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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 775 records · Page 43
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Assessment of NDE for Key Indicators of Aging Cables in Nuclear Power Plants – Interim Status

Degradation of the cable jacket, electrical insulation, and other cable components of installed cables within nuclear power plants (NPPs) is known to occur as a function of age, temperature, radiation, and other environmental factors. System tests verify cable function under normal loads; however, the concern is over cable performance under exceptional loads associated with design-basis events (DBEs). The cable’s ability to perform safely over the initial 40 year planned and licensed life has generally been demonstrated and there have been very few age-related cable failures. With greater than 1000 km of power, control, instrumentation, and other cables typically found in an NPP, replacing all the cables would be a severe cost burden. Justification for life extension to 60 and 80 years requires a cable aging management program to justify cable performance under normal operation as well as accident conditions. Currently the gold standard for determining cable insulation degradation is the elongation-at-break (EAB). This, however, is an ex-situ measurement and requires removal of a sample for laboratory investigation. A reliable nondestructive examination (NDE) in-situ approach is desirable to objectively determine the suitability of the cable for service. A variety of tests are available to assess various aspects of electrical and mechanical cable performance, but none of these tests are suitable for all cable configurations nor does any single test confirm all features of interest. Nevertheless, the complete collection of test possibilities offers a powerful range of tools to assure the integrity of critical cables. Licensees and regulators have settled on a practical program to justify continued operation based on condition monitoring of a lead sample set of cables where test data is tracked in a database and the required test data are continually adjusted based on plant and fleet-wide experience. As part of the Light Water Reactor Sustainability program sponsored by the U.S. Nuclear Regulatory Commission, the U.S. Department of Energy, and industry (represented by the Electric Power Research Institute), an assessment of cable NDE methods was commissioned. Technologies include both bulk electrical measurements (Tan d, time domain reflectometry, frequency domain reflectometry (FDR), partial discharge, and other techniques) and local insulation measurement (indenter, dynamic mechanical analysis interdigital capacitance, infrared spectral measurement, etc.). This aging cable NDE program update reviews the full range of techniques but focuses on the most interesting test approaches that have a chance to be deployed in-situ, particularly including Tan d, FDR, and ultrasound methods that have been reviewed most completely in this progress period.

cable aging inspection↗
DOE OSTI2016

Investigation of similarity metrics for simulation based scaling analysis

Scaling analysis is a technique for characterizing degree of similarity between processes to occur in a real prototypic system and those to occur in its reduced scale (experimental) analog. In the context of nuclear reactor thermal-hydraulics and safety, scaling analysis is instrumental in assessing the applicability of data obtained from a scaled test facility for calibration of models and validation of simulation codes. Quantification of scale-related distortions is the key element in scaling analysis. In this work, we examine an approach for scaling analysis using computer simulation codes. Similarity analysis using information-theoretic approach is performed by comparing performance of experiment and application to variation of system conditions. Three different information theoretic measures namely mutual information, Kullback-Leibler divergence and overlap coefficient are selected for their potential use as metric for similarity analysis. We test the metrics by comparing them against physics-based scaling ratios, which are well known in fundamental thermal-hydraulic processes. In this exploratory analysis, a case of pressure drop of a steady-state flow through a channel is employed to test the proposed approach. It was found that in the absence of model uncertainties, mutual information serve as an effective measure of the scale effects present in the experiment. However, model uncertainty can significantly impact the quality of similarity analysis. Our analysis indicates that model uncertainties can have significant impact on the information theoretic metrics used for scaling analysis. Furthermore, if the model uncertainties exceeds beyond a critical value, the scale effect present in the experiment cannot be adequately evaluated. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2021

A Discussion of Strength Reduction Factor Development for Thermal Aging Effect on Nuclear Structural Alloys

