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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Materials Characterization using the Facilities of the Center for Synchrotron Radiation Research and Instrumentation (CSRRI), a NSUF Partner Facility

Advanced nuclear energy systems are envisioned to operate at high temperatures, high burn-up, harsh neutron irradiation, and corrosive environments. New materials and nuclear fuels with improved performance, safety and reliability are needed for the increasingly demanding operating conditions. The fundamental challenge in developing new materials with improved irradiation, temperature, and corrosion resistance is to understand and control chemical and physical phenomena in complex systems over multiple time and length scales. Solving these scientific and technical challenges will require an integrated approach of all available computational and experimental capabilities. This talk will focus on the application of synchrotron radiation techniques to characterize materials of interest to nuclear material science at the facilities operated by the Center For Synchrotron Radiation Research and Instrumentation (CSRRI) at the Illinois Institute of Technology. The CSRRI is a partner facility with the U. S. Department of Energy's Nuclear Science User Facility (NSUF). Using the CSRRI facilities, such as the Materials Research Collaborative Access Team (MRCAT) beamline, we can study the materials science of irradiated and model materials to improve the understanding of the performance of materials in nuclear reactors. (author)

36 MATERIALS SCIENCE↗
DOE OSTI2016

Experimental test plan for PWR sister rods in the high burnup spent fuel data project

The High Burnup Spent Fuel Data Project (herein referred to as the 'project'), sponsored by the US Department of Energy (DOE) Office of Nuclear Energy (NE), is focused on experimentally defining the effects of long-term storage and transportation on high burnup (HBU) pressurized water reactor (PWR) fuel (>45 gigawatt days per metric ton uranium). The goals of the project are to provide confirmatory data for model validation and potential improvement; to provide input to future dry storage cask design; to support license renewals and new licenses for Independent Spent Fuel Storage Installation facilities; and to support transportation licensing for high burnup used fuel. In support of these project goals, an instrumented TN-32B bolted-lid cask at the North Anna reactor site will be loaded with intact, HBU PWR fuel assemblies (project assemblies) having four different kinds of cladding: standard Zircaloy-4, low-tin Zircaloy-4, ZIRLO, and M5. In approximately 10 years, the instrumented cask will be opened for inspection and several fuel rods from the project assemblies will be characterized. To support the project goals, the project pulled 25 'sister rods', 9 from the project assemblies and 16 from similar HBU assemblies, for characterization. The 25 sister rods are all high burnup and cover the range of cladding alloys. The 25 sister rods were shipped to Oak Ridge National Laboratory (ORNL) in early 2016 for detailed nondestructive and destructive examination. Examinations are intended to provide baseline data on the initial physical state of the cladding and fuel prior to the loading, drying, and long-term dry storage process. Further examinations are focused on determining the effects of temperatures encountered during and following drying. Similar tests will be performed on rods taken from the project assemblies at the end of the storage period (the 'cask rods') to identify any significant changes in the fuel rods that may have occurred during the dry storage period. Additionally, some of the sister rods will be used for separate effects testing to expand the applicability of the project data to the fleet, and to address some of the data-related gaps associated with extended storage and subsequent transportation of high burnup fuel. A draft test plan is being developed that describes the experimental work to be conducted on the sister rods. This paper summarizes the draft test plan and necessary coordination activities for the multi-year experimental program to supply data relevant to the assessment of the safety of long-term storage followed by transportation of high burnup used fuel. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
NASA NTRS2008

Advanced Thermal Simulator Testing: Thermal Analysis and Test Results

Work at the NASA Marshall Space Flight Center seeks to develop high fidelity, electrically heated thermal simulators that represent fuel elements in a nuclear reactor design to support non-nuclear testing applicable to the development of a space nuclear power or propulsion system. Comparison between the fuel pins and thermal simulators is made at the outer fuel clad surface, which corresponds to the outer sheath surface in the thermal simulator. The thermal simulators that are currently being tested correspond to a SNAP derivative reactor design that could be applied for Lunar surface power. These simulators are designed to meet the geometric and power requirements of a proposed surface power reactor design, accommodate testing of various axial power profiles, and incorporate imbedded instrumentation. This paper reports the results of thermal simulator analysis and testing in a bare element configuration, which does not incorporate active heat removal, and testing in a water-cooled calorimeter designed to mimic the heat removal that would be experienced in a reactor core.

