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At least 199 records · Page 11
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

A Portable Miniature Cryogenic Environment for In Situ Neutron Diffraction

Neutron diffraction instruments offer a platform for materials science and engineering studies at extended temperature ranges far from ambient. As one of the widely used neutron sample environment types, cryogenic furnaces are usually bulky and complex, and they may need hours of beamtime overhead for installation, configuration, cooling, and sample change, etc. To reduce the overhead time and expedite experiments at the state-of-the-art high-flux neutron source, we developed a low-cost, miniature, and easy-to-use cryogenic environment (77–473 K) for in situ neutron diffraction. A travel-size mug serves for the environment where the samples sit inside. Immediate cooling and an isothermal dwell at 77 K are realized on the sample by direct contact with liquid N 2 in the mug. The designed Al inserts serve as the holder of samples and heating elements, alleviate the thermal gradient, and clear neutron pathways. Both a single-sample continuous measurement and multi-sample high-throughput measurements are demonstrated in this environment. High-quality and refinable in situ neutron diffraction patterns are acquired on model materials. The results quantify the orthorhombic-to-cubic phase transformation process in LiMn 2 O 4 and differentiate the anisotropic lattice thermal expansions and bond length evolutions between rhombohedral perovskite oxides with composition variation.

47 OTHER INSTRUMENTATION↗
DOE OSTI2024

Godiva IV Simulated Radiation Field Characterization and Variance Reduction

Godiva IV is a system comprised of highly enriched uranium alloyed with molybdenum in the form of fuel plate rings. The reactor, along with its predecessors, was designed with the unique ability to satisfy interests in the super-prompt-critical reactor operation space. Originally, the reactor was part of the Los Alamos Critical Experiments Facility (LACEF) at Technical Area-18 (TA-18). The radiation field around Godiva at this facility was well characterized and understood. As a fast neutron system, the neutron spectrum in and around Godiva was close to a Watt Fission spectrum. The Kiva where Godiva IV was located at LACEF was made of thin, sheet metal walls which did not contribute significantly to the neutron spectrum. Following the transition of LACEF to the National Critical Experiments and Research Center (NCERC) in Nevada, Godiva-IV was moved from TA-18 to the Device Assembly Facility (DAF) at the Nevada National Security Site (NNSS). Part of this move brought renewed interest in radiation field characterization. The new facility introduced significant changes to the environment surrounding Godiva, and preliminary foil irradiation results suggested that the room contribution to the neutron spectrum was significant. Unlike at TA-18, a large thermal neutron signature was added to the fast spectrum from Godiva due to significant room return. A primary goal due to the additional complexity that the room return adds to the Godiva IV radiation emission spectrum was the development of an efficient Monte Carlo N-Particle (MCNP) calculation capable of characterizing the neutron spectrum anywhere in the room around Godiva. A campaign of activation foil irradiations and analysis were completed to support the validation of the MCNP model. The modeling of these foils in MCNP can be easily done with a standard volumetric neutron flux tally. However, given the multitude of locations and reaction rates to be modeled, further steps must be taken to increase the efficiency of these calculations in MCNP. During this study, a benchmark model currently under development for Godiva IV was used. A qualitative assessment of the thermal neutron contributors was performed using spatial neutron distribution plots. Additional detail was added to the model based on the qualitative results showing the thermal spectrum’s large sensitivity to hydrogenous material. Neutron energy spectra was evaluated at discrete locations in the room around Godiva to quantify the relative contribution of various components. It was discovered that the concrete walls are the largest contributor to the thermal signature, with minor contributions from plastic components surrounding Godiva. Following these results, two different variance reduction techniques were implemented to improve the problem efficiency in these calculations. In the first approach, an F5 point detector tally was implemented in the standard Godiva IV criticality problem. The second approach involved a weight-window generator implementation with an F5 point detector tally in a fixed source problem. The weight window implementation reduced the runtime from 42739.55 minutes to 1803.34 minutes (computer time), compared to the F5 KCODE implementation.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2023

