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DOE OSTI2015

Preliminary Framework for Human-Automation Collaboration

The Department of Energy’s Advanced Reactor Technologies Program sponsors research, development and deployment activities through its Next Generation Nuclear Plant, Advanced Reactor Concepts, and Advanced Small Modular Reactor (aSMR) Programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Human Automation Collaboration (HAC) Research Project is located under the aSMR Program, which identifies developing advanced instrumentation and controls and human-machine interfaces as one of four key research areas. It is expected that the new nuclear power plant designs will employ technology significantly more advanced than the analog systems in the existing reactor fleet as well as utilizing automation to a greater extent. Moving towards more advanced technology and more automation does not necessary imply more efficient and safer operation of the plant. Instead, a number of concerns about how these technologies will affect human performance and the overall safety of the plant need to be addressed. More specifically, it is important to investigate how the operator and the automation work as a team to ensure effective and safe plant operation, also known as the human-automation collaboration (HAC). The focus of the HAC research is to understand how various characteristics of automation (such as its reliability, processes, and modes) effect an operator’s use and awareness of plant conditions. In other words, the research team investigates how to best design the collaboration between the operators and the automated systems in a manner that has the greatest positive impact on overall plant performance and reliability. This report addresses the Department of Energy milestone M4AT-15IN2302054, Complete Preliminary Framework for Human-Automation Collaboration, by discussing the two phased development of a preliminary HAC framework. The framework developed in the first phase was used as the basis for selecting topics to be investigated in more detail. The results and insights gained from the in-depth studies conducted during the second phase were used to revise the framework. This report describes the basis for the framework developed in phase 1, the changes made to the framework in phase 2, and the basis for the changes. Additional research needs are identified and presented in the last section of the report.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Generation and Use of Thermal Energy in the U.S. Industrial Sector and Opportunities to Reduce its Carbon Emissions

The industrial sector was the third-largest source of direct U.S. greenhouse gas (GHG) emissions in 2014 behind electricity generation and transportation and accounted for roughly 20% of total emissions (EPA 2016). The Energy Information Administration (EIA) projects that total U.S. energy consumption will grow to about 108 exajoules (1 EJ = 10 18 J) or 102 quads (1 quad = 10 15 British thermal units) in 2025, with nearly all of the growth coming from the industrial sector (DOE 2015b). Energy consumption in the industrial sector is forecast to increase to 39.5 EJ (37.4 quads)—a 22% increase, exceeding 36% of total energy consumption in the United States. Therefore, it is imperative that industrial GHG emissions be considered in any strategy intent on achieving deep decarbonization of the energy sector as a whole. It is important to note that unlike the transportation sector and electrical grid, energy use by industry often involves direct conversion of primary energy sources to thermal and electrical energy at the point of consumption. About 52% of U.S. industrial direct GHG emissions are the result of fuel combustion (EPA 2016) to produce hot gases and steam for process heating, process reactions, and process evaporation, concentration, and drying. The heterogeneity and variations in scale of U.S. industry and the complexity of modern industrial firms’ global supply chains are among the sector’s unique challenges to minimizing its GHG emissions. A combination of varied strategies—such as energy efficiency, material efficiency, and switching to low-carbon fuels—can help reduce absolute industrial GHG emissions. This report provides a complement to process-efficiency improvement to consider how clean energy delivery and use by industry could reduce GHG emissions. Specifically, it considers the possibility of replacing fossil-fuel combustion in industry with nuclear (specifically small modular reactors [SMRs]), solar thermal (referred to herein as solar industrial process heat [SIPH]), and geothermal energy sources. The possibility of applying electrical heating and greater use of hydrogen is also considered, although these opportunities are not discussed in as much detail.

