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At least 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Pre-Transient Characterization of Historic EBR-II Pins for Transient Testing

Current interest in sodium-cooled fast reactor (SFR) designs, such as TerraPower’s Natrium Reactor, has highlighted the need for advanced reactor fuel technology development. Modern U-Zr and U- Pu-Zr pin designs are primary candidates to fuel SFRs and boast high fuel utilization capacity, increased fuel-cladding compatibility, and improved safety through inherent feedback mechanisms. Despite over 60 years of metallic fuel irradiation, uncertainties exist in the performance of the fuel system, particularly under transient overpower (TOP) and loss of flow (LOF) scenarios. Throughout historical testing within the Experimental Breeder Reactor II (EBR-II) and the Fast Flux Test Facility (FFTF), fuel behavior has demonstrated benign response to transient reactor conditions; however, accurate predictions of failure thresholds to inform operational limitations rely heavily on fuel composition, burnup, and irradiation history. In expanding TOP and LOF testing, the Transient Heat sink Overpower Response (THOR) Capsule will be used to test modern fuel technologies in a static sodium environment in the Transient Reactor Test (TREAT) Facility. The THOR capsule is highly instrumented and will provide time-dependent thermal behavior of SFR fuel pins subjected to accident conditions within TREAT. The THOR-Metallic (THOR- M) campaign aims to validate and expand historical TOP and LOF testing on high burnup U-Zr and U-Pu- Zr fuel alloys previously irradiated in EBR-II by running the rods to failure. This contribution focuses primarily on the pre-transient engineering-scale destructive and non- destructive characterization that has been conducted on both the test and sibling pins used for the TOP and LOF tests. All pins underwent visual examination, neutron radiography, element contact profilometry, and precise gamma scan. The sibling pins used for each test were further analyzed using gas assay, sampling, and recharge analysis (GASR), and optical microscopy. The results from each technique confirmed that the fuel pins were intact and devoid of any atypical developments when compared to historical data. Additionally, the analyzed measurements establish a baseline for comparison to post-transient analysis. Key fuel behaviors quanitifed include axial elongation of the fuel column, diametral strain of the pin, patterns in fluff structure geometry, changes in axial isotope distribution, evolution of constituent redistribution, porosity, and fission gas release. The pre-transient measurements and changes attributed to transient behavior from post-transient measurement will be compared to historical data to capture the behavioral dependence on composition, burnup, and irradiation history. Results from this work advance the initiatives of the THOR-M campaign, which aid in informing fuel performance models and establishing safety criteria for SFR operational limits. The novel combination of test environment, in-situ instrumentation, and comprehensive suite of characterization methods provides greater understanding of transient fuel behavior. Overall, information on the time and condition of pin failure for high burnup U-Pu-Zr will greatly expand the limited existing TOP and LOF test data.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2011

Neutron Instruments Added at Oak Ridge

The neutron scattering facilities at Oak Ridge National Laboratory continue their development as new instruments are commissioned and join the user program at the Spallation Neutron Source and High Flux Isotope Reactor. More than 640 proposals were received for beam time during the January-May 2011 period on SNS and HFIR instruments with about half either being accepted or identified as alternates. The proposal call for the period June-December 2011, announced at http://neutrons.ornl.gov, will close February 23, 2011.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Level monitoring system with pulsating sensor—Application to online level monitoring of dashpots in a fast breeder reactor

An innovative continuous type liquid level monitoring system constructed by using a new class of sensor, viz., pulsating sensor, is presented. This device is of industrial grade and it is exclusively used for level monitoring of any non conducting liquid. This instrument of unique design is suitable for high resolution online monitoring of oil level in dashpots of a sodium-cooled fast breeder reactor. The sensing probe is of capacitance type robust probe consisting of a number of rectangular mirror polished stainless steel (SS-304) plates separated with uniform gaps. The performance of this novel instrument has been thoroughly investigated. The precision, sensitivity, response time, and the lowest detection limit in measurement using this device are <0.01 mm, ∼100 Hz/mm, ∼1 s, and ∼0.03 mm, respectively. The influence of temperature on liquid level is studied and the temperature compensation is provided in the instrument. The instrument qualified all recommended tests, such as environmental, electromagnetic interference and electromagnetic compatibility, and seismic tests prior to its deployment in nuclear reactor. With the evolution of this level measurement approach, it is possible to provide dashpot oil level sensors in fast breeder reactor for the first time for continuous measurement of oil level in dashpots of Control and Safety Rod Drive Mechanism during reactor operation.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Microstructural characterization of the oxidation of nuclear graphite under chronic and accident conditions via XCT and synchrotron XCT