β- and γ-Counting for Pre-detonation Nuclear Forensics on Eu-155

Post-detonation nuclear forensics was performed at Los Alamos National Laboratory (LANL) on 155Eu, a fission product on the wing of the fission product production curve whose yield is sensitive to fission fuel and neutron energy. With a half-life of 4.753 years, 155Eu provides a longer-lived option for these measurements than other fission products with similar mass numbers. The Chemistry Division Group - Nuclear and Radiochemistry, at LANL routinely measures a suite of fission products from 235U fissions in thermal neutron flux experiments known as thermal calibration exercises, using a mixture of gas proportional β-decay counting and γ-spectrometry on HPGe detectors. The fission products of interest are reported relative to a high-yield reference fission product from the same sample to create a running average ratio specific to neutron energy and fuel type; Equation 1 below shows the ratio-of-ratios R-value measured fission products are reported in: R i x = [A( i X)/A( 99 Mo)] unknown [A( i X)/A( 99 Mo)] 235 U n,th where A denotes activity, i X is the nuclide of interest, and 235 U n,th denotes irradiations of 235 U with thermal neutrons. Measurements of unknown fission spectra are ratioed to a running-average of thermal calibration results, and the resulting R-value can be referred to as against a library of irradiation conditions.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

Characterizing secondary neutrons at BLIP for isotope production applications

Fast secondary neutrons created at the Brookhaven Linac Isotope Producer (BLIP) facility following proton irradiation were characterized by the foil activation technique and compared with FLUKA Monte Carlo simulations. The FLUKA-simulated neutron flux was spectrally adjusted following the maximum entropy formalism using the International Reactor Dosimetry and Fusion File (IRDFF-II), with predictions agreeing with experimental measurements to within 9 % following the adjustment procedure. Here, a multitude of degrader configurations were simulated to assess the feasibility of improving the fast (E n > 20 MeV) secondary neutron yield at the proposed neutron target position (“N-slot”). A configuration where the N-slot is closest to the proton degrader produced the highest fast neutron yield, with tungsten degraders achieving the best performance. Assuming the optimized target-degrader configuration proposed in this work, we discuss potential isotope production opportunities with secondary neutrons. In most cases the yields are in the order of several mCi.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Validation of Thermal Scattering Laws for Light Water at Elevated Temperatures with Diffusion Experiments

Light water is the most important neutron moderator in many reactor applications. Thermal neutron scattering kernels for H bound in H 2 O are represented by ENDF-format thermal scattering laws (TSLs) evaluated at discrete temperatures T . Nuclear data evaluations are conventionally validated utilizing a combination of critical benchmarks and experimental cross section data. Existing public critical benchmarks may be inadequate for testing water TSLs over the full range of T of interest in reactor applications, and experimental thermal scattering cross section data for water is sparse at elevated T . In this work, MC21 is used to simulate the decay of the equilibrium thermal neutron flux in light-water spheres of several radii. The fundamental-mode time eigenvalue $\mathcal{a}$ is calculated at 22 °C and 227 °C (at saturation pressure) for each sphere using three different H-H 2 O TSLs. Fitting $\mathcal{a}$ to a polynomial function of geometric buckling allows calculation of the thermal neutron diffusion length L . Extrapolation lengths are treated as a function of the transport mean free path and geometry. Experimental L data from 25 publications at 49 temperatures (from 10 °C to 295 °C), computed by several different time-dependent and space-dependent decay methods, is used to develop an empirical fit of L vs. T . The MC21-calculated L for the TSLs tested are within ±1% of the predicted values at 22 °C and 227 °C. This validation approach, which may be repeated at arbitrary T , constitutes an integral benchmark specific to and characterizing the detailed physics of the thermal scattering kernel applied.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2020