In consideration of structural alloy property deterioration during long-term exposure to elevated temperatures, the yield and ultimate tensile strength reduction factors are provided in the Boiler and Pressure Vessel Code Section III for nuclear reactor component design and operation analysis. Because the Gen IV reactor requirement of 40 ~ 60 years of service life makes it difficult to acquire such long exposure test data for developing the reduction factors, they must be derived from test data with relatively short exposure by predictive methods considered to be reasonably reliable. A novel approach with a physically-based model has recently been proposed for application to development of reduction factors for 9Cr-1Mo-V. In the model, contributors to the tensile strength are first identified and related to definite microstructural features of the alloy, then some physically-based methods are employed to simulate the microstructural evolution, and finally the model is assembled with test-data-calibrated parameters to generate the yield and ultimate tensile strength reduction factors covering elevated temperature exposure for up to 57 years. The approach is undoubtedly a trailblazing development that will, if proven reliable, lead to a paradigm shift in predicting thermal aging behavior of many other alloys. Its debut application to Section III, however, concerns nuclear safety and naturally warrants objective, impartial, and thorough technical scrutiny. In the present paper, the novel and conventional approaches are discussed. Necessary improvements to the novel approach are recommended for its application to nuclear structural component design and analysis, and for its potential expanded use to other alloys.

Ren, Weiju↗
DOE OSTI2021

Plans for Expanded Data Testing at ORNL [Slides]

VALID expansion is on-going and should provide coverage over a wider range of systems in the future. Adding additional application types beyond critical experiments can also be used in expanded data testing. Two other test suites have been developed at ORNL for other purposes, but never waste a good test suite. These suites are more likely to be relevant to testing differences in data as compared to assessing accuracy of evaluations.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Distributed Temperature Sensing within a 19-pin Wire-Wrapped Fuel Bundle Model

It has always been difficult to measure local temperature and flow velocity within fuel bundle geometries because of their complex and compact layouts. High-resolution mapping of the flow field is especially challenging. High-resolution data is typically obtained using optical instruments but optical access to the interior regions of a bundle is severely limited. Velocity fields within fuel bundle models have been mapped with particle image velocimetry (PIV), an optical technique that employs lasers and cameras. Optical access to the bundle interior is obtained using transparent materials and matching the refractive index of the physical structures with that of the fluid. Testing is necessarily isothermal because refractive indices vary with temperature. However, the ultimate aim of such studies is understanding fluid dynamics and heat transfer within real reactors and so flow field mapping under nonisothermal conditions is of particular interest. Temperature fields can be mapped at high-resolution using laser-induced fluorescence (LIF). However, as an optical technique it is unsuitable for nonisothermal liquid flows. This is due to the same sort of optical distortion that restricts PIV. It is also unsuitable for opaque fluids and setups with restricted optical access. Fiber optic distributed temperature sensing (DTS) offers a high-resolution alternative that does not rely upon imaging. A single sensor can map temperature fields at resolutions far exceeding that achievable with conventional sensors such as thermocouples. Though fiber optic sensors cannot match the spatial resolution and bandwidth of LIF, they are suitable for nonisothermal and opaque flows. And though fibers are physical probes that perturb the flow, they can be strung across complex geometries to provide data in regions lacking the optical access necessary for techniques like LIF. Distributed temperature sensing has the potential to provide fuel bundle temperature data of a type that cannot be obtained with optical methods or conventional sensors. This paper reports on preliminary results of efforts to use fiber optic distributed temperature sensors in a fuel bundle model. We are conducting a feasibility study exploring implementation issues such as strain control in the mounting configuration and signal noise associated with flow-induced vibration. We describe here the setup and present preliminary data from our first phase testing with stagnant flow. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2010

Conceptual Design of Forced Convection Molten Salt Heat Transfer Testing Loop

This report develops a proposal to design and construct a forced convection test loop. A detailed test plan will then be conducted to obtain data on heat transfer, thermodynamic, and corrosion characteristics of the molten salts and fluid-solid interaction. In particular, this report outlines an experimental research and development test plan. The most important initial requirement for heat transfer test of molten salt systems is the establishment of reference coolant materials to use in the experiments. An earlier report produced within the same project highlighted how thermophysical properties of the materials that directly impact the heat transfer behavior are strongly correlated to the composition and impurities concentration of the melt. It is therefore essential to establish laboratory techniques that can measure the melt composition, and to develop purification methods that would allow the production of large quantities of coolant with the desired purity. A companion report describes the options available to reach such objectives. In particular, that report outlines an experimental research and development test plan that would include following steps: •Molten Salts: The candidate molten salts for investigation will be selected. •Materials of Construction: Materials of construction for the test loop, heat exchangers, and fluid-solid corrosion tests in the test loop will also be selected. •Scaling Analysis: Scaling analysis to design the test loop will be performed. •Test Plan: A comprehensive test plan to include all the tests that are being planned in the short and long term time frame will be developed. •Design the Test Loop: The forced convection test loop will be designed including extensive mechanical design, instrument selection, data acquisition system, safety requirements, and related precautionary measures. •Fabricate the Test Loop. •Perform the Tests. •Uncertainty Analysis: As a part of the data collection, uncertainty analysis will be performed to develop probability of confidence in what is measured in the test loop. Overall, the testing loop will allow development of needed heat transfer related thermophysical parameters for all the salts, validate existing correlations, validate measuring instruments under harsh environment, and have extensive corrosion testing of materials of construction.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2012

