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All sourcesNASA NTRSDOE OSTI
DOE OSTI2013

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

This report describes the status of ongoing research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Development of a Technical Basis and Guidance for Advanced SMR Function Allocation

This report presents the results from three key activities for FY13 that influence the definition of new concepts of operations for advanced Small Modular Reactors (AdvSMR: a) the development of a framework for the analysis of the functional environmental, and structural attributes, b) the effect that new technologies and operational concepts would have on the way functions are allocated to humans or machines or combinations of the two, and c) the relationship between new concepts of operations, new function allocations, and human performance requirements.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Summary of the Preliminary Optical ICHMI Design Study: A Preliminary Engineering Design Study for a Standpipe Viewport

This summary report examines an in-vessel optical access concept intended to support standoff optical instrumentation, control and human-machine interface (ICHMI) systems for future advanced small modular reactor (AdvSMR) applications. Optical-based measurement and sensing systems for AdvSMR applications have several key benefits over traditional instrumentation and control systems used to monitor reactor process parameters, such as temperature, flow rate, pressure, and coolant chemistry (Anheier et al. 2013). Direct and continuous visualization of the in-vessel components can be maintained using external cameras. Many optical sensing techniques can be performed remotely using open optical beam path configurations. Not only are in-vessel cables eliminated by these configurations, but also sensitive optical monitoring components (e.g., electronics, lasers, detectors, and cameras) can be placed outside the reactor vessel in the instrument vault, containment building, or other locations where temperatures and radiation levels are much lower. However, the extreme AdvSMR environment present challenges for optical access designs and optical materials. Optical access is not provided in any commercial nuclear power plant or featured in any reactor design, although successful implementation of optical access has been demonstrated in test reactors (Arkani and Gharib 2009). This report outlines the key engineering considerations for an AdvSMR optical access concept. Strict American Society of Mechanical Engineers (ASME) construction codes must be followed for any U.S. nuclear facility component (ASME 2013); however, the scope of this study is to evaluate the preliminary engineering issues for this concept, rather than developing a nuclear-qualified design. In addition, this study does not consider accident design requirements. In-vessel optical access using a standpipe viewport concept serves as a test case to explore the engineering challenges and performance requirements for sodium fast reactor (SFR) and high-temperature gas reactor (HTGR) AdvSMR applications. The expected environmental conditions for deployment are reviewed for both AdvSMR designs. Optical and mechanical materials that maximize component lifetime are evaluated for the standpipe viewport design under these conditions. Optical components and opto-mechanical designs that provide robust optical-to-metal seals and stress-free optical component mounting are identified, and then key performance specifications are developed for a sapphire optical viewport concept. Design strategies are examined that protect the internal optical surfaces from liquid-coolant condensation and impurity deposits. Finally, a conceptual standpipe viewport design that is suggestive of how this concept could be assembled using standard nuclear-qualified pipe components, is presented.

42 ENGINEERING↗
DOE OSTI2013

Preliminary Scaling and controls Analysis of an FHR-HTSE System Idaho National Laboratory Summer 2013 Final Report

For new nuclear reactor system designs to be approved by regulatory agencies like the Nuclear Regulatory Commission (NRC), the details of system operation must be validated with respect to standards of safety, control, and output. A scaled experiment that replicates certain properties of the system can be used to validate compliance with regulatory standards, while avoiding the prohibitive cost and labor required to develop a fully functional prototype system; therefore, designing such an experiment is of special interest to current efforts to develop hybrid energy systems (HES) that integrate small modular reactors (SMRs), renewable energy systems, and industrial process applications such as hydrogen production and desalination. In addition, a scaled experiment can be an economical method of analyzing the interconnections between HES components and understanding the time constants associated between inter-component energy and information flows. This report discusses the results of a preliminary scaling analysis done for the primary loop of a 300 MWth Fluoride-Salt-Cooled High Temperature Reactor (FHR) that is coupled with a High-Temperature Steam Electrolysis system (HTSE), as well as the basic control logic that governs the primary components and the necessary hardware to achieve optimal functionality. The scaled facility will be a 1 MWth system that uses Dowtherm A as the simulant fluid for Flibe (the coolant of choice for the primary loop of molten salt reactors), and can validate the heat transfer and steady-state operational requirements of the 300 MWth prototype. The scaled facility matches the Prandtl and Reynolds numbers associated with steady-state operation of the FHR-HTSE’s primary loop without having to deal with very high temperatures, flow rates, or power inputs. This will allow the facility to run experiments that analyze various thermophysical and fluid-dynamic properties that characterize reactor operation, such as pressure drops, radial temperature distribution, heat exchanger conditions. The facility also has potential to integrate additional components of the prototype system, such as intermediate thermal-hydraulics loops, real-time grid-demand data, energy storage, and HTSE.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2014