Graphite is a candidate material to moderate fast neutrons and for structural components in the US next-generation graphite-moderated reactors. A graphite core is conceived as a large formation of interconnected bricks that primarily serves as the moderator of fast neutrons, holds instrumentation, fuel elements, control rods, and is a receptacle for molten salts for Molten Salt Reactors (MSRs) designs. During the operation of a graphite-moderated power plant, graphite components might be subjected to chronic oxidation during normal operating conditions or aggressive oxidation as a result of accidental ingress of air that reacts vigorously with the graphite core. Using synchrotron, x-ray computed tomography (XCT), this research systematically characterized microstructural changes that accompanied these two oxidation scenarios. Chronic oxidation was studied by characterizing IG-110, PCEA, and NBG-18 specimens that were gradually oxidized in air at a low temperature (520°C). The accidental ingress of air into the graphite reactor was simulated by subjecting the grades mentioned above to high-temperature oxidation at approximately 750°C. This research is the first, in situ, systematic characterization of nuclear graphite microstructural evolution that can be associated with the two possible oxidation scenarios and provide insight into related repercussions. The results show that the microstructure and pore connectivity influence the rate of oxidation and evolution of the microstructure under the two oxidation regimes. These results are crucial to understanding which grades are more resilient to each type of oxidation and documenting the damage created in the graphite.

Arregui Mena, Jose'↗
DOE OSTI2017

MODELING THE AMBIENT CONDITION EFFECTS OF AN AIR-COOLED NATURAL CIRCULATION SYSTEM

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). One such variant, air-cooled RCCS, uses natural convection to drive the flow of air from outside the reactor building to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a half-scale model of the primary features of one conceptual air-cooled RCCS design. The facility was constructed to carry out highly instrumented experiments to study the performance of the RCCS concept for reactor decay heat removal that relies on natural convection cooling. Parallel modeling and simulation efforts were performed to support the design, operation, and analysis of the natural convection system. Throughout the testing program, strong influences of ambient conditions were observed in the experimental data when baseline tests were repeated under the same test procedures. Thus, significant analysis efforts were devoted to gaining a better understanding of these influences and the subsequent response of the NSTF to ambient conditions. It was determined that air humidity had negligible impacts on NSTF system performance and therefore did not warrant consideration in the models. However, temperature differences between the building exterior and interior air, along with the outside wind speed, were shown to be dominant factors. Combining the stack and wind effects together, an empirical model was developed based on theoretical considerations and using experimental data to correlate zero-power system flow rates with ambient meteorological conditions. Some coefficients in the model were obtained based on best fitting the experimental data. The predictive capability of the empirical model was demonstrated by applying it to the new set of experimental data. The empirical model was also implemented in the computational models of the NSTF using both RELAP5-3D and STARCCM+ codes. Accounting for the effects of ambient conditions, simulations from both codes predicted the natural circulation flow rates very well.

Hu, Rui↗
DOE OSTI2019

Status Report: Design and Fabricate Installation-Ready Test Capsule for the In-Situ Measurement of Elastic Properties Using an Optical Fiber-Based Measurement Technique

Microstructure evolution due to irradiation in a nuclear reactor can have a dramatic effect on material properties. A better understanding of this evolution is necessary for developing improved nuclear fuels and materials. The ability to measure such changes in real time is extremely limited due to the harsh conditions, high radiation fields, and limited access of the reactor environment. Through carefully designed experiments, measurement of elastic properties can be tied directly to microstructure. We present the methodology, design, and deployment plan for an instrument that has been developed to monitor grain microstructural changes during irradiation. Our measurement approach involves exciting and measuring the resonant frequency of a thin cantilever beam. Excitation and detection of the flexural vibrations of the beam are accomplished using optical methods that require only an optical fiber connection between the instrumentation and the sample. This technique has been demonstrated in a laboratory setting to monitor the recrystallization of highly textured copper during high temperature annealing. A test capsule incorporating this technique has been developed for in reactor testing. The capsule has been designed to be compatible with a reusable test module that allows simplified insertion in the TREAT reactor at INL. An exact replica of the instrument described in this Status Report, constructed from Quality Level 3 materials, is slated to be inserted into the TREAT reactor in FY19.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

MARVEL Instrumentation and Control System (ICS)

MARVEL plans to be the first new operational reactor at INL in many years. These documents are being made publicly available to disseminate key information on the MARVEL reactor for private sector use towards reactor development.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
NASA NTRS1974

Trace-element analysis of 1000 environmental samples per year using instrumental neutron activation analysis

The technology and methods developed at the Plum Brook Reactor to analyze 1000 samples per year and report data on as many as 56 elements are described. The manpower for the complete analysis of 20 to 24 samples per week required only 3 to 3.5 hours per sample. The solutions to problems encountered in sample preparation, irradiation, and counting are discussed. The automation of data reduction is described. Typical data on various sample matrices are presented.