Piezoelectric material for use in a nuclear reactor core

In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d{sub 33} was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d{sub 33} for many as-grown samples.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Hyperfine structure of the ground state in muonic-lithium ions

Small intervals of the hyperfine structure of the ground state in the muonic-lithium ions (μe{sub 3}{sup 6,7}Li){sup +} were calculated by perturbation theory in the fine-structure constant and in the electronto- muon mass ratio. Vacuum-polarization, recoil, and nuclear-structure effects and electron vertex corrections were taken into account. The values obtained for the small hyperfine-splitting intervals can be used in a comparison with future experimental data and in tests of quantum electrodynamics.

74 ATOMIC AND MOLECULAR PHYSICS↗
DOE OSTI2018

Advanced Sensors and Instrumentation Issue 10

The Advanced Sensors and Instrumentation (ASI) program, on its seventh fiscal year, continues to address key technology gaps for future nuclear energy systems. ASI’s goal is to enable the advanced instrumentation and control (I&C) technologies essential to the Office of Nuclear Energy’s research and development (R&D) efforts needed to realize its mission goals. Information on our projects can be found in our ASI annual awards summaries, annual review webinars, and newsletters posted in the Office of Nuclear Energy website (energy.gov/ne). This year, we have initiated three newly awarded projects in the area of data analytics. We also initiated a project to develop and demonstrate a prototype passive wireless sensor network. This sensor will use advanced manufacturing and printing, and be tested and demonstrated at a U.S. nuclear power plant facility. In addition, we are developing a sensor database to capture advanced sensor and instrumentation technology that is applicable to nuclear power plants. Under the In-Pile Instrumentation program, we continue to develop unique in-pile sensors and instrumentation to provide real-time, accurate, and spatially resolved information regarding the performance of fuels and materials. Due to the closure of the Halden reactor, we are focusing our near-term research in providing equivalent capability for the instrumentation of irradiation experiments in the U.S. that are needed by the nuclear industry. As the ASI program continues to award projects through competitive solicitations, we encourage the I&C community to visit our website for current funding opportunity announcements, especially for the industry-focused solicitation entitled, “U.S. Industry Opportunities for Advanced Nuclear Technology Development.”

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Polyethylene Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 19, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc depositing energy in a Si solid state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2017

A NOVEL TOOL FOR IMPROVING THE DATA COLLECTION PROCESS DURING CONTROL ROOM MODERNIZATION HUMAN-SYSTEM INTERFACE TESTING AND EVALUATION ACTIVITIES

The U.S. Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) Program is developing a scientific basis to extend the existing U.S. nuclear power plant (NPP) operating life beyond the current 60-year licensing period and to ensure NPP long-term reliability, productivity, safety, and security. Under the Advanced Instrumentation, Information, and Control (II&C) Systems Technologies pathway, the LWRS Program conducts targeted research and development to address long-term aging and obsolescence of existing instrumentation and control technologies, as well as to implement and test new digital II&C technology that enables broad innovation and business improvement in the NPP operating model. To this end, the role of human factors engineering (HFE) is important as part of ensuring that the control room modernization (CRM) activities account for the operator’s cognitive and physiological characteristics. Part of the HFE process for CRM entails conducting tests and evaluations (T&Es) that will inform the design of new human-system interface (HSI) technologies. Qualitative data that captures operator feedback and important human-system interactions are an important part of T&E. However, the analyzing this qualitative data in order to provide actionable design recommendations can be time-consuming and error prone. This paper presents an early version of a tool that can be used to analyze this qualitative data to (1) sift through freeform text for improved design issue identification, (2) maintain consistency in theme coding conventions, and (3) readily provide actionable design guidance that can be traceable to state-of-the-art HFE guidelines such as NUREG-0700. A description of this tool is discussed next.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Neutron Activation Foil and Thermoluminescent Dosimeter Responses to a Lead Reflected Pulse of the CEA Valduc SILENE Critical Assembly

