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At least 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2021

Tonopah Test Range Air Monitoring: CY2020 Meteorological, Radiological, and Wind Transported Particulate Observations

In 1963, the U.S. Department of Energy (DOE) (formerly the Atomic Energy Commission [AEC]), implemented Operation Roller Coaster on the Tonopah Test Range (TTR) and an adjacent area of the Nevada Test and Training Range (NTTR) (formerly the Nellis Air Force Range). This operation resulted in radionuclide-contaminated soils at the Double Tracks site and Clean Slate I, II, and III sites. This report documents observations made during ongoing monitoring of radiological, meteorological, and dust conditions at stations installed adjacent to the Clean Slate sites and at the TTR Sandia National Laboratories (SNL) Range Operations Center (ROC). The monitoring effort supports the DOE Office of Legacy Management in its stewardship of the Clean Slate sites as part of their mission to protect human health and the environment at sites of former nuclear device development and testing. Specific monitoring objectives include determining if wind blowing across the Clean Slate sites is transporting particles of radionuclide-contaminated soil beyond the physical and administrative boundaries of the sites, and providing information for planning long-term stewardship of the sites.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2021

Materials and Fuels Complex: Five-Year Investment Strategy (FY2021-2025)

The Department of Energy Office of Nuclear Energy (DOE-NE) vision is to “Advance nuclear energy science and technology to meet U.S. energy, environmental, and economic needs.” The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Thermocouple Testing in Support of the AGR-5/6/7 Experiment

This report documents thermocouple testing performed in IRC Lab C-15 over a period of seven years. This testing supported selection and characterization of the thermocouple set used in the AGR-5/6/7 experiment. The following summary was taken directly from the report. Temperature measurement is a challenging aspect of very high temperature irradiation experiments because commonly used high-temperature commercial thermocouples such as platinum-rhodium (Types S, R, and B) and tungsten-rhenium (Type C), suffer dramatic drift because of neutron-induced transmutation. As a result, these types of thermocouples, which are used routinely for industrial temperature measurements outside of reactors, are used only in very special circumstances for reactor experiments. Conversely, because of their low neutron cross-sections, Type N thermocouples are affected to only a limited extent by neutron irradiation. However, the use of these nickel-based thermocouples is limited when the temperature exceeds 1050°C due to drift arising from minor alloying elements migrating from the thermocouple's metal sheath to the thermoelements. This change in the composition of the thermo-elements results in significant decalibration of the signal. The issues described above were recognized during the early planning stages of the final AGR experiment (designated AGR-5/6/7), and a thermocouple furnace testing program was performed over a seven-year period (2014-2019, 2021) to first select and then characterize the best thermocouple set for the high temperature regions of the AGR-5/6/7 irradiation experiment. The calculated temperature range of the AGR-5/6/7 experiment was 600–1500°C. For temperatures below 1000°C standard Type N thermocouples were deemed adequate. The furnace testing campaign identified two thermocouple types suitable for measuring temperatures above 1000°C, a Mo/Nb thermocouple developed at INL called HTIR-TC, and a Type N thermocouple developed by Cambridge University (called herein Cambridge Type N), which featured a custom high nickel alloy sheath. One of the original goals of the furnace testing program was to identify a thermocouple capable of low drift operation near the peak temperature expected in AGR-5/6/7, i.e., about 1400°C. The HTIR-TC design appeared promising in this regard, however a manufacturing difficulty proved to be a barrier and instead the furnace testing focused on drift performance at 1250°C. The manufacturing difficulty was that the Nb sheaths of the HTIR-TCs experienced extreme embrittlement when heat treated at 1600°C or greater. Heat treatment is needed to stabilize the emf output of this TC type, and the higher the heat treatment temperature the higher the peak temperature of stable operation. Because of the sheath embrittlement the heat treatment temperature had to be lowered to 1450°C resulting in a stable operating temperature of about 1250°C. One of the successes of the furnace testing program was identification of a shortcoming in the heat treatment procedure that had been traditionally used in the production of HTIR-TCs. The shortcoming was that the entire heated length of the HTIR-TC sensor was not being heat treated, but rather only the part of the sensor expected to experience temperatures above 1000°C. The problem manifested itself when the thermocouples were removed from the heat treat furnace and placed in another furnace with a different geometry, their indicated temperatures would be widely scattered, but mostly in the negative direction. The solution was to heat treat the entire heated length of the sensor. Since the deepest immersion depth in the AGR-5/6/7 experiment was about 40 inches, a heat treatment length of 48 inches was used. After this change was implemented, thermocouples which were moved into a new environment with a different temperature profile (i.e., a different furnace), produced accurate temperature measurements. Although assembly of the AGR-5/6/7 experiment was completed in September of 2017 (and irradiation begun in 2018), furnace testing of thermocouples continued in 2018 and 2019. The main purpose of this testing was to establish very long-term drift characteristics of the HTIR and Cambridge Type N thermocouples installed in the experiment. Representative thermocouples from the same lots as those installed in the AGR-5/6/7 experiment were used. Additionally, thermocouples of different designs, (particularly variations on the HTIR-TC design) were "piggy-backed" on this testing program to provide insights for instrumenting future very high temperature irradiation experiments. This two-year testing program demonstrated that HTIR-TCs and Cambridge Type N TCs could operate at 1250°C for up to 10,000 hrs (and in some cases longer) while experiencing negative drifts on the order of 2-4°C/1000 hrs. This performance was considered acceptable given the extreme operating environment the sensors faced.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2023