Sheibley, D. W.↗
DOE OSTI2014

Light Water Reactor Sustainability Program: Digital Technology Business Case Methodology Guide

The Department of Energy’s (DOE’s) Light Water Reactor Sustainability Program aims to develop and deploy technologies that will make the existing U.S. nuclear fleet more efficient and competitive. The program has developed a standard methodology for determining the impact of new technologies in order to assist nuclear power plant (NPP) operators in building sound business cases. The Advanced Instrumentation, Information, and Control (II&C) Systems Technologies Pathway is part of the DOE’s Light Water Reactor Sustainability (LWRS) Program. It conducts targeted research and development (R&D) to address aging and reliability concerns with the legacy instrumentation and control and related information systems of the U.S. operating light water reactor (LWR) fleet. This work involves two major goals: (1) to ensure that legacy analog II&C systems are not life-limiting issues for the LWR fleet and (2) to implement digital II&C technology in a manner that enables broad innovation and business improvement in the NPP operating model. Resolving long-term operational concerns with the II&C systems contributes to the long-term sustainability of the LWR fleet, which is vital to the nation’s energy and environmental security. The II&C Pathway is conducting a series of pilot projects that enable the development and deployment of new II&C technologies in existing nuclear plants. Through the LWRS program, individual utilities and plants are able to participate in these projects or otherwise leverage the results of projects conducted at demonstration plants. Performance advantages of the new pilot project technologies are widely acknowledged, but it has proven difficult for utilities to derive business cases for justifying investment in these new capabilities. Lack of a business case is often cited by utilities as a barrier to pursuing wide-scale application of digital technologies to nuclear plant work activities. The decision to move forward with funding usually hinges on demonstrating actual cost reductions that can be credited to budgets and thereby truly reduce O&M or capital costs. Technology enhancements, while enhancing work methods and making work more efficient, often fail to eliminate workload such that it changes overall staffing and material cost requirements. It is critical to demonstrate cost reductions or impacts on non-cost performance objectives in order for the business case to justify investment by nuclear operators. The Business Case Methodology (BCM) addresses the “benefit” side of the analysis—as opposed to the cost side—and how the organization evaluates discretionary projects (net present value (NPV), accounting effects of taxes, discount rates, etc.). The cost and analysis side is not particularly difficult for the organization and can usually be determined with a fair amount of precision (not withstanding implementation project cost overruns). It is in determining the "benefits" side of the analysis that utilities have more difficulty in technology projects and that is the focus of this methodology.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Estimation of sensor measurement errors in reactor coolant systems using multi-sensor fusion

A nuclear power plant is typically instrumented with a variety of sensors to continually monitor its variables, and their sensor’s measurements may be used to assess the plant state and initiate safety actions, if needed. Errors in sensor measurements, due to factors such as calibration drifts, critically affect such state assessments. Here, we address a problem of estimating sensor errors using physics-informed machine learning methods that use measurements collected under known plant conditions. For a given sensor, we propose an information fusion method that uses measurements from other sensors to estimate its output assuming it is error-free and provides its difference from an actual measurement as an error estimate. We present the ensemble of trees and support vector machine fusers, and evaluate their performance using measurements collected over an emulated test loop of a pressurized water reactor. The plant variables are related to each other through the underlying physical laws under inertial constraints that place bounds on their derivatives, which analytically justify the applicability of machine learning methods for computing these fusers. Under twenty scenarios, we assess their sensor error estimates for pressure sensors of the heat exchanger of a reactor’s primary coolant system. Multiple types of errors are captured by both fusers under externally induced calibration drifts, blockages, minor leaks and air gaps in sensing lines, and electromagnetic interference; the root mean square error of the estimation of error is under 2.2% percent of the maximum measurement. We present generalization equations, in the framework of statistical learning theory, for these methods that characterize the confidence probability that the estimation error is bounded by a specified parameter in future test scenarios.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

The suite of small-angle neutron scattering instruments at Oak Ridge National Laboratory

Oak Ridge National Laboratory is home to the High Flux Isotope Reactor (HFIR), a high-flux research reactor, and the Spallation Neutron Source (SNS), the world's most intense source of pulsed neutron beams. The unique co-localization of these two sources provided an opportunity to develop a suite of complementary small-angle neutron scattering instruments for studies of large-scale structures: the GP-SANS and Bio-SANS instruments at the HFIR and the EQ-SANS and TOF-USANS instruments at the SNS. This article provides an overview of the capabilities of the suite of instruments, with specific emphasis on how they complement each other. As a result, a description of the plans for future developments including greater integration of the suite into a single point of entry for neutron scattering studies of large-scale structures is also provided.