NDX: Neutron Dose Rate Meters with Extended Capabilities

Particle accelerators, irradiation facilities, and nuclear power plants would benefit from using neutron dose rate meters featuring improved energy response covering full range of expected neutron energies, improved dynamic range of measured dose rates while providing the real-time data, and improved ability to operate in harsh radiation environments including heavy photon backgrounds and pulsed beams. The novel design of the neutron moderator for the NDX dose rate meter implements a layer of a Beryllium-loaded medium Z material which, similar to the earlier known WENDI detectors, allows the device to detect higher energy neutron flux and convert it to dose. Using Beryllium helps to fill the known dip in the response functions around 5-50 MeV. Optimized design parameter set, obtained using FLUKA Monte Carlo simulations, features the NDX moderator made of polyethylene embedding a layer of BeCu alloy inside, with the dose rate response constant within ±10% in the neutron energy range from about 100 keV to 500 MeV, and ±25% essentially in the full energy range above thermal. A new sensor technique implemented in the NDX is based on using the Ionization Chamber (IC) filled with 3He gas to capture the moderated neutron fluence, and includes a symmetric IC filled with regular 4He to evaluate the contributions due to the non-neutron ionizing radiation. Advantages of the method include large accessible dynamic range, relative insensitivity of the IC probes to photon radiation fields, and the ability to subtract their contribution. Two prototype NDX detectors were built, calibrated, and operated in various conditions at the CEBAF 12 GeV electron accelerator at JLab. The first results showed reliable neutron dose rate measurements in the presence of overwhelming photon radiation fields, and at practical high dose rate levels of up to approximately 10 Sv/h. Detector design, operation, first results, and further development plans are presented.

Degtiarenko, Pavel V.↗
DOE OSTI2023

Accident Tolerant Fuel Test (ATF-2c) Irradiation Test Report

A new fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2C has completed its first cycle of prototypic steady state irradiation. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 24 test pins has been calculated.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

Survey of U.S. Research Reactor Auxiliary Facilities Used for Material Testing and Basic Neutron Science

A catalogue of and attributes of the Materials Irradiation and Testing facilities (MIF) and the Basic Neutron Science facilities (BSF) of auxiliary (AUX) facilities of research reactors are compiled. The survey of these facilities is drawn from the set of U.S. university and DOE research reactor facilities and should be considered as a reference point when comparing commonalities in international AUX facilities. The size and shielding capabilities of the facilities, the typical characterization equipment used, and the neutron flux and irradiation capabilities of the facilities are listed. Short descriptions of experimental activities and current practices are detailed for MIFs and BSFs. The attributes are important in consideration of the reconfiguration of the facilities for purposes other than stated mission, i.e., for proliferation of weapons-usable nuclear material (WUNM). An evaluation of the capacity for production rate of WUNM from neutron beams-on-targets or a sample that has been placed in an irradiation position for a period is provided. Further evaluation of reconfiguration of these MIF and BSF are recommended to refine the proliferation risks. Specifically, MIF hot cells and potential configurations, additional modeling of facility throughput and development of methods for determining levels of concern for MIF and BSF have all been identified as steps for refinement of determination of the risks associated with these facilities. Future work will focus on: 1. A model of projected throughput for different MIF and BSF configurations that will be used to provide a window of potential operational misuse for a facility and better understand the capability of production and the rate of processing for WUNM in these facilities. 2. A misuse study for various configurations of MIF hot cell layouts detailing capacity, specifications, shielding constraints for MIF operations, identification of a maximum shielding thickness, and reasonable need associated with each activity performed in a MIF. 3. Development of a rule of thumb rating system utilizing data gathered from previous reports to show levels of concern at a given power level, flux, and experiment set.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

The importance of delayed neutron precursors in gamma dose calculations for activated primary heat exchanger components in the Molten Salt Breeder Reactor

The Molten Salt Reactor (MSR) Multiphysics Applications technical area in the Nuclear Energy Advanced Modeling and Simulation program has supported the development of 3D Monte Carlo models of the Molten Salt Breeder Reactor (MSBR) over the last couple of years. This MSBR model was previously run with the Shift Monte Carlo code to perform radiation shielding calculations in the reactor cell area. The MSBR is a 2250 MWth (1000 MWe) liquid-fueled molten salt reactor design developed at Oak Ridge National Laboratory in the 1970s. Determining the source terms from activated primary heat exchanger (HX) components is important because delayed and prompt neutron fluxes incident on these components affect the dose rate in the primary HX maintenance areas. This information can be used in the development of remote handling procedures required during shutdown for maintenance. A methodology has been developed and is proposed in this paper to quantify the activated source term from the primary HX components as a result of the movement of the delayed neutron precursors in flowing primary fuel salt through the primary HXs in the MSBR. The goal of this research is to evaluate the gamma dose rates in the maintenance hatches above the primary HX using the activated HX source terms. The study showed that the gamma dose rates are approximately two orders of magnitudes higher when accounting for the neutron activation from the movement of delayed neutron precursors through the HXs than when flowing fuel is not considered. Thus, the movement of delayed neutron precursors must be taken into account for accurately predicting the neutron activation of primary loop components.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Overview of fast particle experiments in the first MAST Upgrade experimental campaigns

