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Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.

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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Rancor Microworld Simulation Environment for Nuclear Process Control

The Rancor Microworld is a simplified nuclear power plant simulator. It represents the process of an integrated pressurized light water reactor core with a secondary Rankine cycle turbine/generator. The reactor core is governed by simplified core physics. Operators have the ability to raise and lower control rods to change the reactivity and consequently generate more or less electricity. The microworld provides a digital human machine interface for operating and monitoring the plant efficiently to maximize profits within the simulation. The simulation supports multi-unit operations analogous to small modular reactor plants and supports faults on various components within each unit to simulate some common training scenarios such as reactor and turbine trip events. The microworld includes experimental modules to log process parameters and both automated and user control inputs and events, administer situation awareness probes, and interface with eye tracking software.

Ulrich, ThomasA.↗
DOE OSTI2022

Initial study on cross-section generation requirements for a PBR equilibrium core

A Serpent model of the equilibrium core HTR-PM small modular nuclear reactor in China, was developed for use in cross-section preparation studies in order to guide methods development for the Griffin reactor multiphysics application. The model includes detailed isotopics for 10 distinct pebble burnup groups in 126 core zones with unique fuel and moderator temperatures obtained from a coupled neutronics-thermal-fluids equilibrium core calculation using Griffin-Pronghorn. A sensitivity study of the fuel and moderator temperatures for various core regions was performed with the MOOSE stochastic tools. The results show that the uncertainties are, not unexpectedly, dominated by the value of the fluid temperature and that the power level, heat transfer coefficient and effective conduction to neighboring pebbles and fluid constitute, at best, second order effects. The temperature uncertainty range varies from 28 K to 57 K between the core entry and exit planes, respectively, but these values are probably higher. We still have to quantify the significance of these uncertainties in the preparation of cross-sections in future work. In addition, we verified that the effective pebble approximation used in the PEBBED and V.S.O.P. computer codes works well for the preparation of region averaged cross-sections. Nevertheless, there are some discrepancies in the cross-sections when compared to the multi-pebble model, which could affect the prediction of peak values and the depletion calculation. We conclude that is highly desirable for future studies with Griffin to be able to handle both the 'effective' pebble approximation and the multi-pebble approach for various pebble burnup groups. This enables Griffin users with the flexibility to perform higher-fidelity studies. Finally, we initiated the preparation of cross-sections for various core regions from the full core Serpent reference model. We quantified the differences in 26 group cross-sections from infinite domain models versus the full core approach. These reference cross-sections will serve to verify the double heterogeneity, self-shielding, and spectrum-correction methods in Griffin. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Simulation of CEFR neutronic start-up tests with FENNECS

This paper presents simulation results of selected Neutronic Start-up Tests of the China Experimental Fast Reactor (CEFR) obtained by the neutronics code FENNECS that have been performed within the frame of the IAEA Coordinated Research Program I31032. The Finite Element Neutronics code FENNECS is developed at GRS and solves the few-group steady-state and transient diffusion equation using a Galerkin-based finite element approach. Its main purpose is the safety assessment of Small Modular Reactors and Micro Reactors with complex geometry (e.g., rotating control drums) which gain increased interest internationally. Serpent has been applied to create reference models of the CEFR and cross-section libraries in 10 energy groups for FENNECS. Using these libraries, the following experiments have been simulated with FENNECS: net criticality, control rod integral and differential worth, void reactivity effect, subassembly exchange reactivity effects and reaction rate distribution. The obtained satisfactory agreements with the measurements represent a valuable contribution to the validation of FENNECS. (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Deterministic modelling of the SPERT IV reactor transients using the multi-physics capability of WIMS

