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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 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Radiation Resistant Electrical Insulation Materials for Nuclear Reactors: Final Report

The instrument and control cables in future nuclear reactors will be exposed to temperatures, dose rates, and accumulated doses exceeding those originally anticipated for the 40-year operational life of the nuclear power plant fleet. The use of nanocomposite dielectrics as insulating material for such cables has been considered a route to performance improvement. In this project, nanoparticles were developed and successfully included in three separate material systems [cross-linked polyvinyl alcohol (PVA/XLPVA), cross-linked polyethylene (PE/XLPE), and polyimide (PI)], and the chemical, electrical, and mechanical performance of each was analyzed as a function of environmental exposure and composition. Improvements were found in each material system; however, refinement of each processing pathway is needed, and the consequences of these refinements in the context of thermal, radiation, and moisture exposures should be evaluated before transferring knowledge to industry.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1973

Fabrication of capsule assemblies, phase 3

Thirteen capsule assemblies were fabricated for evaluation of fuel pin design concepts for a fast spectrum lithium cooled compact space power reactor. These instrumented assemblies were designed for real time test of prototype fuel pins. Uranium mononitride fuel pins were encased in AISI 304L stainless steel capsules. Fabrication procedures were fully qualified by process development and assembly qualification tests. Instrumentation reliability was achieved utilizing specially processed and closely controlled thermocouple hot zone fabrication and by thermal screening tests. Overall capsule reliability was achieved with an all electron beam welded assembly.

Keeton, A. R.↗
DOE OSTI2015

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Current Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.

Crack Growth↗
DOE OSTI2022

Predicting deviations in ex-core detector response in Krsko NPP during an earthquake with Monte-Carlo neutron transport methods

To study the response of ex-core detectors during an earthquake, an approach using Monte Carlo code MCNP was developed. As a first step to predict the response of the ex-core detectors a detailed MCNP model of the Krsko reactor core was created. A detailed MCNP core model was used to calculate 3D power distributions inside the core. Calculated power distributions were verified by comparison to the CORD-2 calculations, which is currently used for core design calculation verification of the Krsko nuclear power plant. Due to the outside the core position of the neutron detectors, the hybrid code ADVANTG is used to generate weight windows to speed up neutron transport outside the reactor core. To be able to use ADVANTG, fixed neutron source had to be reconstructed from the criticality core calculation. It was found that the power range ex-core detector movement has a negligible effect on the value of the thermal neutron flux at the active part of the detector. However, the 5 cm radial movement of the intermediate-range detector leads to 7 % - 8 % change in thermal neutron flux within the active intermediate-range detector region. The analysis continued with the evaluation of the effect of core barrel movement on the ex-core detector response. It was determined that the 2 mm core barrel radial oscillation can lead to ∼ 4% change in thermal neutron flux within the active detector region. It was evaluated that reactor core movement inside the baffle can lead up to ∼ 1.9% and ∼ 1.3% deviation in power-range and intermediate-range detector signal respectively. Analysis showed that the mechanical movement of ex-core neutron detectors could not explain the fluctuations in the ex-core detector signal. However, combined core barrel and reactor core inside baffle oscillations could be a probable reason for the observed fluctuations in the ex-core detector signal during an earthquake. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

Conceptual Design Report for the I2 Instrumentation Experiment in ATRC

The US Department of Energy In-Pile Instrumentation (I2) program is tasked with developing advanced sensors for in-pile service as well as supporting the engineering required to implement existing nuclear sensor technologies in US high-power test reactors. The ATRC instrumentation test has been conceived to address both of these missions, specifically with respect to neutron flux measurements. Space has been allocated in the test for Self-Powered Neutron Detectors (SPNDs), fission chambers, and Micro-Pocket Fission Detectors (MPFDs). The real-time flux readings obtained from these instruments will be compared with time integrated data obtained from passive neutron dosimetry (i.e., “flux” wires). This work will be complementary to data obtained from the TREAT MIMIC-N experiment. This experiment is expected to support instrumentation efforts related to the proposed Medium-I loop and to instrumenting loop experiments conducted in the ATR center flux trap. While the MIMIC-N experiment was conducted in a dry environment, the ATRC I-19 experiment will be conducted in an aqueous environment, similar to the center flux trap loop experiments but will take place at ambient temperature and pressure, and at a much lower neutron flux. One of the challenges of the ATRC I-19 experiment will be low amplitude signals produced by the sensors in the low neutron fluxes of ATRC. However, related work conducted several years ago in ATRC was able to reliably detect signals using a similar instrument set.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Machine-Learning-aided Approach for Predicting the Thermal Expansion Behaviors in Advanced Test Reactor Capsules (NURETH-20 full paper)

