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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 595 records · Page 33
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Investigating Wireless Quantum Key Distribution for Advanced Reactor Communications

Remote operation of small modular reactor (SMR) facilities is an appealing prospect for streamlined operation costs, safety concerns, and user convenience. The ability to guarantee security of communication channels between offsite users and nuclear reactor facilities is critical to enabling remote operations, given the sensitive nature of reactor state data. This report proposes a configuration implementing quantum key distribution (QKD) and advanced encryption standard (AES) protocols to establish secure reactor data feeds. The model outlined, which utilizes the Argonne-developed SeQUeNCe software package and custom MATLAB code, determines free-space channel losses for microwave-band communications in a variety of weather conditions and simulates QKD via ground-satellite links. This report demonstrates the feasibility of repeatable, real-time key generation and distribution for this configuration towards enabling secure wireless reactor communications.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

Summary of GNU Octave Lady Godiva Simulator

A pseudo real-time interactive reactor simulator for the Lady Godiva reactor has been developed using the GNU Octave software package. The simulator tracks the time-dependent power and fuel temperature behavior of the Lady Godiva reactor during sub-prompt critical transients. Users interact with the simulator by controlling reactivity input. This input occurs through users repositioning a control rod as desired during simulator operation. Various reactor parameters are displayed graphically in near “real time,” facilitating user assessment and feedback input. The reactor model consists of a lumped-fuel core, a single control rod, and a simple heat transfer mechanism that allows heat to be rejected from the core. The goal of the simulator is to demonstrate in real-time the system response to reactivity perturbations, the effect of reactivity feedback, and the effect of temperature-dependent equilibrium conditions. The dynamic behavior of the reactor is primarily driven by the relative characteristic time-constants between the delayed neutrons and the overall heat transfer parameter. The simulator will allow users to gain experience performing standard reactor operations, such as establishing steady-state at different fuel temperatures and making power level changes

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Control strategies for coupling thermal energy storage systems with small modular reactors - 188

The contribution of intermittent (renewable) energy sources such as wind and solar continues to increase as renewables improve in both efficiency and price-point. However, the variability of renewables generates additional challenges for the electric grid in the form of rapidly varying electric loads. Proposed options for accommodating this load have included operating nuclear reactors in a load follow mode, or operating the reactor at or near steady state and bypassing steam directly to the condenser. Both of these strategies result in lost energy potential. In addition to lost energy potential, load follow operation can result in increased stress on the fuel and other mechanical components. A more attractive approach is to operate the reactor at or near steady state and bypass excess steam to a thermal energy storage system. The thermal energy can then be recovered, either for electric generation during periods of peak electric demand, or for use in ancillary applications such as desalination and hydrogen production. Sensible Heat Thermal Energy Storage is a mature technology currently used in solar energy systems. This research focuses on the design and coupling of such a system to Small Modular Reactors (SMRs), typical of Integral Pressurized Water Reactor (IPWR) designs currently under development. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2025

SAM Two-Phase Flow Model Development and Applications for Operational Transients in Advanced Reactors

As advanced nuclear technologies continue to develop, the need for the flexible operation and generation of these advanced reactors becomes necessary to maximize economic potential. As large-scale experiments are not always feasible, modeling and simulations of advanced reactors play a crucial role in design optimization and analysis. The SAM (System Analysis Module) code developed at Argonne National Laboratory is a state-of-the-art system-level thermal-hydraulic code aimed at simulating advanced reactor systems. Recent code developments have implemented two-phase flow modeling using the homogeneous equilibrium model, and a new steam generator component has been developed to utilize the two-phase flow implementation. In addition to verification tests, a load-following simulation was performed to model a realistic load-following transient in a proposed integrated system consisting of a conceptual advanced reactor known as the Advanced Burner Test Reactor (ABTR) and thermal energy storage (TES) tanks. The integrated system model uses two large TES tanks designed for sodium and a model helical coil steam generator to simulate the operational load-following transient. The flow rates of the feedwater and secondary loops are regulated to meet a prescribed steam generator load consistent with the electricity demand over a 24-h period. In conclusion, the results found the ABTR system was able to maintain stable reactor conditions and primary- and secondary-side characteristics over the course of the load-following transient.

