integrity and reactor safety during irradiation. Fuel rods were fabricated, and irradiation tests of them were performed in both the thermal and the fast research reactors. For higher performance, advanced cladding, FC92 and Cr barrier applied on cladding inner surface were developed, fabricated and tested in comparison with HT9. A fuel assembly was fabricated, and mechanical and hydraulic tests were performed.

TRISO Fuels

TRISO (TRi-structural ISOtropic) particle fuel has been developed for use in modular high temperature gas reactors (HTGR) designed to passively maintain core temperatures below fission product release thresholds under all licensing basis events and accident scenarios. This webinar will give an overview of the US DOE Advanced Gas Reactor (AGR) TRISO Fuel Qualification and Development Program’s activities focused on enhancing TRISO fuel performance by using uranium oxicarbide (UCO) fuel kernels and improving coated particle and compact fabrication methods for deployment in advanced HTGRs. Topics include fuel characterization and qualification methods, TRISO production scale fabrication process improvements, AGR TRISO irradiation experiments, post-irradiation examination and safety heating test results, and fuel performance modeling efforts. Current US TRISO fuel reactor vendor efforts, and the first TRISO topical report submitted to the NRC will be presented.

Overview of Nuclear Graphite R&D in Support of Advanced Reactors

As arguably the very first nuclear reactor core material, graphite has been utilized in a variety of nuclear applications since Enrico Fermi first stacked up bricks of graphite in a university squash court. But why? Graphite is not the first material that comes to mind when considering the extreme environment anticipated within a nuclear core. Materials with high strength, toughness, hermeticity, and hardness are traditional material choices for this demanding application. Graphite exhibits only moderate, or even low, values for these material properties. This presentation will address these issues and attempt to demonstrate that graphite is nearly the perfect material choice for these (Very) High Temperature Reactor designs. The latest information on graphite's unique crystal structure and bulk microstructure which provide the desired properties, the (baffling) irradiation behavior, the expected response to anticipated degradation, and how the nuclear graphite community is establishing the operational safety envelop of the core components within these new advanced reactor designs will be discussed. We'll finish up with a short demonstration of why nuclear graphite cannot burn (No, Chernobyl graphite fires did not happen).

General Considerations on Thorium As Nuclear Fuel

This webinar will present an overview of the basic concepts behind the historical interest on the use of thorium as a nuclear fuel. It will aim at reviewing thorium’s real potential and the many challenges it is facing before it can be part of the solution to the world’s energy problems. It is aimed at giving some of the scientific elements to a general audience in order to “demystify the thorium question” which has regained some prominence in recent years when talking about future nuclear concepts.

Neutrino and Gen IV Reactor Systems

Neutrinos were first conceived of in a desperate ploy to save the conservation of energy in nuclear beta decay. From a nuclear engineering perspective, they are an ever present but ultimately inconsequential by-products of nuclear fission. But to particle physicists, nuclear reactors are an essential and extraordinarily bright source of neutrinos, which are responsible for some of the most important advancements of the last 70 years, from the discovery of the neutrino at the Savannah River’s P reactor in 1956, through current times when reactors are used as the source for studies of neutrino fundamental properties. Advances in neutrino detector technology and a recent finding that neutrinos can be used to track the production of plutonium in the core have made it possible for particle physicists to contemplate giving back to the nuclear industry. As a new type of non-invasive instrumentation, neutrino detectors may, for example, have a role to play in non-proliferation safeguard regimes, particularly for advanced reactors. Ultimately, the potential of neutrino applications can only be properly assessed in a collaborative effort between particle physicists and nuclear engineers. This webinar will review the discoveries and other major advances in neutrino physics that have been enabled by nuclear reactors, and explore the ways that neutrino detectors may be used to monitor reactors or as reactor instrumentation.