08 HYDROGEN↗
DOE OSTI2016

Generation and Use of Thermal Energy in the U.S. Industrial Sector and Opportunities to Reduce its Carbon Emissions

This report quantifies greenhouse gas (GHG) emissions from the industrial sector and identifies opportunities for non-GHG-emitting thermal energy sources to replace the most significant GHG-emitting U.S. industries based on targeted, process-level analysis of industrial heat requirements. The intent is to provide a basis for projecting opportunities for clean energy use. This provides a prospectus for small modular nuclear reactors (including nuclear-renewable hybrid energy systems), solar industrial process heat, and geothermal energy. This report provides a complement to analysis of process-efficiency improvement by considering how clean energy delivery and use by industry could reduce GHG emissions.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2017

Corrosion of Structural Materials for Advanced Supercritical Carbon- Dioxide Brayton Cycle

The supercritical carbon-dioxide (referred to as SC-CO 2 hereon) Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts, including the sodium fast reactor (SFR), fluoride saltcooled high temperature reactor (FHR), and high temperature gas reactor (HTGR), and several types of small modular reactors (SMR). The SC-CO 2 direct cycle gas fast reactor has also been recently proposed. The SC-CO 2 Brayton cycle (discussed in Chapter 1) provides higher efficiencies compared to the Rankine steam cycle due to less compression work stemming from higher SC-CO 2 densities, and allows for smaller components size, fewer components, and simpler cycle layout. For example, in the case of a SFR using a SC-CO 2 Brayton cycle instead of a steam cycle would also eliminate the possibility of sodium-water interactions. The SC-CO 2 cycle has a higher efficiency than the helium Brayton cycle, with the additional advantage of being able to operate at lower temperatures and higher pressures. In general, the SC-CO 2 Brayton cycle is well-suited for any type of nuclear reactor (including SMR) with core outlet temperature above ~ 500°C in either direct or indirect versions. In all the above applications, materials corrosion in high temperature SC-CO 2 is an important consideration, given their expected lifetimes of 20 years or longer. Our discussions with National Laboratories and private industry early on in this project indicated materials corrosion to be one of the significant gaps in the implementation of SC-CO 2 Brayton cycle. Corrosion can lead to a loss of effective load-bearing wall thickness of a component and can potentially lead to the generation of oxide particulate debris which can lead to three-body wear in turbomachinery components. Another environmental degradation effect that is rather unique to CO 2 environment is the possibility for simultaneous occurrence of carburization during oxidation of the material. Carburization can potentially lead to embrittlement of structural alloys in SC-CO 2 Brayton cycle. An important consideration in regards to corrosion is that the temperatures can vary widely across the various sections of the SC-CO 2 Brayton cycle, from room temperature to 750°C, with even higher temperatures being desirable for higher efficiencies. Thus the extent of corrosion and corrosion mechanisms in various components and SC-CO 2 Brayton cycle will be different, requiring a judicious selection of materials for different sections of the cycle. The goal of this project was to address materials corrosion-related challenges, identify appropriate materials, and advance the body of scientific knowledge in the area of high temperature SC-CO 2 corrosion. The focus was on corrosion of materials in SC-CO 2 environment in the temperature range of 450°C to 750°C at a pressure of 2900 psi for exposure duration for up to 1000 hours. The Table below lists the materials tested in the project. The materials were selected based on their high temperature strength, their code certification status, commercial availabilities, and their prior or current usage in the nuclear reactor industry. Additionally, pure Fe, Fe-12%Cr, and Ni-22%Cr were investigated as simple model materials to more clearly understand corrosion mechanisms. This first phase of the project involved testing in research grade SC-CO 2 (99.999% purity). Specially designed autoclaves with high fidelity temperature, pressure, and flow control capabilities were built or modified for this project.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Coupled Calculations in Helical Steam Generator: Validation on Legacy Data

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past five years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. In this regime one-way coupling is judged sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators at the low flows expected in integral PWRs. The legacy data is obtained from two datasets developed at Argonne and B&W.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Development of Novel Functionally Graded Transition Joints for Improving the Creep Strength of Dissimilar Metal Welds in Nuclear Applications

