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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 1,603 records · Page 89
All sourcesNASA NTRSDOE OSTI
DOE OSTI2022

Status of the data acquisition, trigger, and slow control systems of the Mu2e experiment at Fermilab

The Mu2e experiment at the Fermilab will search for a coherent neutrinoless conversion of a muon into an electron in the field of an aluminum nucleus with a sensitivity improvement by a factor of 10,000 over existing limits. In this work, the Mu2e Trigger and Data Acquisition System (TDAQ) uses otsdaq framework as the online Data Acquisition System (DAQ) solution. Developed at Fermilab, otsdaq integrates several framework components — an artdaq-based DAQ, an art-based event processing, and an EPICS-based detector control system (DCS), and provides a uniform multi-user interface to its components through a web browser. Data streams from the Mu2e tracker and calorimeter are handled by the artdaq-based DAQ and processed by a one-level software trigger implemented within the art framework. Events accepted by the trigger have their data combined, post-trigger, with the separately read out data from the Mu2e Cosmic Ray Veto system. Foundation of the Mu2e DCS, EPICS – an Experimental Physics and Industrial Control System – is an open-source platform for monitoring, controlling, alarming, and archiving. A prototype of the TDAQ and the DCS systems has been built and tested over the last three years at Fermilab’s Feynman Computing Center, and now the production system installation is underway. This work presents their status and focus on the installation plans and procedures for racks, workstations, network switches, gateway computers, DAQ hardware, slow controls implementation, and testing.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2024

Passive Temperature Sensors for Nuclear Applications

Thermocouples are generally used to provide real-time temperature indications in instrumented tests performed at material and test reactors. Passive temperature monitors, such as Silicon Carbide (SiC) and melt wires, may be included in such tests as an independent technique of detecting peak temperatures experienced during irradiation. In less expensive static (drop-in) capsule tests, which have no leads attached for real-time data transmission, melt wires, and SiC temperature monitors (TMs) are essentially the only possibility for peak temperature indication. A melt wire involves placing materials (wires) of a known composition and melting temperature in a test. An inventory is maintained at Material Science Laboratory (MSL) for melt wires ranging in temperatures from 30°C to 1500°C. Unfortunately, melt wires are limited in that it can only detect whether a single temperature is or is not exceeded (melt wire melted or not). SiC TMs, which can also be used to detect peak irradiation temperatures, are advantageous because a single monitor can allow to determine the peak temperature reached within a relatively broad range (100 – 1200°C) resulting in accuracies within ±20°C. Irradiation temperature is determined by measuring a property change after isochronal annealing or during a continuously monitored annealing process using specialized equipment at MSL. Recent research has produced a passive monitor known as sublime temperature monitor. This passive sensor has the capability of recording temperature gradients and pinpointing exactly where a temperature is located along that gradient. Long measurement lengths are achieved with very high accuracy in the location of desired temperature measurements (±2 mm over a 1 m span); however, this sensor has not been deployed in a nuclear reactor. This article will focus only on passive temperature sensors currently being researched and implemented under the Advanced Sensors and Instrumentation (ASI) program at Idaho National Laboratory (INL).

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Natural Gas Drilling near the Underground Nuclear Test Site of Rulison, Colorado - 19571

In 1969, the Project Rulison test was conducted in western Colorado using a nuclear device to stimulate natural gas production by fracturing low-permeability, gas-bearing rock. The presence of radionuclides (specifically tritium) in the produced gas persisted above acceptable levels, and the test was abandoned. Advances in hydrofracturing technology over the past 10-15 years have revolutionized the extraction of hydrocarbons from low-permeability formations, leading to a significant increase of drilling in the Rulison area. Drilling activity nearing the Rulison site has raised concerns that remnant radioactivity might be encountered. The site, now under the care of the US Department of Energy (DOE), is protected by a 16-hectare (40 acre) institutional control boundary (Lot 11) that prohibits drilling or extraction of materials below 1830 meters (6000 feet) within the lot boundary. Drilling within 0.8 kilometer (0.5 mile) of the site requires a hearing before the Colorado Oil and Gas Conservation Commission. DOE recommends that gas developers adopt a conservative, staged drilling approach, allowing gas reserves near the Rulison site to be recovered in a manner that minimizes the likelihood of encountering contamination. The staged approach calls for collecting data from wells away from the site before drilling closer. For instance, the wells 1.2 kilometer (0.75 mile) west of the site did not encounter any unexpected geologic conditions and now have 8 years of production without encountering any detonation-related radionuclides. The next set of wells would be 0.8 kilometer (0.5 mile) west of the site. This sequential method drills first at locations with a low probability of encountering contamination and then moves closer. This process protects public safety while allowing the collection of additional data to confirm that contamination has not migrated beyond the 16-hectare (40-acre) institutional control boundary. The staged drilling approach was simulated with a modeling effort that estimates the current extent of contamination and the potential for production-induced contaminant migration to nearby wells. The model incorporates geologic data from nearby wells and was calibrated to production data from both the reentry well production testing (1970-1971) and the wells 1.2 kilometer (0.75 mile) from the site that began producing in 2010. The geologic data include both sand/shale distribution statistics as well as the tendency of hydrofractures to preferentially propagate east and west due to the regional stress field. The model results indicate that it is unlikely that wells near the site would ever encounter site-related radionuclides. (authors)