Instrumented experiments at test reactors are essential to deploying new advanced reactor systems. Designing new experiments and generating data on specific conditions require both time and cost investment. A high-fidelity model of the experiment environment can be created using finite element analysis software to support the actual experiments, but computation time is still a concern in applying outcomes to real-time usage (e.g., a digital twin). This research proposes a machine-learning-aided approach to temperature and displacement predictions, based on the thickness of the outer gas gap on the experimental capsule used for the in-pile demonstration of a novel thermal conductivity probe in the Advanced Test Reactor. The capsule consisted of U10Zr fuel, a rodlet, sodium, and inner and outer capsules. There were gas gaps between the fuel and rodlet and between the inner and outer capsule. The learning data consisted of an experimental capsule’s radial distributions of temperature and displacement, as obtained from Abaqus and the physical features. For the first step, temperature was predicted using three positional parameters. Then the displacement was predicted using six different positional parameters. Each physical feature was normalized to be both nondimensional and standardized. The temperature and displacement predictions showed good agreement in all cases involving interpolation and extrapolation. Also, data similarity enhancement increased the similarity between training and target data increasing the predictive accuracy of machine-learning models. In some cases of extrapolation, the accuracy of the machine-learning model showed limited performance, but still data similarity enhancement improved the accuracy.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Systems and methods for managing shared-path instrumentation and irradiation targets in a nuclear reactor

Systems and methods permit discriminate access to nuclear reactors. Systems provide penetration pathways to irradiation target loading and offloading systems, instrumentation systems, and other external systems at desired times, while limiting such access during undesired times. Systems use selection mechanisms that can be strategically positioned for space sharing to connect only desired systems to a reactor. Selection mechanisms include distinct paths, forks, diverters, turntables, and other types of selectors. Management methods with such systems permits use of the nuclear reactor and penetration pathways between different systems and functions, simultaneously and at only distinct desired times. Existing TIP drives and other known instrumentation and plant systems are useable with access management systems and methods, which can be used in any nuclear plant with access restrictions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Formalization of the functional analysis methodology to improve NPP I and C system design process - 244

Functional analysis is a universal methodology, which is able to support NPP and I and C design process. The paper proposes an approach to combining the functional analysis process and I and C design process into a single I and C functional design process. Operation of a system is considered as a set of functions providing not only the achievement of goal, but also keeping material and energy balances. A notation for graphical representation of semantic network (hierarchy of functions) describing a system is suggested. This network includes four levels, namely functional goals, abstract functions level, process functions level, and equipment level. The elements of the hierarchy are connected by three kinds of relationships, namely: parent-child, main-support, and cause-consequence. Then the semantic network (hierarchy) of functions is converted into a functional network which looks as a directed graph. This graph is used for development of control algorithm and for building of instrumentation and control functions hierarchy. Farther analysis of process functions reveals a set of criteria indicating real-time status and effectiveness of functions. The criteria are used for allocation of instruments and actuators. The process of functional analysis and design is illustrated in application to a simple heat generation system. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

Development of an In Situ Fission Gas Release Instrument for Fuel Sample Irradiations in the High Flux Isotope Reactor

Experimental measurement of gaseous fission product release with respect to temperature and burnup is a critical aspect of understanding nuclear fuel performance, validating predictive models, and qualifying new fuels. To measure this phenomenon in real-time, Oak Ridge National Laboratory has developed an instrument for measuring in situ fission gas release from small-scale fuel samples irradiated in the High Flux Isotope Reactor (HFIR). The instrument uses a continuous flow of Heover the fuel samples to sweep gaseous fission products from a sealed capsule in the HFIR Be reflector to an instrument station adjacent to the reactor. The instrument station houses two high-purity germanium (HPGe) detectors that measure decay gamma rays from fission products passing through a room temperature dwell chamber placed over the detector crystal. The sealed capsules in the reactor are designed to modulate fuel sample temperatures between 700 and 1,100°C by changing the Ar/He gas mixture surrounding the capsules during irradiation. N-type thermocouples are incorporated into the capsule housing to record real-time fuel temperatures. The capsules are heated primarily by prompt gamma rays emitted from the HFIR core with minimal heat contributions from fission in the fuel samples to minimize temperature gradients in the specimens for separate-effects characterization of the material. This paper describes modeling of time-dependent nuclear heating and fission product formation in fuel samples, thermal characteristics of the in-core capsules, and expected gaseous fission product gamma spectra at the HPGe instrument station.