Steam generators of the small modular Very High Temperature Reactor (VHTR) will require use of Alloy 800 and 2.25Cr-1Mo steel in different regions depending on temperature and corrosion resistance requirements. As a result, Dissimilar Metal Welds (DMWs) between these two materials are critical for design, development, and manufacturing of the VHTR. DMWs are also used in many other applications in the power generation and petrochemical industries. Experience has demonstrated that failures of such DMWs can occur prematurely well below the expected creep life of either base metal, and these failures have significant consequences in terms of plant reliability, safety, and economic impact. These DMWs failures are caused by the abrupt change in composition that occurs at the weld interface between the two alloys. In this research, the concept of developing and optimizing Graded Transition Joints (GTJ) was developed through an integrated modeling and experimental approach. With GTJs, the composition is gradually graded between the two alloys, thus eliminating the sharp change in composition that causes failure. Microstructural modeling was first conducted to identify the optimum grade length required for improved resistance to carbon diffusion that is largely responsible for premature failure. Next, GTJs were fabricated and characterized in detail to understand how the microstructure evolves during fabrication and long term aging. Finally, the creep behavior of the GTJs was investigated and compared to that of conventional DMWs to demonstrate the superior creep strength of the GTJs. The results of this research provide significant insight into design, fabrication, microstructural evolution, and resultant performance of GTJs that can be used for reducing or eliminating DMW failures in a number of high temperature applications. The results also demonstrate that the GTJs can be readily fabricated with either conventional or additive manufacturing processes, and detailed guidelines are provided on the length of grades needed for optimal performance. The results of this research benefits the general public by improving the reliability of power generation and petrochemical plants that rely on DMWs for safe and reliable operation to deliver power and other products.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Enhanced Accident-Tolerant Fuel Performance and Reliability for Aggressive iPWR/SMR Operation

This project will develop a process that evaluates the whole-core operational and safety performance of a reactor-and-fuel combination by integrating core-wide three-dimensional neutronics with Moose/Bison and Peregrine multi-physics fuel performance assessments using explicit time and space dependent fuel rod operational power histories generated in the whole-core analysis. Demonstration of the process will target large contemporary PWRs with traditional UO2 fuel and Zircaloy cladding to establish a baseline analysis, while the innovative application of this project will extend the developed process to evaluate operational and safety performance, via reactor physics and fuel performance analyses, of key emerging/leading concepts of enhanced Accident Tolerant Fuels (ATFs) within increasingly aggressive operational scenarios such as loadfollow maneuvers in PWRs and long-cycle rodded and unborated operation in small modular reactors (SMRs). The ATFs to be considered include standard UO2 fuel pellets cladded with advanced iron-based alloys (FeCrAl) as well as particle-based Fully Ceramic Microencapsulated (FCM) fuels with FeCrAl or Zircaloy cladding, but the process being developed and demonstrated could ultimately be applied to any candidate ATF concept in a light water reactor. In fact, given that ATF R&D priorities are in a dynamic state, the scope related to ATFs will be targeted for the second half of this three-year project so to provide some opportunity to strengthen the programmatic alignment of this project with the DOE Fuels Campaign, by identifying the most promising and relevant ATFs at that time for further evaluation, while the early part of the project will focus upon fine-tuning the process herein developed and upon generating the baseline analyses against which a larger scope of ATF concepts can be compared.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Enabling Situation Awareness Under High Levels of Automation: Results From An Experimental Study

The Department of Energy’s Advanced Reactor Technologies (ART) program sponsors research, development and deployment activities through its Next Generation Nuclear Plant, Advanced Reactor Concepts, and Advanced Small Modular Reactor (aSMR) programs to promote safety, technical, economical, and environmental advancements of innovative Generation IV nuclear energy technologies. The Human Automation Collaboration (HAC) research project is located under the aSMR program, which identifies developing advanced instrumentation and controls and human-machine interfaces as one of four key research areas. New nuclear plants are expected to employ significantly more advanced technology than the systems in the current reactor fleet as well as utilize automation to a greater extent. However, advancing technology and increasing automation does not necessarily imply improved efficiency and safety of the plant. Instead, a number of concerns about how these technologies will affect human performance and the overall safety of the plant need to be addressed. More specifically, it is important to investigate how the operator and the automation work as a team to ensure effective and safe plant operation, also known as the human-automation collaboration (HAC). The HAC project at INL, investigates collaboration methods between a system’s automation and human operators and how that collaboration impacts the performance and reliability of overall system performance. Some well documented human performance issues associated with typical human automation collaboration methods include reduced operator situation awareness (SA), complacency, and over reliance on the system automation (Endsley, 1995, 1996, 1997; Endsley & Kaber, 1999; Kaber & Endsley, 1997, 2004; Parasuraman et al., 2000; Sheridan, 2002; Wickens & Hollands, 2000; Wright & Kaber, 2005). The HAC research effort addresses how to best design the collaboration between the operators and the automated systems in a manner that has the greatest positive impact on overall plant performance and reliability, hence addressing many of the documented issues. The ultimate goal of the HAC research effort is to develop design guidance that supports optimal interaction between humans and automated systems.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Advanced SMRs Using S-CO2 Power Conversion with Dry Cooling