Advanced Burner Test Reactor (ABTR)↗
DOE OSTI2024

Advanced Test Reactor Long-Term (20-Year) Operational Strategy

The Department of Energy Office of Nuclear Energy (DOE-NE) and the Naval Nuclear Propulsion Program (NNPP) have identified the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) as a critical need for continued thermal irradiation testing capability to support and advance U.S. commercial and naval nuclear power systems. The ATR is currently the only suitable test reactor operating in the world that can provide these capabilities at volume. ATR must, then, plan to operate for at least the next 20 years. As of the beginning of 2024, ATR has operated for 57 years, with many of the original reactor structures, systems, and components installed and operated for 60 years. Some ATR Complex and reactor support equipment that is currently in use was installed 75 years ago with the startup of the Materials Testing Reactor (MTR).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS2020

Operational Considerations for Fission Reactors Utilized on Lunar In-Situ Resource Utilization Missions: A Report to the Nuclear Power & Propulsion Technical Discipline Team

This report identifies the considerations related to the operation of space nuclear power reactors that would be utilized for In Situ Resource Utilization (ISRU) missions on the lunar surface. The objective of this study is to evaluate the operational complexities of reactors that would provide electrical and thermal power to a growing lunar ISRU and mining outpost/base.

Fission reactor↗
DOE OSTI2017

Multi-layered adaptative human-machine interface to support NPP operators - 104

The paper is aimed at improvement of situational awareness and reduction of information overload and during transients and abnormal operational modes. We investigate the possibility of implementation of a multi-layered interface concept for presentation of information to NPP operators. Four layers are suggested, namely: the structural layer representing process system in the form of a mimic diagram; the analytical layer intended to support cognitive activity performed by operator; the functional layer aimed at representation of the status of process functions performed by the plant equipment; and the reference layer containing any relevant information depending on the situation. All the layers can be superimposed on each other. Two factors influence on activation of layers, namely: category of operator and operational mode. The process of designing a multi-layered interface is based on functional analysis and illustrated in application to a simple heat generation system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

SFCOMPO-2.0: An OECD NEA database of spent nuclear fuel isotopic assays, reactor design specifications, and operating data

SFCOMPO-2.0 is the new release of the Organisation for Economic Co-operation and Development (OECD) Nuclear Energy Agency (NEA) database of experimental assay measurements. These measurements are isotopic concentrations from destructive radiochemical analyses of spent nuclear fuel (SNF) samples. We supplement the measurements with design information for the fuel assembly and fuel rod from which each sample was taken, as well as with relevant information on operating conditions and characteristics of the host reactors. These data are necessary for modeling and simulation of the isotopic evolution of the fuel during irradiation. SFCOMPO-2.0 has been developed and is maintained by the OECD NEA under the guidance of the Expert Group on Assay Data of Spent Nuclear Fuel (EGADSNF), which is part of the NEA Working Party on Nuclear Criticality Safety (WPNCS). Significant efforts aimed at establishing a thorough, reliable, publicly available resource for code validation and safety applications have led to the capture and standardization of experimental data from 750 SNF samples from more than 40 reactors. These efforts have resulted in the creation of the SFCOMPO-2.0 database, which is publicly available from the NEA Data Bank. Our paper describes the new database, and applications of SFCOMPO-2.0 for computer code validation, integral nuclear data benchmarking, and uncertainty analysis in nuclear waste package analysis are briefly illustrated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Development of the Delphi display concept for alternative explanations decision making support - 120

Human operators of advanced technology systems, such as nuclear power, command and control, and deepwater oil drilling, must be able to evaluate and respond to a wide array of operational anomalies. Unfortunately, these circumstances are often complex and ambiguous, complicating decision making tasks. One common decision making error, called confirmation bias, occurs when decision makers prematurely focus on one explanation instead of systematically considering all plausible alternative explanations that could equally well account for an operational anomaly (Nickerson, 1998). The Delphi display is designed to counteract decision makers' tendencies towards confirmation bias and to encourage them to consider all plausible explanations, including high risk explanations, they might otherwise overlook. An initial Delphi display prototype was designed to help identify the underlying cause of operational anomalies in the complex, high-consequence technology of deepwater drilling. This prototype decision aid incorporates major indicators related to operational anomalies and presents corresponding plausible explanations for each indicator, ranging from high-risk system-critical explanations to lower-risk but still significant events. The Delphi display incorporates several innovative features to help decision makers detect and monitor operational anomalies and view the relationships among indicators and their associated plausible explanations. Preliminary evaluations of an early prototype Delphi display for detecting kicks in deepwater oil drilling operations have been promising. Moreover, the methodology used to design and develop the Delphi display can be readily generalized to numerous other operational contexts that require humans to make system-critical decisions. To illustrate, an initial Delphi display design for nuclear power plant control is described. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Modernizing the Legacy Fission Wire Measurement System for the Advanced Test Reactor-Critical Facility