This benchmark experiment was conducted as a joint venture between the US Department of Energy (DOE) and the French Commissariat à l'Energie Atomique (CEA). Staff at the Oak Ridge National Laboratory (ORNL) in the US and the Centre de Valduc in France planned this experiment. The experiment was conducted on October 13, 2010 in the SILENE critical assembly facility at Valduc. Several other organizations contributed to this experiment and the subsequent evaluation, including CEA Saclay, Lawrence Livermore National Laboratory (LLNL), the Y-12 National Security Complex (NSC), Babcock International Group in the United Kingdom, and Los Alamos National Laboratory (LANL). The goal of this experiment was to measure neutron activation and thermoluminescent dosimeter (TLD) doses from a source similar to a fissile solution critical excursion. The resulting benchmark can be used for validation of computer codes and nuclear data libraries as required when performing analysis of criticality accident alarm systems (CAASs). A secondary goal of this experiment was to qualitatively test performance of two CAAS detectors similar to those currently and formerly in use in some US DOE facilities. The detectors tested were the CIDAS MkX and the Rocky Flats NCD-91. The CIDAS detects gammas with a Geiger-Muller tube, and the Rocky Flats detects neutrons via charged particles produced in a thin 6 LiF disc, depositing energy in a Si solid-state detector. These detectors were being evaluated to determine whether they would alarm, so they were not expected to generate benchmark quality data.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Sister Rod Destructive Examinations (FY23) Appendix E2: Four-Point Bend Fractured Specimen Images

This report documents work performed under the Spent Fuel and Waste Disposition’s Spent Fuel and Waste Science and Technology program for the US Department of Energy (DOE) Office of Nuclear Energy (NE). This work was performed to fulfill Level 2 Milestone M2SF-24OR010201024, “FY23 ORNL Testing on Sibling Pins,” within work package SF-24OR01020102 and is an update to the work reported in M2SF-23OR010201024, M2SF-22OR010201047, M2SF-21OR010201032, M2SF-19ORO010201026, and M2SF-19OR010201028. As a part of the DOE NE High Burnup Spent Fuel Data Project, Oak Ridge National Laboratory (ORNL) is performing destructive examinations (DEs) of high burnup (HBU) (>45 GWd/MTU) spent nuclear fuel (SNF) rods from the North Anna Nuclear Power Station operated by Dominion Energy. The SNF rods, called sister rods or sibling rods, are all HBU and include four different kinds of fuel rod cladding: standard Zircaloy-4 (Zirc-4), low-tin Zirc-4, ZIRLO, and M5. The DEs are being conducted to obtain a baseline of the HBU rods’ condition before they are sent to dry storage and are focused on gaining an understanding of overall SNF rod strength and durability. Composite fuel and defueled cladding will be tested to derive material properties. Although the data generated can be used for multiple purposes, one primary goal for obtaining the post-irradiation examination data and the associated measured mechanical properties is to support SNF dry storage licensing and relicensing activities by (1) addressing identified knowledge gaps and (2) enhancing the technical basis for post-storage transportation, handling, and subsequent disposition. This report documents the status of the ORNL Phase 1 DE activities related to the mechanical testing of selected sister rods in Phase 1 of the sister rod test program. This appendix provides images of the segments broken in four-point bending (4PB).

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

R.Q. Wright’s Hf updates [Slides]

This presentation details R.Q. Wright’s Hf updates. It discusses the stable Hf isotope (174, 176-180), RRR and URR + fast (n, g), and the motivation and deficiencies in the current ENDF/B-VIII.0. It also discusses what new data and/or theory will motivate a new evaluation and/or update and what validation testing has been/will be done.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2012