This report concludes the LDRD entitled "Advanced SMRs using S-CO2 Power Conversion with Dry Cooling." The goal of this project was to demonstrate the feasibility of using sCO 2 as the working fluid in a dry-cooled natural circulation loop. The reason for doing this is to demonstrate that such a loop could be utilized in small modular reactors (SMRs) as a method for (a) passively removing decay heat from the reactor and (b) cooling the reactor without using the copious amounts of water needed for wet-cooled reactors. The dry-cooling aspect of this work is possible due to the working conditions of a sCO 2 Brayton power conversion cycle. The loop was designed, built, and operated over the three-year life of the LDRD project. This report outlines some of the key accomplishments and some of the future work that we are continuing to try to complete in the future. A more comprehensive report will be completed and submitted in the Fall of 2015.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

United States Nuclear Manufacturing Infrastructure Assessment

This report provides a review and assessment of the existing U.S. infrastructure capabilities to manufacture on a commercial scale, the major components and unique materials required by the various U.S. designs for advanced nuclear reactors and small modular reactors (SMRs). The approach was to compare the needs of SMRs and advanced reactors against the capability of U.S. manufacturers to ascertain gaps that could inhibit U.S. industrial support for deployment of SMRs or advanced reactors. This report focused on the reactor plant designs being developed by NuScale Power (the NuScale Power Module), TerraPower (both the Traveling Wave Reactor and the Molten Chloride Fast Reactor to a limited extent), and X-energy (the Xe-100). To gather the information, MPR visited each of these companies and selected candidate suppliers to tour their facilities and conduct in-person interviews. Information was also obtained by telephone interviews, internet research, and literature review of technical reports. The overarching conclusion from this study is that the U.S. has existing capability or can readily develop capability to support and sustain most, but not all, aspects of commercial deployment of advanced reactors and SMRs. The greatest concerns for capability gaps in the U.S. infrastructure for supply of components for SMRs and advanced reactors are associated with reactor pressure vessels, steam generators, and fuel fabrication. Foreign vendors can supply reactor pressure vessels and steam generators. Issues with fuel fabrication are associated with uranium enrichment and development of fabrication methods for the innovative fuel concepts planned for advanced reactors. The technical capabilities of U.S. manufacturers appear satisfactory with the possible exceptions noted above. However, the capacity of the existing U.S. manufacturing infrastructure may be insufficient for a high rate of construction. Manufacturers contacted as part of this study indicated a willingness to expand capacity if sufficient orders are in place. The review of U.S. infrastructure included a detailed review of capabilities in many different processes including forging, tube drawing, machining, fabrication, and testing. In addition, MPR researched the advanced manufacturing processes for additive manufacturing (i.e., 3D printing) and powder metallurgy - hot isostatic pressing. Reactor plant designers and candidate suppliers are planning to rely on traditional processes for the initial plants. If advanced manufacturing processes are developed and accepted by the nuclear industry, these processes may be used in the future. In the immediate term, ongoing research (much of which is funded by DOE) is necessary to develop and refine advanced manufacturing methods. This report includes several recommendations for DOE that will advance deployment of SMRs and advanced reactors from a manufacturing perspective and other areas identified during the course of this project.

36 MATERIALS SCIENCE↗
DOE OSTI2019

Advanced Reactor Technologies Advanced Reactor Technologies Program Management Plan

The Advanced Reactor Technologies (ART) program grew out of earlier Department of Energy efforts to promote next-generation, small modular, and other advanced reactor concepts. It is now sponsored by Office of Advanced Reactor Deployment, NE-52. Because of its peculiar genesis, the program resides at multiple national laboratories and brings in researchers from subcontracting entities and universities. This creates a need for Idaho National Laboratory’s (INL’s) ART to authorize and define work through memorandum purchase orders, inter-entity work orders, and subcontracts. ART manages research and development (R&D) for the Gas-Cooled Reactors (GCRs) Campaign and other advanced-reactor technologies and ensures that Nuclear Regulatory Commission (requirements and stakeholder needs are factored into the R&D activities. The GCR Campaign supports the Next Generation Nuclear Plant High Temperature Reactor as outlined in the Energy Policy Act of 20051 by