Mulligan, Padhraic L [ORNL] (ORCID:000000025826540↗
DOE OSTI2019

Methodology for Irradiation Creep Testing of SiC / SiC Composite Cladding

Irradiation creep compliance is a critical parameter for evaluating the stress evolution of silicon carbide (SiC) components during operation in various reactor concepts. Recent instrumented irradiation creep testing of SiC, performed at low dose in the Halden reactor, showed a creep compliance that is orders of magnitude higher than previous un-instrumented bend relaxation experiments performed in the High Flux Isotope Reactor (HFIR). This discrepancy illustrates the need for additional irradiation creep data for SiC. To this end, a novel irradiation capsule design has been developed to test miniature SiC tensile specimens under loading during irradiation in the HFIR. The experiment design uses the coolant pressure of the HFIR to compress a bellows, which applies the loading on the specimen. Preliminary thermal finite element models were developed to estimate the temperature distribution within the specimen and surrounding capsule components. Similar structural models were developed to evaluate the stress distribution within the specimen. The thermal analyses show that within the gauge section, the specimen temperatures remain within ~10°C of the nominal 300°C design temperature. Similarly, the structural analysis predicts axial stresses within ~10 MPa of the target stress of 100 MPa. These initial analyses show that irradiation creep testing of SiC in the HFIR is feasible and could provide a low-cost, high-throughput vehicle for gathering needed irradiation creep data.

36 MATERIALS SCIENCE↗
DOE OSTI2013

Note: Readout of a micromechanical magnetometer for the ITER fusion reactor

We present readout instrumentation for a MEMS magnetometer, placed 30 m away from the MEMS element. This is particularly useful when sensing is performed in high-radiation environment, where the semiconductors in the readout cannot survive. High bandwidth transimpedance amplifiers are used to cancel the cable capacitances of several nanofarads. A frequency doubling readout scheme is used for crosstalk elimination. Signal-to-noise ratio in the range of 60 dB was achieved and with sub-percent nonlinearity. The presented instrument is intended for the steady-state magnetic field measurements in the ITER fusion reactor.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Selection and Ranking of Experiments from the Halden Database in support of Multiscale Model Validation

Validating fuel performance codes, such as BISON, requires an extensive amount of experiments covering a wide range of operating conditions and fuel types. Within the light-water reactor (LWR) space, there have been several international experimental programs that have contributed to the wealth of available experimental data available for use. One of those international programs, the Halden Reactor Project (HRP), began in 1958 and utilized the Halden Boiling Water Reactor to conduct many highly instrumented experiments until the reactor closed in 2018. Idaho National Laboratory, through the U.S. Department of Energy, has utilized several Halden experiments to perform the initial validation of the BISON code based upon their inclusion in international modeling and simulation benchmarks. Recently, the HRP has provided member organizations a complete copy of all available data, reports, and presentations since the HRP began. This report provides an initial exploration of the data available in the database for use in validating the multiscale models under development in the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program for LWR applications. Ranking tables that identify potential validation cases are provided for the high priority models of interest. It was found that some of the recommended high priority experiments correspond to additional rods in existing assemblies already available in the BISON validation suite. It is expected that several of these cases will be incorporated into future NEAMS milestones in the fuels technical area for increased validation of BISON.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Investigation of instrumentation containing an embedded digital device - 257

The U.S. nuclear power industry primarily relies on mature, well-proven analog technology as the basis for most instrumentation and control (I and C) equipment employed in its nuclear power plants (NPPs). However, the market for this legacy instrumentation is very small and availability of such equipment for spare parts or replacement is increasingly limited. Frequently, modern I and C equipment contains embedded digital devices (EDDs), such as microprocessors and programmable logic devices. While the enhanced functional capabilities provided by the digital components offer improved performance, reliability, and maintainability, the corresponding complexity of sophisticated digital implementations also exacerbate concerns about an increased potential for undetected systematic faults, which can lead to common-cause failure (CCF). As part of a research project sponsored by the U.S. Department of Energy, an investigation has been conducted of commercially available instrumentation marketed for nuclear or industrial application. The objective was to identify the usage of EDDs and to determine the types of functional roles allocated to the devices. Examples of the equipment surveyed include smart sensors, sequencers for emergency diesel generators, pumps, valve actuators, motor control centers, breakers, priority logic modules, and uninterruptible power sources. Where sufficient detail was found, an evaluation was performed to ascertain the nature and role of the digital technology implemented in the instrumentation. The information gained through this investigation serves as technical input for defining a classification approach for EDDs based on the functional impact of the device. For example, the role of an EDD can involve any of the following functions: performance monitoring and diagnostics, information extraction and communication, instrument condition monitoring and protection, or execution of core instrument functions. Classifying digital devices according to their role in the operation of the instrumentation in which they are embedded can enable determination of the potential functional impact of failure of that device and contribute to determination of the potential safety significance of a CCF event disabling all instances of that device. This paper will report on the findings of the investigation of instrumentation with EDDs and indicate considerations for the emerging classification approach. The higher-level goal of this effort is to determine whether a graded approach to qualification testing may be possible based on classification. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Spurious actuations in digital instrumentation and control systems - evaluation framework - 137