Operational lifetime extensions of existing research reactors have emphasized the need for refurbishment, replacements, and upgrades to supporting equipment and instrumentation. The Advanced Test Reactor (ATR) at Idaho National Laboratory (INL), which entered service in 1967, has recently completed the sixth core internals change-out and has scheduled operations until at least 2040. Reactor maintenance and operational risk management is critically important in the research reactor community, however supporting measurement systems sometimes get overlooked when maintenance is planned. The Fission Wire Measurement System (FWMS) is a custom measurement system designed in the 1960s to measure the beta-particle activity of irradiated uranium-aluminum fission wires. This measurement is conducted to determine the fission rate profile of the Advanced Reactor Test Critical (ATR-C) facility. The ATR-C is an open-pool, low-power test reactor that was purpose driven to resemble ATR and is used to qualify experiment configurations and verify core models prior to full-power experiment irradiations in ATR. A power distribution measurement in ATR-C uses uranium-aluminum wires that are distributed throughout the ATR-C core to validate simulation and modeling results. These measurements require 340 to 1500 wires to be irradiated and measured within a 12-hour window. The activity of the wires is measured in the required time with the FWMS, which was put into service in 1965 at the Radiation Measurements Laboratory (RML). The system consists of 4 measurement channels and one reference channel, each with a 2-pi proportional gas flow detector and the measurement channels each have an automated sample changer. This legacy system is crucial to the continued operations of ATR and has undergone some minor hardware upgrades since 1965, however the system presently relies on custom control boards, custom gas ion chambers, analog amplifiers/discriminators, and a user interface (UI) for the system written in outdated code. Much of the equipment and software is custom with no commercial replacements or support and limited documentation. The existing control software requires an operating system that is no longer supported, creating more vulnerabilities to continued operations. A project is underway with a third-party vendor to design, build, and document a new control and data acquisition system (CDAS) for the FWMS. The new upgrade will replace the control system, computer, UI, sample changer motors, and main power supply while maintaining the interface with existing detector hardware. The upgraded system will be operated in parallel with the current hardware and software to conduct validation testing. This equipment upgrade demonstrates the commitment at ATR to ensuring successful operations and potential future research reactors at INL.

46 - INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AN↗
NASA NTRS1980

Design of energy-storage reactors for single-winding constant-frequency dc-to-dc converters operating in the discontinuous-reactor-current mode

This paper presents an algorithm and equations for designing the energy-storage reactor for dc-to-dc converters which are constrained to operate in the discontinuous-reactor-current mode. This design procedure applied to the three widely used single-winding configurations: the voltage step-up, the current step-up, and the voltage-or-current step-up converters. A numerical design example is given to illustrate the use of the design algorithm and design equations.

Chen, D. Y.↗
DOE OSTI2021

Recent advances in CO 2 hydrogenation to value-added products — Current challenges and future directions

Climate change, global warming, fossil fuel depletion and rising fuel prices have created great incentives to seek alternative fuel production technologies. CO 2 transformation to value-added products using renewable H 2 has proven to be an emerging solution to enable this goal. In this regard, three different promising processes, namely methane, methanol and hydrocarbon synthesis via CO 2 hydrogenation are thoroughly discussed. In addition, the influential factors affecting process efficiencies such as catalyst design and mechanistic insight, operating conditions as well as reactor types are investigated, with key pathways that dictate catalyst activity and selectivity of the most promising materials described. Furthermore, a brief overview of the reactor configuration and its crucial role in the improving process viability is analyzed. Accordingly, fixed-bed, fluidized-bed, annular and spherical reactors along with H 2 O/H 2 perm-selective membrane reactors are disscussed for hydrocarbon production. In addition, different reactor configurations are compared to assess the best one that is adjustable depending on the reaction mechanism. In conclusion, a corrugated-wall dual-type membrane reactor is proposed as an emerging alternative for CO 2 hydrogenation to value-added products.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2018

Oxide Morphology of C26M at 300 - 600 °C

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on an Oak Ridge National Laboratory developed Fe-12Cr-6Al- 2Mo-0.2Si-0.03Y alloy (designated ‘C26M’). Aging treatments were performed for 100-2000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, the oxide is amorphous. At 500 °C, crystalline α-Al 2 O 3 oxide film develops and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening on the cladding inner surface during reactor operation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Oxide Morphology of a FeCrAl Alloy, Kanthal APMT, following Extended Aging at 300-600C

Iron-chromium-aluminum (FeCrAl) alloys are of interest to the nuclear materials community due to their resistance to high temperature steam oxidation under accident conditions. The present work investigates oxide formation at temperatures relevant to light water reactor cladding operation following extended aging to assess growth kinetics, chemical composition, and microstructure of oxide formation on a commercial FeCrAl alloy, Fe-21wt.%Cr-5wt.%Al-3wt.%Mo (Kanthal APMT). Aging treatments were performed for 100-1000 hours in stagnant air at 300, 400, 500, and 600 °C, respectively. Oxide growth behavior under the investigated conditions follows a logarithmic time dependence. When the oxidization temperature is 400 °C or below, the oxide is amorphous. At 500 °C, isolated crystalline regions start to appear during short period aging time and expand with extended exposures. Crystalline α-Al2O3 oxide film develops at 600 °C and the correlated logarithmic rate constant decreases significantly, indicating enhanced oxidation resistance of the formed oxide film. In addition, Mo segregation at grain boundaries has been observed when the aging temperature exceeds 500 °C. The results of this study can be viewed as an upper bounding result for potential oxide coarsening during reactor operation.