Nuclear fuel irradiation testbed for nuclear security applications

The nuclear security community has long been interested in the identification and quantification of nuclear material signatures to understand a material’s provenance, use, and ultimate application. New forensics signatures and methods intended for non-traditional or advanced nuclear fuel applications require fuel irradiation experiments to demonstrate viability and validity. Integral fuel irradiations have historically required significant costs and long timelines to design, irradiate, and characterize. This paper describes how a recently developed nuclear fuel irradiation testbed can be used to provide a low cost, rapid turnaround, modular test environment for irradiation and evaluation of nuclear fuel specimens for nuclear security applications. The irradiation testbed houses six small ‘MiniFuel’ samples within hermetically sealed capsules inside targets that can be removed in between each ∼25-day operating cycle of the High Flux Isotope Reactor (HFIR). As many as nine targets can be irradiated using a single irradiation position (reflector region) in HFIR, allowing for varying irradiation temperatures and burnups. A suite of hot cell capabilities have been established to perform post-irradiation examination for measuring performance (e.g., fuel swelling, fission gas release) and facilitating experiment disassembly for subsequent property measurements, microstructural analysis, or chemical assay. This new testbed allows fuel irradiations to be conducted on an accelerated timeframe to enable rapid proof of concept testing and to provide reference material for nuclear fuel security applications. Recent applications using this testbed include the testing of isotopic taggants in UO 2 fuel (intentional forensics), testing of U-10Mo fuel for down-conversion of highly enriched uranium–fueled reactors, and the production of irradiated UO 2 fuel material for signature analysis of its isotopic composition (plutonium, fission gases, etc.).

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2023

Short Communication: Observation of Initial Burst Release of Fission Gas from High-Burnup UO 2 Nuclear Fuel During Thermal Transient

A system was developed and tested to provide a deeper understanding of the fission gas release kinetics from nuclear fuel during thermal transients. Pressure, temperature, spectral data, and optical images were simultaneously collected during resistive sample heating, and all released gases were collected in liquid nitrogen-cooled charcoal traps. Standup testing was performed with a high-burnup nuclear fuel segment from rod section with an average local burnup of 77 GWd/tU. During heating, the segment released approximately 6 ± 2% of generated fission gas inventory after ramp heating to 460°C. Here, heating ceased when the sample ejected from the holder, as observed by constant imaging.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