Lead Containing Pb-208: New Reflector for Improving Safety of Fast Neutron Reactors

This webinar considers improvement of fast reactor safety through slowing-down power runaway. The idea is surrounding the core by the neutron reflector made of lead-208, a material of heavy atomic weight and extremely low neutron absorption. The power runaways can be slowed down because of a long way for leakage neutrons to come back from distant layers of neutron reflector to the core. It is demonstrated that mean prompt neutron lifetime can be elongated roughly by three orders of magnitude with appropriate slowing-down the reactor power runaway.

Multiphysics Depletion & Chemical Analyses of Molten Salt Reactors

Molten Salt Reactors (MSRs) are an innovative Generation-IV reactor concept which use nuclear fuel dissolved in a high temperature liquid salt and allow for enhanced safety and economic performance. The liquid fuel feature also entails several multiphysics effects that can complicate reactor design. One primary effect, coined here as depletion-driven thermochemistry, is the changing chemical redox potential of the fuel salt due to chemical composition changes driven by depletion. As the fuel is consumed and fission products are formed, the redox potential of the fuel salt shifts toward a more oxidizing state. Without active control, the changing chemistry can have multiple effects on the multiphysics behavior of the reactor that are important for both steady state operation and for accident scenario transients. A new multiphysics framework capability developed at Idaho National Laboratory can now simulate these coupled processes occurring in MSR systems during depletion including neutron transport, nuclide generation, thermal hydraulics, thermochemical equilibrium, chemical species transport, corrosion, and active chemistry control. The application of this work includes modeling source term, decay heat removal, reactivity transients, corrosion, chemistry control, and safeguards analyses. Future work focuses on validation efforts by defining a thermochemical benchmark against the Molten Salt Reactor Experiment (MSRE) and future digital twins of near-term experiments.

Date: Wednesday, April 17, 2024, 8:30 am EST (UTC-4)
Register now at https://attendee.gotowebinar.com/register/3263652212604349019

Thermal Hydraulics in Liquid Metal Fast Reactors

Thermal-hydraulics play a determining role in the design, operation and safety of liquid-metal reactors (LMRs) cooled by sodium, lead or lead-bismuth eutectic. The strong heat transfer performance and high boiling point of liquid metal enable the use of high working temperatures without pressurization. Because no pressure vessel is needed, most reactor designs then adopt a "pool-type" primary circuit, which minimizes the potential consequences of a primary leak and provides a large reserve of thermal inertia in accidental scenarios. While these common design characteristics of LMRs have direct advantages, they are also the source of complex thermal-hydraulic phenomena with potential high impact: strong temperature gradients must be controlled to avoid thermal fatigue on reactor structures, decay heat removal in pool-type designs depends on complex natural convection patterns. In this way, many key aspects of the justification of LMRs depend on understanding and simulating complex thermal-hydraulic phenomena. This webinar provides an overview of these phenomena and the current state-of-the-art for simulating them.

Passive Decay Heat Removal System

A major design goal for Generation IV nuclear energy systems is to reduce or eliminate the likelihood and/or extent of reactor core damage incurred during an off-normal operating event, thereby eliminating the need for offsite emergency response. One approach for achieving this objective is to develop inherently safe reactor designs that can passively dissipate decay heat to the environment without relying on operator action during an event of this type. Historically, this approach has been taken for both sodium- and gas-cooled Generation IV reactor types by providing Reactor Cavity Cooling Systems (RCCS) that are designed to passively dissipate decay heat to the environment by natural convection while maintain fuel temperature below the threshold for onset of core damage. This presentation will begin by providing a high level overview of RCCS systems that have been developed for advanced reactor designs over the years. This will be followed by a summary of large scale integral effect tests that are currently underway at Argonne to provide licensing-quality data for two of these systems; i.e., air- and water-cooled RCCS concepts.