36 MATERIALS SCIENCE↗
DOE OSTI2017

Ensuring operator reliability in digital control rooms - 102

The U.S. fleet of nuclear power plants is aging, and efforts are being made to modernize instrumentation and control systems including integration of digital human-machine interfaces (HMIs) into the control room. Automation and control technologies have advanced since original systems were installed in the plants. This presents unique opportunities and challenges in ensuring the replacements are a benefit to the operators and the overall safe and reliable operation of the plant. We discuss trends we have observed from assisting utilities in their modernization efforts. Automation and evolving control rooms come with their own host of problems that existing human factors and engineering methods fail to capture. Traditional approaches are analytical approaches to ensuring reliability. Here we discuss some of the shortfalls of these traditional approaches as plants become increasingly more complex. We suggest that ensuring reliability for modernized and next generation control room will require expanding the scope of what was traditionally considered human factors as well as exploring novel approaches and methodologies for dealing with complexity and rapidly advancing technology. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Effects of adaptive automation on operator-system performance - 103

Adaptive automation (AA) is the dynamic, real-time change in the degree of automation in response to situational changes. AA systems are being developed in many industries, including for nuclear power plant applications. AA is a promising means of mitigating human performance issues that often arise in highly automated systems. The purpose of this Nuclear Regulatory Commission study was to characterize important AA design features, define the effects of AA on performance, examine the human factors engineering (HFE) guidance available to design and evaluate AA systems, and identify research and development needs. In general, we found that AA improves task performance and situation awareness (SA) while managing workload. Although the research is limited, AA also supports operator recognition of automation failure and recovery. Research on human-system interfaces has identified principles showing promise for supporting SA, including failure detection, and limiting workload. Research further suggests that AA systems are both more effective, and less disruptive and distracting, when they follow rules of interaction similar to those used by human crewmembers. Our general conclusion about the HFE guidance available to designers and reviewers is that it is limited and has not kept pace with automation technology. We discuss several research and development needs that will help support a better understanding of where and how AA should be used and support the development of more comprehensive guidance for AA safety reviews. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Ontology to guide scenario design to evaluate new technologies for control room modernization - 127

Modernization of nuclear power plants (NPPs) are introducing new human-system interface (HSI) technology that may have unanticipated impact on control room operations. Introduction of HSI thus requires careful evaluation and integrated system validation involving human-in-the-loop experiments with high-fidelity simulation and realistic scenarios. To support the technology evaluation in modernization of NPPs, this paper reports our research on constructing an 'scenario development ontology' for the nuclear industry. We conducted a literature review to produce a preliminary flow chart of the scenario development process and verified the flow chart through interviews with ten domain professionals. The interviews confirmed the key steps in the process flow chart and provided a compendium of approximately 100 scenario attributes, which were divided according to their relevance to HSI evaluation. The next step will be to conduct additional interviews to expand and generalize the ontology that can facilitate vendor and industry evaluation of current and future control room modernization projects. This ontology is particularly intended to facilitate vendor and industry evaluation of current and future control room modernization projects. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Study on the assessment method of human error probabilities in the digitalized main control room - 170

Recently, nuclear power plants (NPPs) in Republic of Korea have been constructed with the new type of main control room (MCR) which is called as digitalized MCR. In digitalized MCR, as digital technologies are being adopted, the environment of MCR has been considerably changed. The operators may obtain the plant data via digitalized human system interfaces (HSIs), large display panel (LDP), computerized procedure system (CPS), soft controls and other system features. Accordingly, the necessity to consider the new environment of MCR has been raised when performing human reliability analysis (HRA) methods. However, none of HRA methods consider the newly installed MCRs. The goal of this paper is to propose new frameworks to assess the human error probabilities (HEPs) in digitalized MCR. Generally, the HEPs are obtained by summing the probabilities of diagnosis errors and the probabilities of execution errors and those probabilities are modified by performance shaping factors (PSFs). In this paper, two parts are included; (1) Assessing the probabilities of diagnosis errors, and (2) Assessing the probabilities of execution errors. In this paper, the process to develop the frameworks are introduced and we expect that the HEPs affecting the features of digitalized MCR can be calculated by using the proposed frameworks. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