Enhanced In-pile Instrumentation for Material Testing Reactors

An increasing number of U.S. nuclear research programs are requesting enhanced in-pile instrumentation capable of providing real-time measurements of key parameters during irradiations. For example, fuel research and development funded by the U.S. Department of Energy now emphasize approaches that rely on first principle models to develop optimized fuel designs that offer significant improvements over current fuels. To facilitate this approach, high fidelity, real-time data are essential for characterizing the performance of new fuels during irradiation testing. Furthermore, sensors that obtain such data must be miniature, reliable and able to withstand high flux/high temperature conditions. Depending on user requirements, sensors may need to obtain data in inert gas, pressurized water, or liquid metal environments. To address these user needs, in-pile instrumentation development efforts have been initiated as part of the Advanced Test Reactor (ATR) National Scientific User Facility (NSUF), the Fuel Cycle Research & Development (FCR&D), and the Nuclear Energy Enabling Technology (NEET) programs. This paper reports on recent INL achievements to support these programs. Specifically, an overview of the types of sensors currently available to support in-pile irradiations and those sensors currently available to MTR users are identified. In addition, recent results and products available from sensor research and development are detailed. Specifically, progress in deploying enhanced in-pile sensors for detecting elongation and thermal conductivity are reported. Results from research to evaluate the viability of ultrasonic and fiber optic technologies for irradiation testing are also summarized.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

Detection with Enhanced Energy Windowing Phase I Report

This document reviews the progress of Phase I of the Detection with Enhanced Energy Windowing (DEEW) project. The DEEW project is the implementation of software incorporating an algorithm which reviews data generated by radiation portal monitors and utilizes advanced and novel techniques for detecting radiological and fissile material while not alarming on Naturally Occurring Radioactive Material. Independent testing indicated that the Enhanced Energy Windowing algorithm showed promise at reducing the probability of alarm in the stream of commerce compared to existing algorithms and other developmental algorithms, while still maintaining adequate sensitivity to threats. This document contains a brief description of the project, instructions for setting up and running the applications, and guidance to help make reviewing the output files and source code easier.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2021

Development of Real-Time, In-pile Creep Test Rigs for Characterizing the Structural Materials of Nuclear Components

New and improved materials are being considered for supporting both existing and next-generation nuclear reactors. Reactor materials can significantly degrade with time, thus limiting or altering their properties in harsh reactor environments. To accurately understand such material degradation, real-time data obtained under prototypic irradiation conditions are required. In particular, understanding the creep behavior of materials exposed to irradiation and elevated temperatures is essential for safety concern evaluations. To provide these capabilities, Idaho National Laboratory (INL)’s High Temperature Test Laboratory (HTTL) developed several instrumented test rigs for obtaining real-time data from specimens in well-controlled pressurized-water reactor (PWR) coolant conditions at the Materials Test Reactor. This technical report focuses on INL’s efforts to evaluate and enhance the former creep test rig prototype that relied on linear variable differential transformers in laboratory settings. Specifically, the test rig can detect changes in the length of creep specimens, which is useful for measuring thermal expansion and creep deformation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2010

Human System Simulation in Support of Human Performance Technical Basis at NPPs

This paper focuses on strategies and progress toward establishing the Idaho National Laboratory’s (INL’s) Human Systems Simulator Laboratory at the Center for Advanced Energy Studies (CAES), a consortium of Idaho State Universities. The INL is one of the National Laboratories of the US Department of Energy. One of the first planned applications for the Human Systems Simulator Laboratory is implementation of a dynamic nuclear power plant simulation (NPP) where studies of operator workload, situation awareness, performance and preference will be carried out in simulated control rooms including nuclear power plant control rooms. Simulation offers a means by which to review operational concepts, improve design practices and provide a technical basis for licensing decisions. In preparation for the next generation power plant and current government and industry efforts in support of light water reactor sustainability, human operators will be attached to a suite of physiological measurement instruments and, in combination with traditional Human Factors Measurement techniques, carry out control room tasks in simulated advanced digital and hybrid analog/digital control rooms. The current focus of the Human Systems Simulator Laboratory is building core competence in quantitative and qualitative measurements of situation awareness and workload. Of particular interest is whether introduction of digital systems including automated procedures has the potential to reduce workload and enhance safety while improving situation awareness or whether workload is merely shifted and situation awareness is modified in yet to be determined ways. Data analysis is carried out by engineers and scientists and includes measures of the physical and neurological correlates of human performance. The current approach supports a user-centered design philosophy (see ISO 13407 “Human Centered Design Process for Interactive Systems, 1999) wherein the context for task performance along with the requirements of the end-user are taken into account during the design process and the validity of design is determined through testing of real end users

22 GENERAL STUDIES OF NUCLEAR REACTORS↗