03 NATURAL GAS↗
DOE OSTI2013

Progress Letter Report on Bending Fatigue Test System Development for Spent Nuclear Fuel Vibration Integrity Study (Out-of-cell fatigue testing development - Task 2.4)

Vibration integrity of high burn-up spent nuclear fuel in transportation remains to be a critical component of US nuclear waste management system. The structural evaluation of package for spent fuel transportation eventually will need to see if the content or spent fuel is in a subcritical condition. However, a system for testing and characterizing such spent fuel is still lacking mainly due to the complication involved with dealing radioactive specimens in a hot cell environment. Apparently, the current state-of-the-art in spent fuel research and development is quite far away from the delivery of reliable mechanical property data for the assessment of spent fuels in the transport package evaluation.

Spent Nuclear Fuel Vibration Integrity↗
DOE OSTI2017

Real-Time In-Pile Acoustic Measurement Infrastructure at the Advanced Test Reactor

The core of any nuclear reactor presents a particularly harsh environment for sensors and instrumentation. The reactor core also imposes challenging constraints on signal transmission from inside the reactor core to outside of the reactor vessel. In this paper, an acoustic measurement infrastructure installed at the Advanced Test Reactor (ATR), located at Idaho National Laboratory, is presented. The measurement infrastructure consists of ATR in-pile structural components, coolant, acoustic receivers, primary coolant pumps (PCPs), a data acquisition system, and signal-processing algorithms. Intrinsic and cyclic acoustic signals generated by the operation of the PCPs are collected and processed. Finally, the characteristics of the intrinsic signal can indicate the process state of the ATR (such as reactor startup, reactor criticality, reactor attaining maximum power, and reactor shutdown) during operation (i.e., real-time measurement).

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

High-temperature oxidation and quenching of chromium-coated zirconium alloy ATF cladding tubes with and w/o pre-damage

Chromium-coated zirconium alloys are one of the promising candidates for accident-tolerant fuel cladding (ATF) tubes for light water reactors (LWRs). In this study, the high temperature oxidation and degradation of two types of Cr coatings (cold spray and physical vapor deposition) with and without pre-damage by scratches were investigated on prototype rod segment samples filled with ZrO 2 pellets and tightly sealed with welded end caps. Isothermal tests at 1100 and 1200 °C were terminated by quenching with water; transient tests were performed up to 1500–1600 °C until complete coating failure. The positive effect of both types of Cr coatings was observed in all tests. Pre-damaged specimens showed only locally increased oxidation of the scratched zone, but no negative effects on the adjacent Cr coating. Finally, the behavior of the two types of coatings is compared and the degradation mechanisms are discussed based on hydrogen release data, extensive metallographic post-test examinations, and the current state of global research.

36 MATERIALS SCIENCE↗
DOE OSTI2023

Spin Alignment of Vector Mesons in Heavy-Ion Collisions

Polarized quarks and antiquarks in high-energy heavy-ion collisions can lead to the spin alignment of vector mesons formed by quark coalescence. Here, using the relativistic spin Boltzmann equation for vector mesons derived from Kadanoff-Baym equations with an effective quark-meson model for strong interaction and quark coalescence model for hadronizaton, we calculate the spin density matrix element $ρ_{00}$ for Φ mesons and show that anisotropies of local field correlations with respect to the spin quantization direction lead to Φ meson's spin alignment. We propose that the local correlation or fluctuation of Φ fields is the dominant mechanism for the observed Φ meson's spin alignment and its strength can be extracted from experimental data as functions of collision energies. The calculated transverse momentum dependence of $ρ_{00}$ agrees with STAR's data. We further predict the azimuthal angle dependence of $ρ_{00}$ which can be tested in future experiments.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2013

Study of the Quasi-Elastic Scattering in the NOvA Detector Prototype

NOvA is a 810 km long base-line neutrino oscillation experiment with two detectors (far 14 KTon and near detector 300 Ton) currently being installed in the NUMI o -axis neutrino beam produced at Fermilab. A 222 Ton prototype NOvA detector (NDOS) was built and operated in the neutrino beam for over a year to understand the response of the detector and its construction. The goal of this thesis is to study the muon neutrino interaction data collected in this test, specifically the identification of quasi-elastic charged-current interactions and measure the behavior of the quasi-elastic muon neutrino cross section.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Sixth Status Report: Testing of Aged Softwood Fiberboard Material for the 9975 Shipping Package