Bragg-Sitton, Shannon M.↗
DOE OSTI2017

Setup for polarized neutron imaging using in situ 3 He cells at the Oak Ridge National Laboratory High Flux Isotope Reactor CG-1D beamline

Here in the present study, we report a new setup for polarized neutron imaging at the ORNL High Flux Isotope Reactor CG-1D beamline using an in situ 3 He polarizer and analyzer. This development is very important for extending the capabilities of the imaging instrument at ORNL providing a polarized beam with a large field-of-view, which can be further used in combination with optical devices like Wolter optics, focusing guides, or other lenses for the development of microscope arrangement. Such a setup can be of advantage for the existing and future imaging beamlines at the pulsed neutron sources. The first proof-of-concept experiment is performed to study the ferromagnetic phase transition in the Fe3Pt sample. Lastly, we also demonstrate that the polychromatic neutron beam in combination with in situ 3He cells can be used as the initial step for the rapid measurement and qualitative analysis of radiographs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Incorporating the latest Neutron NDA analysis techniques into INCC6

International Neutron Coincidence Counting (INCC6) is the IAEA’s neutron NDA software. With FY25 SGTech support we implemented self-diagnostic features into INCC6. For FY26 we propose to implement advanced analysis methods that have been splintered into custom software for one-off instruments. Incorporating these capabilities in INCC6 will make them available for any new or existing application on the ubiquitous INCC6 platform, increasing safeguards efficiency and effectiveness.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

TFR-2574 MARVEL Instrumentation and Control System (ICS)

This document contains the Level 3 requirements associated with the subsystems of the Instrumentation Control System (ICS) in the Microreactor Applications Research Validation and Evaluation (MARVEL) project.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Update on Small Modular Reactors Dynamics System Modeling Tool -- Molten Salt Cooled Architecture

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2026

Postirradiation Examination of WIRE-21 Experiment Irradiated in the High Flux Isotope Reactor

Westinghouse Electric Company is developing wireless sensors to monitor the centerline temperature and internal pressure of commercial light-water reactor fuel rods during irradiation. Oak Ridge National Laboratory and Westinghouse Electric Company developed the Wireless Instrumented RB Experiment 2021 (WIRE-21) to test wireless temperature and pressure sensor technologies in a removable beryllium position in the High Flux Isotope Reactor. The experiment was irradiated for a total of 75 days, at temperatures ranging from approximately 150°C to 400°C, resulting in a peak fast (energy > 0.1 MeV) neutron fluence of about 3 × 10 21 n/cm 2 . The temperature was intentionally cycled multiple times to compare the response of the wireless temperature sensor to collocated thermocouples. Similarly, the pressure sensor was actuated in multiple steps to compare the response of the wireless measurement to excore pressure transducers (Petrie et al., 2023). After irradiation, the experiment was disassembled in the Irradiated Fuels Examination Laboratory (IFEL) hot cell at Oak Ridge National Laboratory with the intent to recover sensor and dosimetry components, document the as-irradiated condition of the hardware, and investigate possible causes of the sensor behavior observed during irradiation. The postirradiation examination successfully recovered and preserved key WIRE-21 components. After the housing was removed using a milling machine, the internal experiment sections were examined. All eight fiber-optic sensors were recovered, cut, and stored. The silicon carbide thermometry, temperature sensor, pressure sensor, lower spacer, and selected pressure and temperature cable sections were also removed and stored. The metal bellows of the pressure sensor was found to be plastically deformed, indicating that it had been properly pressurized during irradiation and generally behaved as expected. X-ray diffraction analysis of a section of one of the irradiated inductor cores within the pressure sensor was performed and confirmed the presence of phase-pure alpha ferrite (i.e., no unexpected phase transformations). Inductance testing was performed on the irradiated pressure sensor cores using an unirradiated test coil, and DC resistance measurements of the transceiver coils were also performed. The measurements with irradiated inductor cores assembled inside unirradiated coils showed slightly reduced inductance compared to measurements made with unirradiated cores, but the difference was not sufficient to explain the more significant reductions in inductance that were observed in-pile. Therefore, it is suspected that degradation of the inductor coils (specifically the wire wrapping) is responsible for the reduced inductance observed in-pile.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