42 ENGINEERING↗
DOE OSTI2019

Controllable 3D Manufacturing System

Next generation Small Modular Reactor (SMR) requires new and improved techniques, instrumentations and strategies to deal with the anticipated high radiation and high temperature environment (esp. in reactors). Component manufacturing technologies will be required that take full advantage of the new 3-D printing methods employed by additive manufacturing technologies. These manufacturing methods must be capable of producing components or sub components on a limited production basis and with nuclear quality. Based on our success in developing a variety of award-winning high energy and high power ultrafast fiber lasers and our pioneered work in fs laser 3D manufacturing (US patents, 9643361 and 9770760), PolarOnyx proposes to use tunable fiber lasers to develop smart AM & SM system to make nuclear quality components for SMR. In Phase I, we have achieved critical milestones and experimentally demonstrated the most critical functions of the proposed controllable laser AM: • Developed and investigated a series of NDI sensing technologies. Combination of IR thermal image and high resolution visible camera provides a powerful tool to characterize thermal distribution and defects. • Built an in-line laser induced breakdown spectroscopy (LIBS) to real time detect compositions of powder and parts, and melting temperature of the melting pool. It proves the LIBS is a very powerful tool to precisely guide the process. • Filed a patent (Application number 16378485) for integrating IR and visible images and LIBS for intelligent AM system. • Developed control algorithm and proved it works well for overhang experiment. • Achieved high density (>99%) and excellent mechanical properties. • Succeeded in developing a working and reliable calibration method to extract true temperature distribution during AM process. • Demonstrated spatial and temporal pulse shaping methods, and impact on microstructures. • Filed one utility patent • Laid out a solid path for phase II to work on controllable 3D manufacturing system development and optimization and to make components with nuclear quality. Our proprietary fs fiber laser based AM processing offers an ultimate solution for manufacturing high temperature components with high strength for nuclear systems. The 3D AM system in conjunction with the intelligent sensing and control is energy efficient and cost effective for the growing nuclear industries. Successful development of the proposed AM technology will provide a vital tool to solve existing and potential issues. This technology will be directly applied in the 3D printing area with combined market share of $10(s) of billions. The achievement in this research will have significant and wide social, scientific, and economic impacts by providing affordable low cost intelligent AM/SM systems, laser micromachining systems, improving energy efficiency, and reducing fabrication and operational cost. Having demonstrated the proof of concept in Phase I, we will prototype the controllable AM/SM system and investigate the efficiency and qualification of fabricating various nuclear quality components in Phase II.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Advancement of Supercritical Carbon Dioxide Technology through Round Robin Testing and Fundamental Modeling

Growing interest in supercritical carbon dioxide (sCO 2 ) cycles for advanced reactors is driving the need for corrosion data on candidate plant materials. The sCO 2 Brayton cycle is being considered for power conversion systems for a number of nuclear reactor concepts including the sodium fast reactor, fluoride salt-cooled high temperature reactor, high temperature gas reactor, and several types of small modular reactors. Multiple organizations have developed test facilities to address the corrosion data knowledge gap in high temperature, high pressure sCO 2 environments but, to date, there has been no formal test program among these organizations to validate the consistency of the data they produce. A demonstration of comparable and reproducible results enables a coordinated effort to explore the sCO 2 parameter space relevant to advanced reactor technology. This proposal establishes a round robin test plan for sCO 2 corrosion testing and the organization of a sCO 2 Materials Group to guide future materials testing directions. Furthermore, this proposal outlines efforts to elucidate the mechanisms of sCO 2 corrosion by performing identical tests in supercritical steam, testing model alloys with various composition and environmental conditions, accompanied by modeling efforts to study rate controlling mechanisms of corrosion and to help understanding the effects of impurities.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Advanced Reactor Siting Policy Considerations