Energy Conversion

The rotary motion, high pressure steam engine was patented by James Watt in 1781. The evolution of steam engines and high pressure boiler technology led directly to the development of the steam turbine coupled to an electrical generator by Charles Parsons in 1884. Since then, over the last 133 years, the world has been using steam turbines to convert heat into electricity in almost all of the world’s thermal power stations and in all of the world’s nuclear power stations. Specifically for the latter, steam turbines and the Rankine thermodynamic cycle in which they operate offer high efficiency for moderate steam temperatures, temperatures typical of first, second and third generation nuclear reactors. Generation IV reactors offer the potential to move away from the steam Rankine cycle to systems such as helium (or nitrogen) Brayton or supercritical CO2 gas turbine cycles to exploit the higher temperatures that some of the systems generate, to offer plant simplification and potentially higher conversion efficiencies. Nonsteam cycles offer other advantages, particularly in connection with the sodium cooled fast reactor, such that the risk of sodium water reactions is massively reduced. Within this webcast, the basic thermodynamics and performance limits of energy conversion systems will be explained and each of the technological options proposed for the energy conversion systems of Generation IV reactors will be presented.

Geometry Design and Transient Simulation of a Heat Pipe Micro Reactor

In recent years, micro-reactor concepts have attracted increasing attention in the nuclear industry due to the market demand for flexible, reliable, and sustainable power and heat on-site for industrial or federal installations or remote communities. Micro-reactor AGile Non-nuclear Experimental Test-bed (MAGNET) is at Idaho National Laboratory (INL) with an initial focus on the thermal and structural performance of heat pipe cooled micro-reactors. System Analysis Module (SAM) code is a multi-dimensional modern software tool provided by Argonne National Laboratory (ANL). This webinar will review the designs of heat-pipe micro reactors, steady-state and transient simulation of MAGNET by SAM, to explore the performance limit by heat transfer and temperature distribution.

Development of Multiple-Particle Positron Emission Particle Tracking for Flow Measurement

Flows in opaque systems can present a significant challenge to experimental investigators. Understanding flow phenomena in reactor components often relies on the use of simulation, as well as experiments using surrogate materials and fluids to allow optical access. Positron emission particle tracking (PEPT) is a radiotracer-based technique that uses the same technology as the medical imaging platform PET (positron emission tomography). As such, PEPT can be used to directly study flows in opaque systems. The research focus has been on the development and deployment of new PEPT reconstruction algorithms that allow the simultaneous tracking of multiple tracers, increasing data collection efficiency and enabling new measurements. Herein Dr. Wiggins will discuss the basics of PEPT, as well as its utility for measurements in pipes, heat exchangers, and pebble beds, among other systems. The data gleaned from such experiments can be used for both fundamental understanding of flow phenomena and validation of the computational fluid dynamics models being used for next generation reactor design.

Materials Challenges for Generation IV Reactors

The Generation IV reactors offer significant advantages over typical light water reactors including increased power conversion efficiency, passive safety features and in some cases process heat for other applications (e.g. hydrogen production). These families of reactors include 3 fast reactors [sodium fast reactor (SFR), lead fast reactor (LFR) and gas-cooled fast reactor (GFR)], one thermal reactor [very high temperature reactor (VHTR)] and two fast or thermal reactors [supercritical water reactor (SCWR) and molten salt reactor (MSR)]. The extreme environments in these families of reactors create significant challenges to materials ranging from high doses from a fast neutron flux (SFR, LFR, GFR, SCWR and MSR), more corrosive environments from molten salt (MSR) or lead coolants (LFR) and high temperatures in the helium-cooled reactor concepts (e.g. GFR and VHTR). This presentation will discuss the materials challenges in Generation IV reactor concepts and summarize radiation effects in reactor metals proposed for these concepts over prototypic irradiation conditions.