Samples have been prepared from several 9975 lower fiberboard subassemblies fabricated from softwood fiberboard. Physical, mechanical and thermal properties have been measured following varying periods of conditioning in each of several environments. These tests have been conducted in the same manner as previous testing on cane fiberboard samples. Overall, similar aging trends are observed for softwood and cane fiberboard samples, with a few differences. Some softwood fiberboard properties tend to degrade faster in some environments, while some cane fiberboard properties degrade faster in the two most aggressive environments. As a result, it is premature to assume both materials will age at the same rates, and the preliminary aging models developed for cane fiberboard might not apply to softwood fiberboard. However, it is expected that both cane and softwood fiberboard assemblies will perform satisfactorily in conforming packages stored in a typical KAC storage environment for up to 15 years. Samples from an additional 3 softwood fiberboard assemblies have begun aging during the past year to provide information on the variability of softwood fiberboard behavior. Aging and testing of softwood fiberboard will continue and additional data will be collected to support development of an aging model specific to softwood fiberboard.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2018

Post-Irradiation Examination of Low Neutron Dose Metal Fuel Samples

The FCT irradiation series at the High Flux Isotope Reactor (HFIR) using the Hydraulic Tube Facility (HT) was initiated in FY-2012 with the goal of performing irradiation testing on advanced fuels and materials. Data from the FCT irradiation series was to be used to further the fundamental scientific understanding of phenomena involved in the evolution of metal fuels microstructure under irradiation based on separate effects testing of factors including elemental composition, temperature, and neutron fluence. This data would then be used to inform theoretical models for in-service performance of metallic fuels.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2022

Verification and Validation Tests of the Gamma Library of the ARC Software Package (Rev.1)

As part of the verification and validation (V&V) efforts of the Argonne Reactor Code (ARC) software system to support the Versatile Test Reactor (VTR) project of the U.S. Department of Energy (DOE), work has been performed to verify and validate the gamma (or photon) library of the multigroup cross section generation code MC 2 -3 of the ARC system, which is the essential part of the coupled neutron and gamma heating calculations to determine the heat generation rate or power distribution in the core. This V&V study was carried out in three steps: 1) review of the procedures and utility programs for generating the MC 2 -3 gamma library, 2) verification tests of cross sections and KERMA factors by comparing the total heat generation rate of coupled neutron and gamma calculation with that obtained with reaction Q values under the assumption of local gamma energy deposition and by comparing the prompt heat generation rate with that calculated with the MCNP6.2 code, and 3) validation tests of cross sections and KERMA factors by analyzing the ZPPR-15D gamma dose data measured with thermoluminescent dosimeters (TLD).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

The Deep Underground Neutrino Experiment (DUNE) Program

The Deep Underground Neutrino Experiment (DUNE) is a next-generation long-baseline neutrino oscillation experiment aimed at determining the neutrino mass hierarchy and the CP-violating phase. The DUNE physics program also includes the detection of astrophysical neutrinos and the search for signatures beyond the Standard Model, such as nucleon decays. DUNE consists of a near detector complex located at Fermilab and four 17 kton Liquid Argon Time Projection Chamber (LArTPC) far detector modules to be built 1.5 km underground at SURF, approximately 1300 km away. The detectors are exposed to a wideband neutrino beam generated by a 1.2 MW proton beam with a planned upgrade to $>$ 2 MW. Two 770 ton LArTPCs (ProtoDUNEs) have been operated at CERN for over 2 years as a testbed for DUNE far detectors and have been optimized to take new cosmic and test-beam data in 2024-2025. The DUNE and ProtoDUNE experiments and physics goals, as well as recent progress and results, are presented.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2016

DETAILED ANALYSIS OF NEAR-IR WATER (H{sub 2}O) EMISSION IN COMET C/2014 Q2 (LOVEJOY) WITH THE GIANO/TNG SPECTROGRAPH

We observed the Oort cloud comet C/2014 Q2 (Lovejoy) on 2015 January 31 and February 1 and 2 at a heliocentric distance of 1.3 au and geocentric distance of 0.8 au during its approach to the Sun. Comet Lovejoy was observed with GIANO, the near-infrared high-resolution spectrograph mounted at the Nasmyth-A focus of the TNG (Telescopio Nazionale Galileo) telescope in La Palma, Canary Islands, Spain. We detected strong emissions of radical CN and water, along with many emission features of unidentified origin, across the 1–2.5 μ m region. Spectral lines from eight ro-vibrational bands of H{sub 2}O were detected, six of them for the first time. We quantified the water production rate [ Q (H{sub 2}O), (3.11 ± 0.14) × 10{sup 29} s{sup −1}] by comparing the calibrated line fluxes with the Goddard full non-resonance cascade fluorescence model for H{sub 2}O. The production rates of ortho-water [ Q (H{sub 2}O){sup ORTHO}, (2.33 ± 0.11) × 10{sup 29} s{sup −1}] and para-water [ Q (H{sub 2}O){sup PARA}, (0.87 ± 0.21) × 1029 s{sup −1}] provide a measure of the ortho-to-para ratio (2.70 ± 0.76)). The confidence limits are not small enough to provide a critical test of the nuclear spin temperature.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2023