A HUMAN FACTORS META MODEL FOR U.S. NUCLEAR POWER PLANT CONTROL ROOM MODERNIZATION

Over the last several years, the United States (U.S.) Department of Energy (DOE) has sponsored human factors research and development (R&D) and human factors engineering (HFE) activities through its Light Water Reactor Sustainability (LWRS) program to modernize the main control rooms (MCR) of commercial nuclear power plants (NPP). Idaho National Laboratory (INL), in partnership with numerous commercial nuclear utilities, has conducted some of this R&D to enable the life extension of NPPs (i.e., provide the technical basis for the long-term reliability, productivity, safety, and security of U.S. NPPs). From these activities performed to date, a human factors meta model for U.S. NPP control room modernization can now be formulated. This paper discusses this emergent HFE meta model for NPP control room modernization, with the goal of providing an integrated high level roadmap and guidance on how to perform human factors R&D and HFE for those in the U.S. nuclear industry that are engaging in the process of upgrading their MCRs.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2017

Spallation Neutron Source Second Target Station Integrated Systems Update

The Spallation Neutron Source (SNS) was designed from the beginning to accommodate both an accelerator upgrade to increase the proton power and a second target station (STS). Four workshops were organized in 2013 and 2014 to identify key science areas and challenges where neutrons will play a vital role [1-4]. Participants concluded that the addition of STS to the existing ORNL neutron sources was needed to complement the strengths of High Flux Isotope Reactor (HFIR) and the SNS first target station (FTS). To address the capability gaps identified in the workshops, a study was undertaken to identify instrument concepts that could provide the required new science capabilities. The study outlined 22 instrument concepts and presented an initial science case for STS [5]. These instrument concepts formed the basis of a planning suite of instruments whose requirements determined an initial site layout and moderator selection. An STS Technical Design Report (TDR) documented the STS concept based on those choices [6]. Since issue of the TDR, the STS concept has significantly matured as described in this document.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2012

Future Vision for Instrumentation, Information and Control Modernization

A Future Vision of a transformed nuclear plant operating model based on an integrated digital environment has been developed as part of the Advanced Instrumentation, Information, and Control (II&C) research pathway, under the Light Water Reactor (LWR) Sustainability Program. This is a research and development program sponsored by the U.S. Department of Energy (DOE), performed in close collaboration with the nuclear utility industry, to provide the technical foundations for licensing and managing the long-term, safe and economical operation of current nuclear power plants. II&C has been identified as a potential life-limiting issue for the domestic LWR fleet in addressing the reliability and aging concerns of the legacy systems in service today. The Future Vision is based on a digital architecture that encompasses all aspects of plant operations and support, integrating plant systems, plant work processes, and plant workers in a seamless digital environment to enhance nuclear safety, increase productivity, and improve overall plant performance. Pilot projects are being conducted as the means for industry to gain confidence in these new technologies for use in nuclear plant work activities. The pilot projects introduce new digital technologies into the nuclear plant operating environment at host operating plants to demonstrate and validate them for production usage. In turn, the pilot project technologies serve as the stepping stones to the eventual seamless digital environment as described in the Future Vision. Initial project results confirm that the technologies can address provide substantial efficiency and human performance benefits while resolving the reliability and aging concerns.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2015