The ANSWERS WIMS reactor physics code is being developed for whole core multi-physics modelling. The established neutronics capability for lattice calculations has recently been extended to be suitable for whole core modelling of Small Modular Reactors (SMRs). A whole core transport, SP3 or diffusion flux solution is combined with fuel assembly resonance shielding. An integrated thermal hydraulic solver permits temperature and density variations to feedback to the neutronics calculation. This capability can be applied to both steady state and time dependent transient problems. Nuclear reactor design and safety case development requires assessment of a range of reactor transients, to inform both normal operation limitations and accident scenario analysis. This paper presents new methodology developed in WIMS to couple the core neutronics to the integrated core thermal hydraulics solver for the simulation of reactor transients in whole core models. To support the validation of the multi-physics capability of WIMS, this capability has been applied to the reactivity insertion transient experiments performed in the plate fuelled SPERT-IV reactor. This study employs WIMS using the whole core solver MERLIN, which calculates the time-dependent flux distribution and magnitude, coupled to ARTHUR, which solves for the time-dependent thermal hydraulics solution, and provides thermal feedback via the cross sections generated by GEOM, which performs resonance shielding calculations and generates cross section data for the plate geometry. This spatial kinetics model, with dynamic cross section generation, allows for variations in the temperature and neutron flux profile with time.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Development of Cable Aging Acceptance Criteria for Nuclear Facilities

The aging of nuclear power plant electrical cables has been the subject of substantial research and development (R&D) projects in the nuclear industry, national and international laboratories, universities, and vendor organizations for several years. More specifically, these R&D efforts were conducted to develop equipment and techniques for assessing age-related degradation of in-service cables in nuclear facilities including power plants, research reactors, waste facilities and fuel fabrication plants. In recent years, cable R&D efforts in the nuclear arena have also included work to develop new cables for the next generation of reactors. Over time, exposure to harsh environmental conditions such as elevated temperatures, radiation, and humidity in nuclear installations can result in age-related degradation and failure of cables. In the current fleet of nuclear reactors, there are thousands of miles of cabling installed in each plant and many of these cables are exposed to these harsh environmental conditions. For these cables, the jacket and insulation polymers harden and become brittle over time, making them more susceptible to crack formation and growth, moisture intrusion, and other mechanisms that can lead to cable failure. Moreover, the existing U.S. fleet of 98 nuclear reactors has an average operating age of 38 years. Many of these nuclear power plants have applied for and almost all have been granted regulatory approval for license renewals to operate for 60 years, 20 years beyond their original 40-year life. Further, subsequent license renewals (SLRs) are underway with a few commercial nuclear power sites already approved to operate up to 80 years. As these reactors pursue operating life extensions, utilities must find a way to address issues associated with age-related degradation of cables. Designers of small modular reactors are also in need of technologies to evaluate the performance of cables that will be installed in harsher environments (e.g. higher temperatures, radiation doses, dose rates, etc.) than those present in current-generation reactors. Today, a variety of cable condition monitoring (CM) techniques have been developed and successfully used in nuclear facilities. These techniques are used to identify age-related degradation and assess the condition of cables to determine if their performance characteristics have changed with age. However, objective criteria must be developed for these CM tests to help quantify cable condition and thereby develop repair and replacement schedules. With operating life extending to 80 years and more SLRs on the horizon, both nuclear facilities and regulators need an objective means to determine the aged condition of cables.

36 MATERIALS SCIENCE↗
DOE OSTI2020

Gen IV Education and Training Working Group Webinars’ Initiative

Collaboration and support among national laboratories, industry, universities, and research and development organizations are vital to not only maintain a skilled and competent nuclear workforce but also to avert the risk of human resource shortages. As stated in a recent IAEA paper [1], the world nuclear electrical generating capacity is projected to increase to 554 GW(e) by 2030 and up to 874 GW(e) by 2050 (Fig. 1). This represents a 42% increase over current levels by 2030 and a doubling of the current capacity by 2050. To promote Education and Training (E&T) on Gen IV reactor systems and other nuclear related topics of interest, the Gen IV Education and Training Working Group (ETWG) identifies and advertises training courses; engages collaboration with other international education and training organizations; delivers webinars dedicated to Gen IV systems; and maintains a modern social medium platform to exchange information and ideas on Gen IV R&D topics, as well as on related GIF education and training activities. At the end of June 2020, the GIF ETWG has produced, podcasted and posted forty-one webinars covering the six Gen IV systems and various subjects addressing e.g. the economics of the nuclear fuel cycle, sustainability aspects of Gen IV systems, nuclear fuels and materials challenges, the thorium fuel cycle, energy conversion systems, and lessons learned for knowledge management and preservation. This paper describes the development of these webinars from the initial concept to its full realization, with future webinars planned in 2021, addressing topics such as materials and fuels performance for advanced reactors, small modular reactor.