Graphite-Molten Salt Interactions

The new High Temperature Reactor (HTR) designs being considered for future Gen IV nuclear reactor deployment include designs utilizing molten salt as the primary coolant. These molten-salt cooled, graphite core designs pose new material compatibility challenges that are not considered within the gas-cooled HTR designs that have been previously built and operated. In MSRs, graphite is not only exposed to fast neutron irradiation but also in continuous contact with the coolant molten salt, the fuel salt, or both, depending on the design. The continuous operation in contact with the molten salts is expected to affect graphite’s local composition and microstructure, which in turn impacts the mechanical, thermal, and irradiation-resistance properties of the graphite. These issues are currently under investigation within the DOE Advanced Reactor Technologies (ART) graphite program and will be presented at this seminar.

Corrosion and Cracking of Supercritical Water Reactor (SCWR) Materials

Steam (with increased temperature and pressure) is always the challenge of power technology for high-efficiency production of electricity. Supercritical water fossil fired plants have been the mainstream for over 5 decays. A supercritical water-cooled nuclear reactor (SCWR) concept has been selected by the Generation IV International Forum (GIF), but scientists and designers are facing unprecedented difficulties in solving the problems in material, neutronics and thermo-hydraulics. The high temperature and pressure operation condition of a SCWR core is tough and challenging to the materials for the nuclear fuel cladding due to its thickness limitation, strict corrosion rate requirement and high reliability expectations. This webinar will present the state-of-art development of materials for SCWR fuel cladding, the mechanisms of general corrosion and stress corrosion cracking, the major factors affecting their corrosion and cracking performances, and the proposed testing procedures to obtain consistent and reproducible corrosion and cracking data under supercritical water conditions.

Cement Matrix for Nuclear Waste (2018 GIF EPiC contest)

This webinar discusses the formulation of an alternative cement matrix for solidification/stabilization of nuclear waste. The presentation provides an overview of the multiple complexities of waste management, and the many challenges that arise from it. Topics include a presentation of the French nuclear waste management methods, specific examples on solidification/stabilization of nuclear waste, the physico-chemical aspects of the interactions between the containment matrix and the waste, and the miniaturization of samples for the development of new matrices allowing human radiation protection. The webinar also highlights current experimental research focused on Portland cement and a magnesium potassium phosphate cement matrix. The latter is a promising cement for the stabilization/solidification of heavy metals. Other potential cementitious matrices will also be discussed.

Interactions between Sodium and Fission Products in Case of a Severe Accident in a Sodium-cooled Fast Reactor (2018 GIF EPiC contest)

An overview of severe accident scenarios in Sodium-cooled Fast Reactors will be presented, focusing on the thermochemistry aspects and how the CALPHAD method could be used to enhance the prediction of the different phases that could form depending on the conditions of the system. CALPHAD, which stands for CALculation of PHAse Diagram, is a semi-empirical method that enables to develop a thermodynamic model based on the Gibbs free energy of the gas, liquid and solid phases as a function of temperature, pressure and composition of the system. Experimental measurements of the thermodynamic properties of some fission product compounds formed in the Joint Oxide Gain after interaction with sodium will be presented. These data will be used as input for the thermodynamic modeling.

Security Study of Sodium-Gas Heat Exchangers in Frame of Sodium-cooled Fast Reactors (2018 GIF EPiC contest)

This webinar provides an overview of a Sodium Fast Reactor system and presents an accident scenario in Compact plates Sodium-Gas heat Exchangers (ECSG) of SFR. The overpressure (180 bar in the nitrogen loop while 5 bar in the sodium loop) could result in nitrogen leaking into the liquid sodium. The present work focuses on the analysis of the predominant physical phenomena in the jet (the viscous diffusion, the momentum exchange between the two fluids) and supersonic gas jet, the development of the compressible multiphase flow model (Baer-Nunziato model) and its numerical schemes. In addition, the model is implemented using the numerical tool CANOP that enables researchers to generate the Adaptive Mesh Refinement and to calculate in parallel.