Online Monitoring of Concrete Structures in Nuclear Power Plants: Interim Report

The existing fleet of nuclear power plants in the United States have initial operating licenses of 40 years, and many of these plants have applied for and received license extensions. As plant structures, systems, and components age, their useful life—considering both structural integrity and performance—is reduced as a result of deterioration of the materials. Assessment and management of aging concrete structures in nuclear plants require a more systematic approach than simple reliance on existing code-based design margins of safety. Structural health monitoring is required to produce actionable information regarding structural integrity that supports operational and maintenance decisions. The online monitoring of concrete structures project conducted under the Advanced Instrumentation, Information, and Control Technologies Pathway of the Light Water Reactor Sustainability program at Idaho National Laboratory is seeking to develop and demonstrate capabilities for concrete structures health monitoring. Through this research project, several national laboratories and Vanderbilt University propose to develop a framework of research activities for the health monitoring of nuclear power plant concrete structures that includes the integration of four elements—damage modeling, monitoring, data analytics, and uncertainty quantification. This report briefly discusses activities in this project during October-December, 2014. The most significant activity during this period was the organizing of a two-day workshop on research needs in online monitoring of concrete structures, hosted by Vanderbilt University in November 2014. Thirty invitees from academia, industry and government participated in the workshop. The presentations and discussions at the workshop surveyed current activities related to concrete structures deterioration modeling and monitoring, and identified the challenges, knowledge gaps, and opportunities for advancing the state of the art; these discussions are summarized in this report

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2025

Towards the development of resonantly enhanced laser-based diagnostics for molten salt reactor safeguards

The adoption of Generation IV molten salt reactors (MSRs) depends on the ability to implement compatible monitoring instrumentation to ensure adherence to nuclear safeguards. This task is nontrivial due to the high temperature, reactive, and chemically complex fuel and coolant components contained within these systems. Optical spectroscopy techniques such as laser-induced fluorescence (LIF) and laser-induced breakdown spectroscopy (LIBS) are candidates for monitoring instrumentation. They offer many advantages for continuous monitoring, including the ability to operate at a standoff and compatibility with liquid-phase analytes. We discuss the use of LIF for the detection of Nd, a common fission fragment, within a liquid matrix. Results show that for NdCl3 dissolved in water, the Nd I 492.45 nm resonant transition is detectable without the need to induce a plasma and that the fluorescence emission can be separated in the time domain from the laser scatter signal. Additionally, we present a setup and preliminary results for measurements of solid Nd and U using resonant LIBS. The results demonstrate element-selective measurement capabilities and provide the foundational data necessary to continue the development of laser-based MSR diagnostics. This work also helps address gaps in the literature regarding energy level assignments of Nd.

98 - NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL↗
DOE OSTI2017

Cybersecurity Vulnerability Assessment Methodologies for Nuclear Power Plants

As any mechanical system ages, broken or outdated parts must be replaced; nuclear power plants (NPPs) are no different. The average age of a NPP in the United States' aging fleet of commercial reactors is 35 years, with many of these plants operating under extensions of their original licenses. These plants, originally designed in the 1960's and 1970's, are controlled by analog instrumentation and control (I and C) systems, and have since been upgraded (to varying degrees) to hybrid analog-digital systems. As these NPPs are upgraded to continue operating and to meet more stringent safety requirements imposed by the Nuclear Regulatory Commission (NRC), the implementation of digital systems to monitor and run the power plant introduces new vulnerabilities that must be addressed. The main objectives of a vulnerability assessment (VA) are to identify, quantify, and prioritize all of the susceptibilities inherent in a system. This allows for all weaknesses to be analyzed, and for the most serious concerns to be addressed first. Traditional VA methodologies deal with the analysis of physical systems associated with the NPP; with the introduction of digital control systems and computer networks, these methodologies must be expanded to include these fundamentally different systems and vulnerability concerns. Digital systems have been used in NPPs since the 1980's, but only for non-safety control systems; attacks on these systems were not deemed consequential, and no dedicated VA methodology was required. However, 'all new NPPs are being designed with integrated digital instrumentation and control systems as the backbone of protection, controls, alarms, and display and monitoring'. This gives rise to new vulnerabilities such as disrupted data displays, the suppression of critical alarms, or unauthorized access to secure systems from connected networks. The NRC released formal cybersecurity regulations in March 2009 (10 CFR 73.54), as well as a guide for the program in January 2010 (RG 5.71). However, the regulations and guide provide limited insight on how to identify these vulnerabilities; the focus is on remediating them once found. Hence, there is a need for a comprehensive, systematic methodology to conduct a VA on digital computer and communications and network systems. In this paper, we review traditional VA methodologies and discuss relevant cases. Then we discuss challenges to vulnerabilities within the context of modernization of the fleet and what needs to be done to address this issue.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Prognostic and health management of active assets in nuclear power plants