Abstract MAST-U is equipped with on-axis and off-axis neutral beam injectors (NBI), and these external sources of super-Alfvénic deuterium fast-ions provide opportunities for studying a wide range of phenomena relevant to the physics of alpha-particles in burning plasmas. The MeV range D-D fusion product ions are also produced but are not confined. Simulations with the ASCOT code show that up to 20% of fast ions produced by NBI can be lost due to charge exchange (CX) with edge neutrals. Dedicated experiments employing low field side (LFS) gas fuelling show a significant drop in the measured neutron fluxes resulting from beam-plasma reactions, providing additional evidence of CX-induced fast-ion losses, similar to the ASCOT findings. Clear evidence of fast-ion redistribution and loss due to sawteeth (ST), fishbones (FB), long-lived modes (LLM), Toroidal Alfvén Eigenmodes (TAE), Edge Localised Modes (ELM) and neoclassical tearing modes (NTM) has been found in measurements with a Neutron Camera (NCU), a scintillator-based Fast-Ion Loss Detector (FILD), a Solid-State Neutral Particle Analyser (SSNPA) and a Fast-Ion Deuterium- α (FIDA) spectrometer. Unprecedented FILD measurements in the range of 1–2 MHz indicate that fast-ion losses can be also induced by the beam ion cyclotron resonance interaction with compressional or global Alfvén eigenmodes (CAEs or GAEs). These results show the wide variety of scenarios and the unique conditions in which fast ions can be studied in MAST-U, under conditions that are relevant for future devices like STEP or ITER.

Physics↗
DOE OSTI2021

ATHENA: A unique radiation environment platform at the National Ignition Facility

This paper describes the ATHENA platform, an energy tuning assembly, which was developed to spectrally shape the National Ignition Facility (NIF) deuterium–tritium fusion neutron source to a thermonuclear (fusion) plus prompt fission neutron spectrum with a capability to act as a short-pulse neutron source. This unique, otherwise inaccessible radiation environment complements existing experimental facilities and capabilities. Here, the flexible ATHENA irradiation positions were modeled using an ensemble of Monte Carlo simulations with stochastic sampling of the nuclear cross-sections to characterize the radiation environments and uncertainty for the platform. Validation of the internal neutron spectrum produced from fielding ATHENA at NIF occurred through neutron flux unfolding with 20 measured activation products.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2021

Accident Tolerant Fuel Test 2B (ATF-2B) Irradiation Test Report

A fully prototypic testing platform for testing new accident tolerant fuel (ATF) designs for light water reactors has been established in the center flux trap of Idaho National Laboratory’s advanced test reactor (ATR). The irradiation experiment named ATF-2 has completed 7 cycles of prototypic steady state irradiation irradiating 41 test pins to burnups as high as 30 MWd/kgU. Irradiation conditions have been maintained via the loop 2a pressurized water coolant loop and have been monitored through in-situ instrumentation during each cycle. Power and fast neutron flux histories of each of the 41 test pins has been calculated using a combined MCNP, and ORIGIN methodology to provide accurate and reliable power histories on a per pin basis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

A revisitation of the ENEA TAPIRO Fast Source Reactor neutronic capabilities

This work is devoted to the analysis of the neutronic characterization of the ENEA-Casaccia center RSV TAPIRO fast neutrons source reactor performed during the 80's in the frame of a joint ENEA-SCK-CEN Mol collaboration. The Belgian scientist Albert Fabry, who was available to investigate a fast research reactor to perform a full characterization of the system neutron flux, led the TAPIRO experimental campaign. In particular, the paper is focused on the comparison between some 'recommended' Fabry experimental reaction rates traverses and nowadays MCNP Monte Carlo and ERANOS deterministic calculation results. It appears that the quality of the results depends on the penetration depth into the copper reflector. The influence of copper nuclear data are more evident in the ERANOS results. Massimo Salvatores has provided a great contribution to the interpretation of the TAPIRO neutronics: this humble work is dedicated to him.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Retractable Sensors for In-Core Use in Material Test Reactors - conf paper