47 OTHER INSTRUMENTATION↗
DOE OSTI2021

Versatile semi-batch apparatus for manometric measurement of gas–solid reaction rates

In this study, we describe the design, construction, operation, and performance of a simple and versatile semi-batch reactor that is especially useful for measurement of gas/solid reaction rates at pressures in the range of 1 mTorr to 1500 Torr. The reactor operates by repeatedly imposing small AC modulations of reactant gas pressure on top of a much larger DC pressure background. Based on the rate of pressure relaxation following each AC pulse, the reaction rate is determined. Our design is characterized by modular construction from off-the-shelf, ultra-high-vacuum-compatible components, which facilitate easy retrofitting and adaptation to a range of experimental conditions. Automated experiment control and data acquisition is accomplished via a custom National Instruments © LabView virtual instrument. Data analysis is automated using a custom series of Mathworks © Matlab scripts. We demonstrate reactor performance through measurements of hydrogenation kinetics for a composite H 2 getter material consisting of 1,4-bis(phenylethynyl)benzene mixed with a palladium/carbon catalyst.

47 OTHER INSTRUMENTATION↗
DOE OSTI2019

Leveraging Connections 2018 Progress Report

This 2018 progress report contains a list of Los Alamos National Laboratory's top innovations of the time, including: Inventing - Los Alamos National Laboratory has been inventing for the past 75 years to accomplish the difficult, the unexpected and at times, what seems impossible. The Feynman Center is rethinking roadmaps for patent and software strategies to enable Los Alamos to achieve greater impact of its innovations; Innovating - The US technology advantage in national security is dependent on continual innovation. Los Alamos’ immense contributions in leading edge science and technology research and development are directed into solving complex and forthcoming national security challenges. Check out the Technology Snapshot platform at https://techsnapshot.lanl.gov to explore Los Alamos’s technologies ready to be developed into disruptive applications; Disrupting - Known well for developing the first nuclear weapon, big and small science at Los Alamos is behind many other disruptive innovations. This year we are highlighting Los Alamos’s leadership in small modular nuclear reactor design and development, work on international nuclear safeguards electronics instrumentation, and first demonstrated deployment of non-destructive testing to measure corrosion in nuclear material storage containers; Partnering - Los Alamos science has assisted partners big and small with their toughest challenges, this year Los Alamos scientists assisted two New Mexico farm owners—Tim Seaman with Manzanar Los Silvestres apple orchard in Abiquiu and Christopher Bassett of Freshies of New Mexico in El Guique — protect their crops from frost.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

ORNL Second Target Station Project: Biological & Environmental Science Workshop

Recent advances in neutron sources and instrumentation have opened up many new opportunities for the application of neutron scattering techniques in the biological and environmental sciences. Neutrons enable studies of the structure and dynamics of biological and environmental samples with a particular sensitivity to light elements, such as hydrogen, which is a key component of biological and environmental samples. Studies using neutrons are complementary to X-rays and have the unique advantage of being non-destructive and highly-penetrating. Oak Ridge National Laboratory’s upcoming Spallation Neutron Source (SNS) Second Target Station (STS) will provide high brightness cold neutron sources that significantly advance the scientific capabilities of neutron scattering instruments. The STS will advance our understanding of biological and environmental processes across spatial and temporal scales. The capabilities will enhance our ability to discover, design, and develop new materials essential for advanced sustainable technologies to address society’s most pressing needs. This report summarizes the discussions and recommendations from a joint workshop held by the STS Project and the Biological and Environmental Systems Science Directorate (BESSD) in June 2022. The purpose of the workshop was to explore science opportunities and capabilities related to biological and environmental systems that could be incorporated into both current and future STS instrument designs, as well as additional instruments at the SNS First Target Station (FTS) and High Flux Isotope Reactor (HFIR). With six breakout sessions, each with two invited plenary speakers from other institutions, the participants discussed a wide range of topics relevant to biological and environmental research. Based on the input from participants, a number of recommendations on instrumentation, sample environments, complementary multi-modal methods, data processing and analysis and sample deuteration are provided in the report. The participants also identified science opportunities that are emerging from the planned instrument capabilities at STS. Selected recommendations and science opportunities are listed in the Executive Summary.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2019