Paviet, Patricia D.↗
DOE OSTI2024

Digital risk analysis in nuclear engineering projects: Designing for safety, performance, reliability, and security

Cyber-informed engineering and security-by-design frameworks are important in promoting the need to identify cybersecurity concerns early in the systems engineering lifecycle so risks from adversarial cyber-attacks can be eliminated or reduced through engineering design practices. In addition to adversarial risk, risk in operational technology systems also includes non-adversarial and unintentional risk from other factors such as human performance errors, environmental conditions, design flaws, and device degradation or failure. This paper introduces a new concept for characterizing digital risk, both adversarial and non-adversarial, and provides the basis for initial research into a novel digital risk analysis approach focused on incorporating attack difficulty into a multi-attribute analysis technique using robust decision-making. This digital risk characterization is also used to frame a discussion on the challenges of competing objectives and competing stakeholder requirements in an integrated energy system project that incorporates a small modular reactor and industrial facility.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Compact fiber-coupled narrowband two-mode squeezed light source

Quantum correlated states of light, such as squeezed states, are a fundamental resource for the development of quantum technologies, as they are needed for applications in quantum metrology, quantum computation, and quantum communications. It is thus critical to develop compact, efficient, and robust sources to generate such states. Here, we report on a compact, narrowband, fiber-coupled source of two-mode squeezed states of light at 795 nm based on four-wave mixing (FWM) in an 85 Rb atomic vapor. The source is designed in a small modular form factor, with two input fiber-coupled beams, the seed and pump beams required for the FWM, and two output fibers, one for each of the modes of the squeezed state. The system is optimized for low pump power (135 mW) to achieve a maximum intensity-difference squeezing of 4.4 dB after the output of fibers at an analysis frequency of 1 MHz. Furthermore, the narrowband nature of the source makes it ideal for atomic-based quantum sensing and quantum networking configurations that rely on atomic quantum memories. Such a source paves the way for a versatile and portable platform for applications in quantum information science.

Jain, Umang [University of Oklahoma, Norman, OK (U↗
DOE OSTI2019

Synchrotron Radiation Beam Diagnostics at IOTA - Commissioning Performance and Upgrade Efforts

The Integrable Optics Test Accelerator is a research electron and proton storage ring recently commissioned at the Fermilab Accelerator Science and Technology facility. A key part of its beam diagnostics suite are synchrotron radiation monitors, used for measuring transverse beam profile, position, and intensity. In this paper, we report on the performance and uses of this system during the year 1 run. We demonstrate sub-100nm statistical beam position uncertainty and high dynamic range from 10⁹ electrons down to a single electron. Commissioning challenges and operational issues are discussed. We conclude by outlining current upgrade efforts, including improved modularity, small emittance measurements, and a multi-anode photomultiplier system for turn-by-turn acquisition.

43 PARTICLE ACCELERATORS↗
DOE OSTI2012

Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 Steel

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in elevated temperature design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module. LANL staff immediately proposed HT-9 for the reactor vessel and components, as well as fuel clad and ducting, due to its superior thermal qualities. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, there are several key deliverables that must first be successfully accomplished. The most important key deliverable is the research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which becomes the basis for an ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Although this report provides a cursory review of the available data, a much more comprehensive study of open-source data would be necessary. This report serves three purposes: (a) provides a list of already existing material data information that could ultimately be made available to the ASME Code, (b) determines the HT-9 material properties data missing from available sources that would be required and (c) estimates the necessary material testing required to close the gap. Ultimately, the gap analysis demonstrates that certain material properties testing will be required to fulfill the necessary information package for an ASME Code Case.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