Preliminary Scaling and controls Analysis of an FHR-HTSE System Idaho National Laboratory Summer 2013 Final Report

For new nuclear reactor system designs to be approved by regulatory agencies like the Nuclear Regulatory Commission (NRC), the details of system operation must be validated with respect to standards of safety, control, and output. A scaled experiment that replicates certain properties of the system can be used to validate compliance with regulatory standards, while avoiding the prohibitive cost and labor required to develop a fully functional prototype system; therefore, designing such an experiment is of special interest to current efforts to develop hybrid energy systems (HES) that integrate small modular reactors (SMRs), renewable energy systems, and industrial process applications such as hydrogen production and desalination. In addition, a scaled experiment can be an economical method of analyzing the interconnections between HES components and understanding the time constants associated between inter-component energy and information flows. This report discusses the results of a preliminary scaling analysis done for the primary loop of a 300 MWth Fluoride-Salt-Cooled High Temperature Reactor (FHR) that is coupled with a High-Temperature Steam Electrolysis system (HTSE), as well as the basic control logic that governs the primary components and the necessary hardware to achieve optimal functionality. The scaled facility will be a 1 MWth system that uses Dowtherm A as the simulant fluid for Flibe (the coolant of choice for the primary loop of molten salt reactors), and can validate the heat transfer and steady-state operational requirements of the 300 MWth prototype. The scaled facility matches the Prandtl and Reynolds numbers associated with steady-state operation of the FHR-HTSE’s primary loop without having to deal with very high temperatures, flow rates, or power inputs. This will allow the facility to run experiments that analyze various thermophysical and fluid-dynamic properties that characterize reactor operation, such as pressure drops, radial temperature distribution, heat exchanger conditions. The facility also has potential to integrate additional components of the prototype system, such as intermediate thermal-hydraulics loops, real-time grid-demand data, energy storage, and HTSE.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Framework for Human-Automation Collaboration: Conclusions from Four Studies

The Human Automation Collaboration (HAC) research project is investigating how advanced technologies that are planned for Advanced Small Modular Reactors (AdvSMR) will affect the performance and the reliability of the plant from a human factors and human performance perspective. The HAC research effort investigates the consequences of allocating functions between the operators and automated systems. More specifically, the research team is addressing 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. Oxstrand et al. (2013 - March) describes the efforts conducted by the researchers to identify the research needs for HAC. The research team reviewed the literature on HAC, developed a model of HAC, and identified gaps in the existing knowledge of human-automation collaboration. As described in Oxstrand et al. (2013 – June), the team then prioritized the research topics identified based on the specific needs in the context of AdvSMR. The prioritization was based on two sources of input: 1) The preliminary functions and tasks, and 2) The model of HAC. As a result, three analytical studies were planned and conduced; 1) Models of Teamwork, 2) Standardized HAC Performance Measurement Battery, and 3) Initiators and Triggering Conditions for Adaptive Automation. Additionally, one field study was also conducted at Idaho Falls Power.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2013

Johnson Noise Thermometry System Requirements

This document is intended to capture the requirements for the architecture of the developmental electronics for the ORNL-lead drift-free Johnson Noise Thermometry (JNT) project conducted under the Instrumentation, Controls, and Human-Machine Interface (ICHMI) research pathway of the U.S. Department of Energy (DOE) Advanced Small Modular Reactor (SMR) Research and Development (R&D) program. The requirements include not only the performance of the system but also the allowable measurement environment of the probe and the allowable physical environment of the associated electronics. A more extensive project background including the project rationale is available in the initial project report [1].

Britton, Jr., Charles L.↗
DOE OSTI2014

Development of alternate methods of determining integrated SMR source terms

The Nuclear Energy Institute (NEI) Small Modular Reactor (SMR) Licensing Task Force (TF) has been evaluating licensing issues unique and important to iPWRs, ranking these issues, and developing NEI position papers for submittal to the U.S. Nuclear Regulatory Commission (NRC) during the past three years. Papers have been developed and submitted to the NRC in a range of areas including: Price-Anderson Act, NRC annual fees, security, modularity, and staffing. In December, 2012, NEI completed a draft position paper on SMR source terms and participated in an NRC public meeting presenting a summary of this paper, which was subsequently submitted to the NRC. One important conclusion of the source term paper was the evaluation and selection of high importance areas where additional research would have a significant impact on source terms. The highest ranked research area was iPWR containment aerosol natural deposition. The NRC accepts the use of existing aerosol deposition correlations in Regulatory Guide 1.183, but these were developed for large light water reactor (LWR) containments. Application of these correlations to an iPWR design has resulted in greater than a ten-fold reduction of containment airborne aerosol inventory as compared to large LWRs. Development and experimental justification of containment aerosol natural deposition correlations specifically for the unique iPWR containments is expected to result in a large reduction of design basis and beyond-design-basis accident source terms with concomitantly smaller dose to workers and the public. Therefore, NRC acceptance of iPWR containment aerosol natural deposition correlations will directly support the industry’s goal of reducing the Emergency Planning Zone (EPZ) for SMRs. Based on the results in this work, it is clear that thermophoresis is relatively unimportant for iPWRs. Gravitational settling is well understood, and may be the dominant process for a dry environment. Diffusiophoresis and enhanced settling by particle growth are the dominant processes for determining DFs for expected conditions in an iPWR containment. These processes are dependent on the areato-volume (A/V) ratio, which should benefit iPWR designs because these reactors have higher A/Vs compared to existing LWRs.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