Development of an Austenitic/Martensitic Gradient Steel Connection by Additive Manufacturing (2021 GIF Pitch contest)

Metal additive manufacturing is already used to build parts for aeronautic or medical applications. The power generation industry is just beginning to use these new technologies to replace or repair parts. Dr Villaret’s work aims to make this technology available for nuclear industry and especially for GEN IV reactors where technological innovation will find all their meaning. Flore Villaret's PhD work focuses on an application of additive manufacturing to obtain bimetallic junctions. Such junctions between two different metals are very common in nuclear reactors, but they are often difficult to realize. The example studied is a junction between an austenitic stainless steel 316L and a martensitic steel Fe-9Cr-1Mo. Such connections are typically found in sodium fast reactors assemblies between wrapper tubes and assembly spikes. The objective is to understand the metallurgical features related to the assembly of these two steels and to evaluate the possibilities of using powder metallurgy and additive manufacturing to produce austenitic/martensitic steel transitions. With this work, the ability of additive manufacturing and powder metallurgy to overcome this challenge, by producing chemically graded parts, is demonstrated. Beyond the replacement of already existing parts, the possibilities opened by the additive manufacturing of gradient materials are very varied. It also constitutes a new tool at the disposal of design offices to imagine new designs and applications. This study demonstrates the potential presented by these gradient materials for future innovations.

Scale Effects Analysis on the Thermal Hydraulic Behavior of Impinging Jets in Sodium Fast Reactors (2021 GIF Pitch contest)

In the industry and especially in the nuclear field, simulation software plays an important part on the study of new prototypes, as building scale 1 prototype is both complex and expensive. To assess new design options and validate associated simulation software, most experiments are performed on small-scale mock-ups. However, distortions of a studied phenomenon can occur from a small-scale model to a prototype: this is known as scale effects. Without understanding of these effects, results from models cannot be transposed to the prototype scale and numerical results cannot be validate either. In the French SFR project, an important problematic is the raise of the core jets in the upper plenum for specific operating conditions. When these jets raise, the flow pattern in the vessel is modified and leads to thermal fluctuations reducing the lifetime of the reactor’s components because of thermal fatigue. While this raise can be observed on small-scale mock-up, it is important to be able to characterize both this phenomenon and its transposition. For this purpose, Benjamin Jourdy works on a methodology based on bibliography synthesis and experimental results, leading to the conception of a new mock-up dedicated to this study.

Development of Nanosized Carbide Dispersed Advanced Radiation Resistant Austenitic Stainless Steels (ARES) for Generation IV Systems (2021 GIF Pitch contest)

Austenitic stainless steels are widely used as in-core materials in current nuclear systems, due to the combination of good corrosion resistance, ductility, and mechanical properties. However, in the next generation nuclear systems to be operated at higher temperature and higher level of neutron irradiation, the use of austenitic stainless steels is largely avoided due to, mostly, poor void swelling resistance. In this regard, our research group developed an austenitic SS with uniformly distributed nanosized NbC precipitates, named as ARES-6P, by controlling chemical composition and thermo-mechanical processing. To demonstrate the radiation resistance, heavy ion irradiation was performed at high temperatures to the damage level of ~200 displacement per atom (dpa). The measured void swelling of ARES-6P was 2‒3%, which was considerably less compared to commercial 316 stainless steel and comparable to ferritic martensitic steels. In addition, increment of hardness measured by nano-indentation was much smaller for ARES-6P compared to 316 stainless steel. Though some nanosized NbC precipitates were dissociated under relatively high dose rate (~5.0 × 10-4 dpa s-1), sufficient number of NbC precipitates remained to act as sink sites for the point defects, resulting in such superior radiation resistance. Both significantly less void swelling and less irradiation hardening indicate the superior irradiation resistance of ARES-6P for the application of next generation nuclear systems.