When a digital Instrumentation and Control (I and C) system or its associated components produce an unintended operation, it is known as a spurious actuation. Spurious actuations could lead to unnecessary challenges to safety equipment, challenge the ability of safety systems to provide their intended functions, or place the plant in an un-analyzed state. There is arguably a lack of clear and sufficient regulatory guidance for assessing spurious actuations. In an attempt to address this guidance gap, this paper provides a generic framework for evaluating spurious actuations in digital I and C systems, components, or supporting systems that are important to safety. The framework provides a methodology for: (1) defining the scope of the evaluation including supporting assumptions; (2) providing options for excluding a spurious actuation from the evaluation; (3) assessing the potential consequences from the assessed spurious actuations; and, (4) defining a high-level acceptance criteria. The methodology discussed by this paper is not to be construed as a requirement, regulation, or acceptable guidance by either domestic or international regulators; instead, it is intended to serve as a potential foundation or technical basis to be used for developing clear and sufficient regulatory guidance for assessing spurious actuations in digital I and C systems. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Impacts of common cause failure regulatory requirements on protection system architectures - 304

Regulatory-based design considerations for potential common cause failures (CCF) in digital instrumentation and control (I and C) systems have evolved over time and have impacted the development of protection system I and C architectures. The single failure criterion used to be the main design driver for a protection system I and C architecture and resulted in the familiar four-fold and three-fold system architectures. The single failure criterion has now been replaced by the conservative assumptions associated with digital CCF concerns (i.e., assumed digital CCF coincident with an anticipated operational transient or postulated accident) which are classified as 'beyond design basis events' by regulatory authorities. It is unusual for a beyond design basis event to be the controlling factor for a safety system design but that is now the reality for I and C system designers. The preferred regulatory solution is to introduce diversity into I and C system designs to guard against the general concern of digital CCF. However, the lack of clear criteria on how to define sufficient diversity (i.e., how much, what kind, and whether it is linked to identified digital CCF vulnerabilities) had led to more complex I and C architectures. The trend has been towards lengthy and more difficult reviews for the treatment of digital CCF vulnerabilities and I and C system architectures because of the subjective definition of digital CCF vulnerabilities and the lack of clear acceptance criteria for diversity strategy. Improvement in the treatment of digital CCF is needed to reverse the trend of increased I and C system architecture complexity and longer regulatory reviews. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Application of a bayesian belief network model to reliability assessment of nuclear safety-related software - 207

As the instrumentation and control (I and C) systems in nuclear power plants (NPPs) have been replaced with digital-based systems, the need to incorporate software failures into NPP probabilistic risk assessments has arisen. In order to assess the probability of software failure on demand, a Bayesian belief network (BBN) model was developed which estimates the number of defects and the resulting probability of software failure on demand in nuclear safety-related software. To assess the feasibility of the BBN framework, the BBN model was applied to the prototype Integrated Digital Protection System-Reactor Protection System (IDiPS-RPS) to estimate the number of remaining faults and the software failure probability of a target software. The developmental- and V and V-activities carried out during the IDiPS-RPS development process were evaluated based on the well-defined checklist derived by the V and V team and were estimated based on expert elicitation. In addition, the attribute evaluations and the number of FPs of the target software is provided as the inputs for the BBN model. The application results showed the feasibility of using BBNs for quantifying software failure probabilities and several insights were gained from the applications of the BBN model. The proposed BBN framework can be applied to estimate the software failure probability for other safety-related NPP software and provide an insight on modeling the software development process that involves iterations between different development phases. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

IEC 62859: A new international standard on the coordination between safety and cybersecurity for nuclear I and C systems - 160

Cybersecurity requirements and controls for digital I and C systems are increasingly added to well-established safety-oriented provisions. Interactions and potential side-effects are possible and must be mastered, taking into account the nuclear I and C system specificities as well as their normative and regulatory contexts. The new international standard IEC 62859 'Nuclear power plants - Instrumentation and control systems - Requirements for coordinating safety and cyber security' aims to optimize the integration of cybersecurity provisions in nuclear I and C architecture and systems, to prevent conflicts between safety and cybersecurity provisions, and to aid the identification and the leveraging of the potential synergies between safety and cybersecurity. The paper presents the structure and the key aspects of this standard. IEC 62859 is structured along three main sections: one dealing with the overall I and C architecture, a second one focusing on the system level, a third one dealing with organizational and procedural issues. The paper also discusses related work in the nuclear industry, and in other sectors. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