Thermal Hydraulic Experimental Test Article: Status Report for FY2019

The Thermal Hydraulic Experimental Test Article (THETA) is a facility that will be used to develop sodium components and instrumentation as well as acquire experimental data for validation of reactor thermal hydraulic and safety analysis codes. The facility will simulate nominal conditions as well as protected/unprotected loss of flow accidents in an SFR. High fidelity distributed temperature profiles of the developed flow field will be acquired with Rayleigh backscatter based optical fiber temperature sensors. The facility is being designed in partnership with systems code experts to tailor the experiment to ensure the most relevant and highest quality data for code validation. Figure 1 illustrates the main components of the primary system. THETA will be installed in the Mechanisms Engineering Test Loop (METL) with the primary system in the 28 inch Test Vessel #4, the secondary system in the 18-inch Test Vessel #6, and its sodium-to-air heat exchanger will occupy the location for 18-inch Test Vessel #5 (Test Vessel #5 will not be installed), Figure 2. The design work for the facility has been completed and is currently in the procurement, manufacturing and initial testing stage. The THETA primary system, including the redan, inlet and outlet plenum, 38 kW electric core simulant heater, submerged flowmeter, pump, and all instrumentation has either been procured or is currently being tested at Argonne. The secondary sodium system for cooling the tube-side sodium of the intermediate heat exchanger has all been analyzed according to ASME code and is in the final development stages before procurement.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

New In-pile Instrumentation to Support Fuel Cycle Research and Development

New and enhanced nuclear fuels are a key enabler for new and improved reactor technologies. For example, the goals of the next generation nuclear plant (NGNP) will not be met without irradiations successfully demonstrating the safety and reliability of new fuels. Likewise, fuel reliability has become paramount in ensuring the competitiveness of nuclear power plants. Recently, the Office of Nuclear Energy in the Department of Energy (DOE-NE) launched a new direction in fuel research and development that emphasizes an approach relying 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. A three-year strategic research program is proposed for developing the required test vehicles with sensors of unprecedented accuracy and resolution for obtaining the data needed to characterize three-dimensional changes in fuel microstructure during irradiation testing. When implemented, this strategy will yield test capsule designs that are instrumented with new sensor technologies for the Advanced Test Reactor (ATR) and other irradiation locations for the Fuel Cycle Research and Development (FC R&D) program. Prior laboratory testing, and as needed, irradiation testing, of these sensors will have been completed to give sufficient confidence that the irradiation tests will yield the required data. Obtaining these sensors must draw upon the expertise of a wide-range of organizations not currently supporting nuclear fuels research. This document defines this strategic program and provides the necessary background information related to fuel irradiation testing, desired parameters for detection, and an overview of currently available in-pile instrumentation. In addition, candidate sensor technologies are identified in this document, and a list of proposed criteria for ranking these technologies. A preliminary ranking of candidate technologies is performed to illustrate the path forward for developing real-time instrumentation that could provide the required data for the FC R&D program. This draft document is a starting point for discussion with instrumentation experts and organizations. It is anticipated that the document will be used to stimulate discussions on a wide-range of sensor technologies and to gain consensus with respect to the path forward for accomplishing the goals of this research program.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
NASA NTRS1983

Investigation of a para-ortho hydrogen reactor for application to spacecraft sensor cooling

The utilization of solid hydrogen in space for sensor and instrument cooling is a very efficient technique for long term cooling or for cooling at high heat rates. The solid hydrogen can provide temperatures as low as 7 to 8 K to instruments. Vapor cooling is utilized to reduce parasitic heat inputs to the 7 to 8 K stage and is effective in providing intermediate cooling for instrument components operating at higher temperatures. The use of solid hydrogen in place of helium may lead to weight reductions as large as a factor of ten and an attendent reduction in system volume. The results of an investigation of a catalytic reactor for use with a solid hydrogen cooling system is presented. Trade studies were performed on several configurations of reactor to meet the requirements of high reactor efficiency with low pressure drop. Results for the selected reactor design are presented for both liquid hydrogen systems operating at near atmospheric pressure and the solid hydrogen cooler operating as low as 1 torr.

Nast, T. C.↗