A Gas Cherenkov Muon Spectrometer for Nuclear Security Applications (2021 ANS pitch contest)

Cosmic ray muons have been considered as a non-conventional radiation probe for imaging large and dense objects because they are high-energetic and penetrative. To maximize the utilizability of cosmic ray muons in engineering applications, two important quantities, trajectory and momentum, must be known. The muon trajectories are easily reconstructed using two-fold detector arrays with a high spatial resolution. However, precise measurement of muon momentum is difficult to be achieved without deploying large and expensive spectrometers. In our research, we develop a novel muon spectrometer using multi-layer pressurized gas Cherenkov radiators and demonstrate its functionality for muon momentum measurement using high fidelity Geant4 simulations. In addition, we develop a new imaging algorithm for muon tomography, or mPoCA, by integrating muon momentum information into the original PoCA algorithm. The results of reconstructed images of various materials using both PoCA and mPoCA algorithms will be presented and compared. Not only the image resolution is significantly improved but also we were able to visually differentiate shielding material (Pb) from special nuclear materials otherwise impossible to see with the original PoCA.

Characterization of U-233 for Thorium Fuel Cycle Safeguards (2023 ANS pitch contest)

Renewed international interest in thorium-fueled advanced reactors has challenged the safeguards community to address future proliferation concerns. Thorium-based technology presents many benefits but does not eliminate the proliferation risks associated with producing and processing fissile material. A byproduct of thorium-fueled reactors is uranium-233(U-233), which is classified as a direct-use material. As a result, the development of new or improved methods to characterize and measure materials containing U-233 must mirror the pace of development of reactors and facilities that produce such material. Research is underway to assess, develop, and test approaches for safeguarding nuclear materials within the thorium fuel cycle. Neutron signatures from the nondestructive assay(NDA) of materials containing U-233 and U-235 are being quantified to inform the potential characterization of these materials. Using a traditional neutron coincidence counter and a series of well-documented U-233 oxide samples, initial measurements have been made to assess the feasibility of U-233 characterization and discrimination from other uranium isotopes, primarily U-235, using a combination of measurement techniques and analysis methods. Data acquisition is performed in list mode, allowing for a variety of analyses to be performed on the raw data that is not available using traditional shift register technology. Measurements were performed with an Active Well Coincidence Counter(AWCC) in passive and active configurations to quantify the strength of signals and validate simulations to demonstrate feasibility of a novel characterization technique.

Analysis of the Reactivity Loss of the PHENIX Core Cycles for the Experimental Validation of the DARWIN-FR Code Package

The development of future reactors relies on design methodologies using scientific calculation tools (SCT), validated and relevant for quantifying the uncertainty for each neutronic quantity of interest. In the validation process, SCT results are compared with experimental data; however, such reference set is limited for fourth-generation reactors. Sodium-cooled fast reactor technology offers substantial overall feedback, making it an essential asset for assessing current SCTs regarding the development of future reactors. In particular, fuel depletion requires the evolution of isotopic composition within a reactor to be mastered for design/operation trade-offs (irradiation time, mass balance of fissile materials, feedback coefficients, among others). As the local isotopic compositions are not accessible “in real time”, the assessment of SCTs for fuel depletion adopts an indirect approach. Analysis of fuel depletion involves studying its consequence: reactivity swing with irradiation. Substantial experimental data on this phenomenon have been obtained in the PHENIX reactor during the whole operation time (40 years). The current analysis involves an exhaustive reinterpretation of the existing experimental data and depletion calculations using the modern deterministic calculation tools. This comprehensive study of the SFR reactivity loss provides valuable insights, establishing links between experimental data and depletion calculations.

Geospatial Analytics for Energy and Resilience Analysis

Energy systems are often compared on the basis of their economics. However, these can vary with location as can access to resources, labor, transportation networks and the receptiveness of host communities. Natural hazards and climate impacts can also vary dramatically from one location to another. This webinar provides an overview of energy related geospatial analytics at the Colorado School of Mines that grew Nuclear Energy University Programs (NEUP) funding.

International Knowledge Management and Preservation of SFRs

The international SFR panel will discuss the lessons learned on the design, construction and operation of SFRs (Superphénix, Monju, and FFTF) and how the transfer of knowledge is passed on in an international context to companies planning on building SFRs ranging in power from 300 to 1,200MWe.