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 - 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on friction drive wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 - 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology's MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 - 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 - 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology's MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Retractable Sensors for In-Core Service in Material Test Reactors

Material Test Reactors (MTRs) such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL) are used to irradiate nuclear fuels and materials to evaluate their performance after high levels of exposure to a reactor in-core environment. The most critical tests are equipped with instrumentation leads, which allow real-time data collection. However, because of the very harsh environment inside high-power MTR experiments, there are very few sensors that can survive and maintain their calibrated readings for the time periods required to obtain the high neutron doses needed for new fuels and materials qualification. As a result, sometimes sponsoring programs are forced to accept low reliability of sensors, collecting useful data for only part of the experiment duration. The work described herein is based on the observation that MTRs normally run at constant power and the corresponding conditions within reactor experiments typically evolve relatively slowly. Therefore, even one or two measurements per day would provide a complete and representative data set. With this in mind, INL has embarked on a program to develop a mechanism capable of pushing a very small-diameter sensor (typically a thermocouple or optical fiber) into the location to be measured, leave the sensor for roughly 60 seconds to allow it to reach equilibrium and transmit the signal, then pull it up and away from the high neutron flux and high-temperature region. Small-diameter capillary tubes, up to 8 m long, are used to guide the sensors to the appropriate locations. These capillary tubes serve as essentially very deep, thin-walled thermowells. The distance a thermocouple or optical fiber would need to traverse is on the order of 40 – 80 cm. By adopting this infrequent cycling strategy, the thermocouple or optical fiber would spend only a few hours in the high-neutron flux/high-temperature environment over the duration of even the longest irradiation experiment. To date, INL has developed two styles of drive mechanisms. The first is based on counter-rotating wheels which drive the sensors in a manner similar to a small MIG welder. This has the advantage of being able to accommodate a very long insertion length. The second is based on a ball screw drive and has the advantages of positive attachment and being able to move more than one sensor at a time. Both drive mechanisms have been fabricated and tested in a laboratory setting. Both systems can handle hard mineral insulated cable (such as thermocouples) or optical fibers encased in small diameter tube. The sizes tested to date are 1 – 1.6 mm diameter. Work in this area is ongoing with an eye toward demonstration in the Massachusetts Institute of Technology’s MITR reactor, followed by deployment in an ATR irradiation experiment.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Assembly and Functional Test of NRAD Heated Instrumentation Rig

This is a technical report (TEV-4296) which documents the configuration and materials of construction for a heated sensor test conducted in the NRAD reactor. It is being reviewed for external release so that operators of university reactors can evaluate whether this level of documentation will be acceptable for similar tests conducted in their reactors. The Objective section of this TEV is reproduced below. The purpose of this TEV is to document assembly and functional testing of a device which is designed to allow advanced sensors to be temporarily placed at the periphery of the NRAD reactor. While in this location, the operational characteristics of the sensors can be evaluated while they are exposed to a neutron flux. This device (or “rig”) consists of a cartridge heater surrounded by six guide tubes, which are in turn surrounded by insulation and an aluminum dry tube. The heater allows for the sensors to be tested while at elevated temperature.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Post-Irradiation Examination to Quantify Irradiation-Induced Bowing of SiGA® Silicon Carbide Composite Structures (Final CRADA Report – NFE-23-09937)

As part of the SiC-based material development at General Atomics Electromagnetic Systems (GA-EMS), this project involved the first-of-a-kind experimental post-irradiation examination of the irradiation-induced bowing response in SiC composite structures under a neutron flux gradient. The SiC composite miniature channel specimen was provided by GA-EMS for the irradiation experiment. To quantify the irradiation-induced bowing of the channel specimen, a series of visual and dimensional inspections were conducted using the unique capabilities at Oak Ridge National Laboratory (ORNL), including a custom profilometry rig. The post-irradiation examination results of this project will assist GA-EMS in validating their fuel performance model for SiC-based core structures in neutron irradiation environments.

36 MATERIALS SCIENCE↗