ASME Code Case Development Strategy for Ferritic/Martensitic HT-9 Steel

The US Department of Energy (DOE), Office of Nuclear Energy (NE), is supporting the development of an ASME Code Case for adoption of 12Cr-1Mo-VW ferritic/martensitic (F/M) steel, commonly known as HT-9, primarily for use in design of liquid-metal fast reactors (LMFR) and components. In 2011, Los Alamos National Laboratory (LANL) nuclear engineering staff began assisting in the development of a small modular reactor (SMR) design concept, previously known as the Hyperion Module, now called the Gen4 Module, being developed by Gen4 Energy, Inc. located in Denver, Colorado. LANL engineering staff had initially proposed either T91 (9Cr-1Mo-VNb) or HT-9 for the reactor vessel and components, as well as fuel clad and ducting. Although T91 is currently an approved/authorized material for elevated temperature design in the ASME Code, the preference is utilization of HT-9 due to its extensive irradiation database and similar creep resistance as T91. In addition, HT-9 has relatively good corrosion properties for fast reactor environment, such as liquid lead bismuth eutectic coolant. Although the ASME material Code Case, for adoption of HT-9 as an approved elevated temperature material for LMFR service, is the ultimate goal of this project, several key deliverables have been accomplished in FY2012 to provide assurance to DOE/NE of the viability and feasibility of pursuing this goal. An important key deliverable was the recently completed, Gap Analysis of Material Properties Data for Ferritic/Martensitic HT-9 (1). This report provided a cursory research, accumulation, and documentation of specific material parameters; physical, mechanical, and environmental, which are necessary for ensuring a successful ASME Code Case. Time-independent tensile and ductility data and time-dependent creep and creep-rupture behavior are some of the material properties required for a successful ASME Code case. Unfortunately, results of the Gap Analysis report show that many of the material properties have limited data availability in open-source documentation, or are contained in Applied Technology documents, or there is insufficient data to support the ASME requirement from 3-separate heats of material.

36 MATERIALS SCIENCE↗
DOE OSTI2012

Digial Technology Qualification Task 2 - Suitability of Digital Alternatives to Analog Sensors and Actuators

The next generation reactors in the U.S. are an opportunity for vendors to build new reactor technology with advanced Instrumentation and Control Systems (control rooms, DCS, etc.). The advances made in the development of many current generation operating reactors in other parts of the world are being used in the design and construction of new plants. These new plants are expected to have fully integrated digital control rooms, computerized procedures, integrated surveillance testing with on-line monitoring and a major effort toward improving the O&M and fault survivability of the overall systems. In addition the designs are also incorporating major improvements in the man-machine interface based on lessons learned in nuclear and other industries. The above relates primarily to the scope of supply in instrumentation and control systems addressed by Chapter 7 of the Standard Review Plan (SRP) NUREG-0800 (Reference 9.5), and the associated Balance of Plant (BOP) I&C systems. This does not relate directly to the actuator and motor, breaker, initiation circuitry, valve position, etc. which is the subject of this report and normally outside of the traditional Distributed Control System (DCS), for both safety and non-safety systems. The recommendations presented in this report will be used as input to I&C research programming for the implementation of lessons learned during the early phases of new build both for large light water reactors (LWR) and also small modular reactors (SMR). This report is intended to support current research plans and provide user (vendor, owner-operator) input to the optimization of these research plans.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2012

EngenuitySC Commercialization and Entrepreneurial Training Project

A team led by EngenuitySC has performed education and outreach on development of advanced energy markets that will enable wider use of clean energy technologies. This report details the efforts that have made significant advances to improve the market place through education, outreach, and increased communications between industry members. The project resulted in two self-funded industry clusters known as the Fuel Cell Collaborative and NuHub. This project has focused on building and strengthening the leading clean energy clusters in South Carolina: nuclear energy and fuel cell technologies. For the nuclear industry, a new cluster was developed that is now known as NuHub. This cluster has already engaged over 25 nuclear industry leaders or suppliers, four public sector partners, six community economic development foundations, and nearly ten academic partners in a 175 mile radius between Augusta, Georgia and Charlotte, North Carolina. Our outreach has touched over 2,000 stakeholders through the website alone, not including the public audiences and members of the business community reached through news stories and releases that were distributed to over 620 print and online publications. NuHub has established a formal leadership structure, developed subcommittees to focus on industry issues, instituted educational programs for the workforce, and created an industry funding structure that will sustain the industry cluster and mission. NuHub has participated in a wide-variety of community building and outreach activities since its formation under this grant. In the two years since its creation in 2010, we have initiated efforts focused in four main areas that correlate with the four NuHub subcommittees including: innovation, workforce development, industry engagement, and marketing and communications. NuHub successfully raised over $\$$160,000 in both public and private funding, which has supported work to grow the cluster and engage partners including NuScale, Fluor, and Holtec International for research about deployment of advanced small modular reactor (SMR) technologies. The workforce training efforts from NuHub have focused on assisting existing industry to fill positions needed to construct and operate new nuclear plants being built at the VC Summer plant in Jenkinsville, SC and at Plant Votgle in Augusta, Georgia both of whom are constructing the first nuclear reactors (Westinghouse AP 1000 units), to be built in over 30 years. This includes a partnership with Midlands Technical College to train reactor operators and the development of training facilities to support workforce development activities. It is anticipated that approximately 70 students a year will be trained through these programs in the next five years, and it will be expanded to meet new industry needs.