This report describes considerations for and challenges to revising the US Nuclear Regulatory Commission (NRC) guidance regarding population for siting an advanced reactor. Current regulations in 10 Code of Federal Regulations (CFR) Part 100, Reactor Site Criteria, and guidance support in Regulatory Guide (RG) 4.7, Revision 3, General Site Suitability Criteria for Nuclear Power Stations, are focused on large light water reactors (LWRs), addressing the distance that a nuclear reactor should be sited from a densely populated center. This report analyzes the alternatives to using a formula distance as the primary figure of merit for siting small modular reactors (SMRs) or non-light water reactors (non-LWRs). Some reactor vendors use a business plan that sites reactors close to an industrial partner to supply heat or to back-fit a reactor at a fossil plant site to take advantage of existing infrastructure. Factors such as smaller source terms (fission product release to the environment), passive safety systems, advances in barrier technology, advances in simulation and modeling, and improved understanding of societal risk may allow for siting these types of reactors closer to densely populated centers than has historically been accepted for large LWRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Demonstration of Fully Coupled Calculations in Helical Steam Generator: Toward Predictions of Fluid Elastic Instability

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. In addition to being more cost- and space-efficient, such steam generators need to be more reliable, since failure of the primary vessel represents a potential loss of coolant and a safety concern. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict their occurrence based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the design stage. Over the past seven years, the Reactor Product Line has developed the integrated multi-physics code suite SHARP. The goal of developing such a tool is to perform multi-physics neutronics, thermal/fluid, and structural mechanics modeling of the components inside the full reactor core or portions of it with a user-specified fidelity. In particular SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. These tools form a strong basis on which to build a flow-induced vibration modeling capability. In the last two years a significant effort was devoted to 1) develop and validate a fully coupled capability in SHARP for Fluid-Structure calculations (FSI) and 2) validate the use of SHARP for one-way coupled calculations in helical steam generators on legacy data available as well as novel experiments. In this report we discuss an extension of previous one-way coupled calculations performed with Nek5000 and Diablo aimed at simulating available FIV experiments in helical steam generators in the turbulent buffeting regime. It was demonstrated that in the buffeting regime one-way coupling was judged sufficient because the pressure loads do not cause substantial displacements. However, higher speeds led to a reduction in accuracy and the need for a fully coupled capability for velocity or approaching the critical velocity and fluid elastic instability. In this report we demonstrate such fully coupled capability on the Argonne test case discussed in previous reports.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Coupled Calculations of the NuScale FIV Tests – Pre-test Simulations and Sensitivity Analysis

The NEAMS program aims to develop an integrated multi-physics simulation capability “pellet-to-plant” for the design and analysis of future generations of nuclear power plants. In particular, the Reactor Product Line code suite's multi-resolution hierarchy is being designed to ultimately span the full range of length and time scales present in relevant reactor design and safety analyses, as well as scale from desktop to petaflop computing platforms. Flow-induced vibration (FIV) is widespread problem in energy systems because they rely on fluid movement for energy conversion. Vibrating structures may be damaged as fatigue or wear occurs. Given the importance of reliable components in the nuclear industry, flow-induced vibration has long been a major concern in safety and operation of nuclear reactors. In particular, nuclear fuel rods and steam generators have been known to suffer from flow-induced vibration and related failures. Advanced reactors, such as integral Pressurized Water Reactors (PWRs) considered for Small Modular Reactors (SMR), often rely on innovative component designs to meet cost and safety targets. One component that is the subject of advanced designs is the steam generator, some designs of which forego the usual shell-and-tube architecture in order to fit within the primary vessel. A significant amount of data exists on flow-induced vibration in shell-and-tube heat exchangers, and heuristic methods are available to predict FIV based on a set of given assumptions. In contrast, advanced designs have far less data available. Advanced modeling and simulation based on coupled structural and fluid simulations have the potential to predict flow-induced vibration in a variety of designs, reducing the need for expensive experimental programs, especially at the conceptual design stage. Over the past several years, the Reactor Product Line has developed a FIV capability in SHARP. SHARP contains high-fidelity single-physics codes Diablo for structural mechanics and Nek5000 for fluid mechanics calculations. Both codes are state-of-the-art, highly scalable tools that have been extensively validated. In previous reports we have discussed the implementation and validation of an FIV capability for helical steam generators and fuel assemblies. In this report we discuss one-way coupled calculations performed with Nek5000 and Diablo for the NuScale FIV tests. In the turbulent fretting regime one-way coupling is judged to be sufficient because the pressure loads do not cause substantial displacements. It is also the most common source of vibration in helical steam generators per the low flows expected in integral PWRs. The simulations precede the tests, as experimental data was not available at the delivery of this report. Several assumptions had to be made on the support-tube interaction and multiple tests and sensitivity analysis have been performed. Additional work remains to be done and it is discussed in the future work section.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