This study presents the development of diagnostic and prognostic capabilities for active assets in nuclear power plants (NPPs). The research was performed under the Advanced Instrumentation, Information, and Control Technologies Pathway of the Light Water Reactor Sustainability Program. Idaho National Laboratory researched, developed, implemented, and demonstrated diagnostic and prognostic models for generator step-up transformers (GSUs). The Fleet-Wide Prognostic and Health Management (FW-PHM) Suite software developed by the Electric Power Research Institute was used to perform diagnosis and prognosis. As part of the research activity, Idaho National Laboratory implemented 22 GSU diagnostic models in the Asset Fault Signature Database and two wellestablished GSU prognostic models for the paper winding insulation in the Remaining Useful Life Database of the FW-PHM Suite. The implemented models along with a simulated fault data stream were used to evaluate the diagnostic and prognostic capabilities of the FW-PHM Suite. Knowledge of the operating condition of plant asset gained from diagnosis and prognosis is critical for the safe, productive, and economical long-term operation of the current fleet of NPPs. This research addresses some of the gaps in the current state of technology development and enables effective application of diagnostics and prognostics to nuclear plant assets.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2011

System Upgrades at the Advanced Test Reactor Help Ensure that Nuclear Energy Research Continues at the Idaho National Laboratory

Fully operational in 1967, the Advanced Test Reactor (ATR) is a first-of-its-kind materials test reactor. Located on the Idaho National Laboratory’s desert site, this reactor remains at the forefront of nuclear science, producing extremely high neutron irradiation in a relatively short time span. The Advanced Test Reactor is also the only U.S. reactor that can replicate multiple reactor environments concurrently. The Idaho National Laboratory and the Department of Energy recently invested over 13 million dollars to replace three of ATR’s instrumentation and control systems. The new systems offer the latest software and technology advancements, ensuring the availability of the reactor for future energy research. Engineers and project managers successfully completed the four year project in March while the ATR was in a scheduled maintenance outage. “These new systems represent state-of-the-art monitoring and annunciation capabilities,” said Don Feldman, ATR Station Manager. “They are comparable to systems currently used for advanced reactor designs planned for construction in the U.S. and in operation in some foreign countries.”

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Assessment of Flow-Enhanced Electrochemical Sensor Testing and Deployments

The goal of our activities in FY24 was to enable advanced materials accountancy for molten salt reactors (MSRs) through the development and optimization of robust flow-enhanced electrochemical sensors (FEES) using a safeguards-focused sensor testing platform called the modular flow instrumentation testbed (MFIT). Flow-enhanced electrochemical sensors are a type of electroanalytical sensor that have been developed at Argonne National Laboratory to be installed directly into MSR flow conduits to make measurements of the salt composition. These sensors represent a significant improvement in capabilities compared to earlier electroanalytical sensors that instead can only be operated in quiescent conditions. Previous work has focused on testing of the FEES in flowing conditions provided by the MFIT to assess the accuracy and precision of the sensor measurements. This year, the FEES technical readiness was increased by (1) demonstrating the use of optimized sensor techniques to achieve measurements in very high-concentration MSR-relevant salts, (2) leveraging new multimodal sensors and automation approaches to reduce measurement uncertainties, and (3) fabricating and deploying sensors to industrial partners to improve their ability to make measurements in challenging environments. Additionally, we also ran the sensors during complex system operations involving combined flow and purification activities to demonstrate safeguards-relevant measurements during online processing for MSRs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