Machining of Alloy 709 Creep-fatigue Specimens from G. O. Carlson Heat

Alloy 709 has been selected as the next candidate material for Section III, Division 5 qualification in the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC) for elevated-temperature nuclear construction. The qualification data package requires an assortment of information and material data including tensile, creep, and creep-fatigue performance at elevated temperatures from different material heats. The goal of the project is to generate a dataset to support qualification of Alloy 709 material for ASME BPVC design code. Working towards the project goal, Idaho National Laboratory needs to conduct a series of tests on three different heats to support the A709 code case development. At present, there is a gap in the data package for one of the heats: Heat number 58776 manufactured by G.O. Carlson. To address this data gap, Argonne National Laboratory transmitted five plates of A709 heat 58776 manufactured by G.O. Carlson to Idaho National Laboratory. These plates were solution annealed at 1150°C and heat treated at 775°C for 10 hours. The objective of this specification is to procure a series of creep-fatigue specimens to support the qualification data package. The creep-fatigue specimen design captures the cyclic material performance at elevated temperature. The material performance data generated from specimens machined herein will support the ASME BPVC code case development and establish design limits, and design life curves for Alloy 709 material.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Validating automated resonance evaluation with synthetic data

The integrity and precision of nuclear data are crucial for a broad spectrum of applications, from national security and nuclear reactor design to medical diagnostics, where the associated uncertainties can significantly impact outcomes. A substantial portion of uncertainty in nuclear data originates from the subjective biases in the evaluation process, a crucial phase in the nuclear data production pipeline. Recent advancements indicate that automation of certain routines can mitigate these biases, thereby standardizing the evaluation process and enhancing reproducibility. This research aims to provide a methodology, framework, and metrics for the validation of automated nuclear data evaluation software leveraging high-quality synthetic data that closely mimic real experimental observables. An introduced error metric provides a scale and intuitive measure of the evaluation quality by quantifying the estimate’s accuracy and performance across the specified energy range. Synthetic data provides access to experimental observables and underlying resonance parameters, enabling comparison of different evaluations. The methodology is demonstrated using Ta-181 isotope data in the resolved resonance region. The Automated Resonance Identification Subroutine (ARIS), which operates without prior resonance information, was used to test and showcase the framework’s capabilities utilizing the proposed error metrics. The results demonstrate the effectiveness of the proposed approach and framework for optimizing software parameters and testing hypotheses through “what-if” controlled experiments, such as modifying assumptions about experimental conditions or average resonance parameters.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

Out-of-Pile Testing and Instrumentation Transient Water Irradiation System

Current initiatives to increase the burnup of conventional nuclear fuels past the approximate 62 GWd/t limit have been spurred on by direct savings to refueling and waste storage. The technical justification for a new license limit requires extensive qualification through experimental testing. Unlike beginning-of-life fuels, high-burnup fuels are more susceptible to fuel fragmentation, relocation, and dispersal (FFRD), therefore more data is needed to characterize fuels under key accident scenarios. The Transient Reactor Test Facility (TREAT) located at the Idaho National Laboratory has developed a testing apparatus architecture to test fuels and claddings at prototypic conditions. The Transient Water Irradiation System (TWIST) is the latest iteration of a testing device capable of conducting loss of coolant accidents (LOCAs) in TREAT. The Out-of-Pile Testing and Instrumentation TWIST (OPTI-TWIST) is an electrically heated device that is analogous to TWIST. OPTI-TWIST allows for detailed instrumentation and thermal-hydraulic characterization. TWIST ultimately aims to conduct the most advanced in-situ diagnostics to evaluate FFRD in a prototypic LOCA. Moreover, it will explore the phenomenological bifurcation of a decay-energy heat up driven LOCA and a stored-energy heat up driven LOCA. The instrumentation suite includes conventional thermocouples and pressure transducers in addition to an electro impedance sensor, an acoustic emission sensor, an optical pressure sensor, and an optical pyrometer. Characterizing these instruments in OPTI-TWIST eliminates complications of irradiation effects while preserving extreme thermal-hydraulic conditions. Finally, benchmarking both devices to a thermal-hydraulic code like the Reactor Excursion and Leak Analysis Program (RELAP)5-3D provides a unique opportunity for iteration. Pre-test predictions and post-test interpretations inform the physical designs, operational procedures, test conditions, and instrumentation types and positions.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