08 HYDROGEN↗
DOE OSTI2012

Reactor Subsystem Simulation for Nuclear Hybrid Energy Systems

Nuclear Hybrid Energy Systems (NHES) could be a key part of the solution to achieving energy security, may provide reliable power availability even with increasing renewable energy penetration into the power grid, and may allow repurposing excess electricity in times of low demand. Incorporating a fission-based power source in a multi-output system (electricity and process heat) can offer significant advantages over carbon-based production sources, such as coal or natural gas, including reduction in atmospheric carbon emissions. In an integrated multi-output system, thermal energy from the nuclear reactor subsystem can be diverted to industrial applications in times of low electricity demand. High temperature, high quality heat from advanced reactor designs might be used for hydrogen production or synthetic fuel production (coal-to-liquid or natural gas-to-liquid processes). Low temperature heat from advanced reactor designs or the current and future fleet of light water reactors (LWRs) might be applied to desalination processes or be stored for later use in low temperature ammonia or hydrocarbon-based thermal cycles. The NHES research team is currently developing a dynamic simulation of an integrated hybrid energy system. A detailed simulation of proposed NHES architectures will allow initial computational demonstration of a tightly coupled NHES to identify key reactor subsystem requirements, identify candidate reactor technologies for a hybrid system, and identify key challenges to operation of the coupled system. This work will provide a baseline for later coupling of design-specific reactor models through industry collaboration. The modeling capability addressed in this report focuses on the reactor subsystem simulation. Many computational tools and methods are available for modeling a nuclear reactor, power conversion systems and associated process applications. One method is to develop a single program capable of modeling the entire system. Another is to combine two or more existing computer programs, each modeling a portion of the system for which it is validated, where the programs exchange information to compute the solution. The former method requires a lengthy development and verification and validation process. The latter method leverages industry standard software to reduce development and funding requirements. Preliminary system models have been developed by Idaho National Laboratory researchers and are currently being enhanced to assess integrated system performance given multiple sources (e.g., nuclear + wind) and multiple applications (i.e., electricity + process heat). Initial efforts to integrate a Fortran-based simulation of a small modular reactor (SMR) with the balance of plant model have been completed in FY12. This initial effort takes advantage of an existing SMR model developed at North Carolina State University to provide initial integrated system simulation for a relatively low cost. The SMR subsystem simulation details are discussed in this report.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

A Review of Sensor Calibration Monitoring for Calibration Interval Extension in Nuclear Power Plants

Currently in the United States, periodic sensor recalibration is required for all safety-related sensors, typically occurring at every refueling outage, and it has emerged as a critical path item for shortening outage duration in some plants. Online monitoring can be employed to identify those sensors that require calibration, allowing for calibration of only those sensors that need it. International application of calibration monitoring, such as at the Sizewell B plant in United Kingdom, has shown that sensors may operate for eight years, or longer, within calibration tolerances. This issue is expected to also be important as the United States looks to the next generation of reactor designs (such as small modular reactors and advanced concepts), given the anticipated longer refueling cycles, proposed advanced sensors, and digital instrumentation and control systems. The U.S. Nuclear Regulatory Commission (NRC) accepted the general concept of online monitoring for sensor calibration monitoring in 2000, but no U.S. plants have been granted the necessary license amendment to apply it. This report presents a state-of-the-art assessment of online calibration monitoring in the nuclear power industry, including sensors, calibration practice, and online monitoring algorithms. This assessment identifies key research needs and gaps that prohibit integration of the NRC-approved online calibration monitoring system in the U.S. nuclear industry. Several needs are identified, including the quantification of uncertainty in online calibration assessment; accurate determination of calibration acceptance criteria and quantification of the effect of acceptance criteria variability on system performance; and assessment of the feasibility of using virtual sensor estimates to replace identified faulty sensors in order to extend operation to the next convenient maintenance opportunity. Understanding the degradation of sensors and the impact of this degradation on signals is key to developing technical basis to support acceptance criteria and set point decisions, particularly for advanced sensors which do not yet have a cumulative history of operating performance.