PDRD (SR13046) Tritium Production Final Report

Utilizing the results of Texas A&M University (TAMU) senior design projects on tritium production in four different small modular reactors (SMR), the Savannah River National Laboratory’s (SRNL) developed an optimization model evaluating tritium production versus uranium utilization under a FY2013 plant directed research development (PDRD) project. The model is a tool that can evaluate varying scenarios and various reactor designs to maximize the production of tritium per unit of unobligated United States (US) origin uranium that is in limited supply.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2014

3D Modeling Engine Representation Summary Report

Computers have been used for 3D modeling and simulation, but only recently have computational resources been able to give realistic results in a reasonable time frame for large complex models. This summary report addressed the methods, techniques, and resources used to develop a 3D modeling engine to represent risk analysis simulation for advanced small modular reactor structures and components. The simulations done for this evaluation were focused on external events, specifically tsunami floods, for a hypothetical nuclear power facility on a coastline.

3D Modeling↗
DOE OSTI2014

Implementation of a Bayesian Engine for Uncertainty Analysis

In probabilistic risk assessment, it is important to have an environment where analysts have access to a shared and secured high performance computing and a statistical analysis tool package. As part of the advanced small modular reactor probabilistic risk analysis framework implementation, we have identified the need for advanced Bayesian computations. However, in order to make this technology available to non-specialists, there is also a need of a simplified tool that allows users to author models and evaluate them within this framework. As a proof-of-concept, we have implemented an advanced open source Bayesian inference tool, OpenBUGS, within the browser-based cloud risk analysis framework that is under development at the Idaho National Laboratory. This development, the “OpenBUGS Scripter” has been implemented as a client side, visual web-based and integrated development environment for creating OpenBUGS language scripts. It depends on the shared server environment to execute the generated scripts and to transmit results back to the user. The visual models are in the form of linked diagrams, from which we automatically create the applicable OpenBUGS script that matches the diagram. These diagrams can be saved locally or stored on the server environment to be shared with other users.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2014

Uncertainty Quantification Techniques for Sensor Calibration Monitoring in Nuclear Power Plants

This report describes research towards the development of advanced algorithms for online calibration monitoring. The objective of this research is to develop the next generation of online monitoring technologies for sensor calibration interval extension and signal validation in operating and new reactors. These advances are expected to improve the safety and reliability of current and planned nuclear power systems as a result of higher accuracies and increased reliability of sensors used to monitor key parameters. The focus of this report is on documenting the outcomes of the first phase of R&D under this project, which addressed approaches to uncertainty quantification (UQ) in online monitoring that are data-driven, and can therefore adjust estimates of uncertainty as measurement conditions change. Such data-driven approaches to UQ are necessary to address changing plant conditions, for example, as nuclear power plants experience transients, or as next-generation small modular reactors (SMR) operate in load-following conditions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Critical Heat Flux Phenomena at High Pressure & Low Mass Fluxes: NEUP Final Report Part I: Experiments

This report is a preliminary document presenting an overview of the Critical Heat Flux (CHF) phenomenon, the High Pressure Critical Heat Flux facility (HPCHF), preliminary CHF data acquired, and the future direction of the research. The HPCHF facility has been designed and built to study CHF at high pressure and low mass flux ranges in a rod bundle prototypical of conceptual Small Modular Reactor (SMR) designs. The rod bundle is comprised of four electrically heated rods in a 2x2 square rod bundle with a prototypic chopped-cosine axial power profile and equipped with thermocouples at various axial and circumferential positions embedded in each rod for CHF detection. Experimental test parameters for CHF detection range from pressures of ~80 – 160 bar, mass fluxes of ~400 – 1500 kg/m 2 s, and inlet water subcooling from ~30 – 70°C. The preliminary data base established will be further extended in the future along with comparisons to existing CHF correlations, models, etc. whose application ranges may be applicable to the conditions of SMRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Development of Seismic Isolation Systems Using Periodic Materials

Advanced fast nuclear power plants and small modular fast reactors are composed of thin-walled structures such as pipes; as a result, they do not have sufficient inherent strength to resist seismic loads. Seismic isolation, therefore, is an effective solution for mitigating earthquake hazards for these types of structures. Base isolation, on which numerous studies have been conducted, is a well-defined structure protection system against earthquakes. In conventional isolators, such as high-damping rubber bearings, lead-rubber bearings, and friction pendulum bearings, large relative displacements occur between upper structures and foundations. Only isolation in a horizontal direction is provided; these features are not desirable for the piping systems. The concept of periodic materials, based on the theory of solid-state physics, can be applied to earthquake engineering. The periodic material is a material that possesses distinct characteristics that prevent waves with certain frequencies from being transmitted through it; therefore, this material can be used in structural foundations to block unwanted seismic waves with certain frequencies. The frequency band of periodic material that can filter out waves is called the band gap, and the structural foundation made of periodic material is referred to as the periodic foundation. The design of a nuclear power plant, therefore, can be unified around the desirable feature of a periodic foundation, while the continuous maintenance of the structure is not needed. In this research project, three different types of periodic foundations were studied: one-dimensional, two-dimensional, and three-dimensional. The basic theories of periodic foundations are introduced first to find the band gaps; then the finite element methods are used, to perform parametric analysis, and obtain attenuation zones; finally, experimental programs are conducted, and the test data are analyzed to verify the theory. This procedure shows that the periodic foundation is a promising and effective way to mitigate structural damage caused by earthquake excitation.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

Economic Conditions and Factors Affecting New Nuclear Power Deployment

This report documents work performed in support of the US Department of Energy Office of Nuclear Energy’s Advanced Small Modular Reactor (AdvSMR) program. The report presents information and results from economic analyses to describe current electricity market conditions and those key factors that may impact the deployment of AdvSMRs or any other new nuclear power plants. Thus, this report serves as a reference document for DOE as it moves forward with its plans to develop advanced reactors, including AdvSMRs. For the purpose of this analysis, information on electricity markets and nuclear power plant operating costs will be combined to examine the current state of the nuclear industry and the process required to successfully move forward with new nuclear power in general and AdvSMRs in particular. The current electricity market is generally unfavorable to new nuclear construction, especially in deregulated markets with heavy competition from natural gas and subsidized renewables. The successful and profitable operation of a nuclear power plant (or any power plant) requires the rate at which the electricity is sold to be sufficiently greater than the cost to operate. The wholesale rates in most US markets have settled into values that provide profits for most operating nuclear power plants but are too low to support the added cost of capital recovery for new nuclear construction. There is a strong geographic dependence on the wholesale rate, with some markets currently able to support new nuclear construction. However, there is also a strong geographic dependence on pronuclear public opinion; the areas where power prices are high tend to have unfavorable views on the construction of new nuclear power plants. The use of government-backed incentives, such as subsidies, can help provide a margin to help justify construction projects that otherwise may not seem viable. Similarly, low interest rates for the project will also add a positive margin to the economic analysis. In both cases, the profitable price point is decreased, making more markets open to profitable entry. Overall, the economic attractiveness of a nuclear power construction project is not only a function of its own costs, but a function of the market into which it is deployed. Many of the market characteristics are out of the control of the potential nuclear power plant operators. The decision-making process for the power industry in general is complicated by the short-term market volatility in both the wholesale electricity market and the commodity (natural gas) market. Decisions based on market conditions today may be rendered null and void in six months. With a multiple-year lead time, nuclear power plants are acutely vulnerable to market corrections.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF 2 ) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR concept, and it will demonstrate key operational features of that design. The FHR DR will be closely scaled to the SmAHTR concept in power and flows, so any technologies demonstrated will be directly applicable to a reactor concept of that size. The FHR DR is not a commercial prototype design, but rather a DR that serves a cost and risk mitigation function for a later commercial prototype. It is expected to have a limited operational lifetime compared to a commercial plant. It is designed to be a low-cost reactor compared to more mature advanced prototype DRs. A primary reason to build the FHR DR is to learn about salt reactor technologies and demonstrate solutions to remaining technical gaps.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Status of the phenomena representation, 3D modeling, and cloud-based software architecture development

Early in 2013, researchers at the Idaho National Laboratory outlined a technical framework to support the implementation of state-of-the-art probabilistic risk assessment to predict the safety performance of advanced small modular reactors. From that vision of the advanced framework for risk analysis, specific tasks have been underway in order to implement the framework. This report discusses the current development of a several tasks related to the framework implementation, including a discussion of a 3D physics engine that represents the motion of objects (including collision and debris modeling), cloud-based analysis tools such as a Bayesian-inference engine, and scenario simulations. These tasks were performed during 2015 as part of the technical work associated with the Advanced Reactor Technologies Program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
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↗