Calibration↗
DOE OSTI2013

Trial Application of the Facility Safeguardability Assessment Process to the NuScale SMR Design

FSA is a screening process intended to focus a facility designer’s attention on the aspects of their facility or process design that would most benefit from application of SBD principles and practices. The process is meant to identify the most relevant guidance within the SBD tools for enhancing the safeguardability of the design. In fiscal year (FY) 2012, NNSA sponsored PNNL to evaluate the practical application of FSA by applying it to the NuScale small modular nuclear power plant. This report documents the application of the FSA process, presenting conclusions regarding its efficiency and robustness. It describes the NuScale safeguards design concept and presents functional "infrastructure" guidelines that were developed using the FSA process.

Coles, Garill A.↗
DOE OSTI2013

Development of a Generalized Methodology for Soil-Structure Interaction Analysis Using Nonlinear Time-Domain TechniquesNEAMS Program, DOE Office of Nuclear Energy (NE-41)

A generalized time-domain method for Soil-Structure Interaction Analysis is developed, based upon an extension of the Bielak Method. The methodology is combined with the use of a simple hysteretic soil model based upon the Ramberg-Osgood formulation and applied to a notional Small Modular Reactor. These benchmark results compare well with those obtained by using the industry-standard frequency-domain code SASSI. The methodology provides a path forward for investigation of other sources of nonlinearity, including those associated with the use of more physically-realistic material models incorporating pore-pressure effects, gap opening/closing, the effect of nonlinear structural elements, and 3D seismic inputs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Framework For Human-Automation Collaboration: Project Status Report

The Human-Automation Collaboration (HAC) research project within the Department of Energy’s Advanced Small Modular Reactor (AdvSMR) program is investigating how advanced technologies that are planned for AdvSMRs might 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. As one step to accomplish this goal, the research team reviewed available information on AdvSMR designs and identified plant functions and operator tasks that might impact the design of the HAC. The identified functions and tasks will later be vetted against insights learned from the current fleet of Light Water Reactors (LWR), other industries where humans are collaborating with highly automated systems, and the results from the activities in the AdvSMR Concepts of Operations project to inform a high-level generic design model. This generic model will be used as a basis for experimental studies to study the effects of different HAC concepts as well as evaluating new human-system interface (HSI) solutions to support a well-functioning HAC. Another important part needed to accomplish the research goal is a model of the collaboration between the human operator and the automated system. The team has developed a model of HAC that defines: (1) the important design dimensions of automation that impact automation’s use by personnel and integrated human-automation performance, (2) what aspects of human cognition, behavior, and performance mediate automation’s use by personnel, and (3) when and how the above factors affect the use of automation and the overall performance of the HAC system. The HAC model is utilized to identify areas in need of additional research, which forms the basis for planned HAC research activities. By detailing the relationships between characteristics of HAC, automation, and human cognition, the HAC model identifies what aspects of the human-automation interaction are important to consider in developing automation for AdvSMR systems. Three analytical studies have been initiated to address questions related to the HAC model. These studies are 1) Models of Teamwork, 2) Standardized HAC Performance Measurement Battery, and 3) Initiators and Triggering Conditions for Adaptive Automation. Additionally, a number of field studies have been identified and are in progress, and a set of experimental studies that are on schedule to be conducted later in the project to address the main research question. This report consists of three main parts: the preliminary identification of AdvSMR functions and tasks, a description of the updates and extensions to the HAC model, and a description of ongoing and planned analytical, field, and experimental research efforts that the research team has developed to accomplish the project’s overall goal of establishing a technical basis for designing the collaboration of personnel and automation.

99 GENERAL AND MISCELLANEOUS↗