ENG 572 Interim Report

Sandia National Laboratories is a Federally Funded Research and Development Center (FFRDC) founded in 1949 with the mission of developing and testing the non-nuclear components of nuclear weapons. Sandia has since been involved with numerous projects to support the Department of Defense’s (DOE) National Nuclear Security Administration (NNSA). One such set of projects has been implementing over-the-road transportation security enhancements. Under this program, Sandia National Laboratories (SNL) has worked to develop interface compatibility modifications for existing shipping configurations. This summer I will be working with a line of trailers that have been used to transport high asset cargo. These vehicles have successfully traveled millions of miles without any accidents over the course of 15 years. The primary motivation for the work that I will complete this summer is to provide support for existing electronic communication technologies implemented at SNL. As part of an ongoing project to implement modifications to a trailer system, I will focus primarily on the characterization and testing of thermal electric coolers (TEC). Within the scope of the trailer project, these devices provide temperature control for lasers used on optical communication boards (OCB).

42 ENGINEERING↗
DOE OSTI2024

Tagged deep inelastic scattering measurement on deuterium with the LAD experiment

Here, the origin of the modification of the quark structure of nucleons in the nuclear medium can be tested with tagged recoil nucleon measurements from deep inelastic scattering off electrons on deuterium. The LAD experiment at the Thomas Jefferson National Laboratory (JLab) will measure the modification of the neutron structure function for high-momentum, highly-virtual neutrons by measuring the spectator recoil protons in coincidence with the scattered electron. An update on the experimental setup and projected results is presented. The experiment will collect data in Fall 2024.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2024

Activation of Insulation in Low Power Nuclear Experiments

Nuclear experiments aim to test new designs, procedures, and materials. Similar to conventional power reactors, potential sources of worker exposure to radiation is a primary concern. One such concern is the insulation materials that are used in or near the reactor, which can often create dust when being handled. If this material receives a substantial neutron dose, then it can become activated, which can pose a health risk if inhaled or ingested. This analysis looked at the activation of, and conservative dose estimates from, industrial grade insulations that may be used in a reactor experiment.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2023

Advanced Reactors Development in USA

Small and micro-reactor advanced reactor development in the USA. Advanced Reactor programs, nuclear renaissance, and INL as a test bed facility. These slides are for presentations in general on the past, INL and nuclear engineering worldwide, GEN IV reactor designs, SMRs and micro Rx designs, multiple commercial designs, and what INL as a test bed means and who uses this to their advantage.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Cumulative short-lived photofission product yields for nuclear forensic application

Experiments have long been carried out with linear accelerator (linac) produced bremsstrahlung X-rays to study high-energy photon-induced fission, known as “photofission”. One benefit of photofission, when compared with neutron-induced fission, is the ability to investigate fundamental physics of fission at excitation energies lower than what is possible with neutrons [1]. Absorption of an incident neutron results in a compound nucleus with an excitation energy at least equal to the neutron binding energy; thus, the lowest energy region of the fission barrier cannot be studied. Fission product yield distributions are dependent on the incident particle energy and the target nucleus mass. They are generally represented by a double-humped curve, with low and high mass “peaks” and an intermediate mass “valley”. As incident particle energy increases, the excitation energy of the nucleus increases, thereby increasing the overall yield of isotopes in the valley with respect to the two maxima or peaks of the fission product curve [3]. Photofission is most probable for incident photons around 14 MeV due to the giant dipole resonance. Induced photofission around this energy results in a fission product yield distribution that resembles 14 MeV DT neutron fission more than thermal or fast (500 keV) neutron fission. In post-detonation nuclear forensics, the peak-to-valley ratio of the fission product yield curve could be used for characterizing device type by providing an indication of the neutron energy spectrum that had induced fission (e.g., 500 keV vs. 14 MeV). The goal of this work is to improve nuclear data for isotopes of interest in nuclear forensics and special nuclear material (SNM) detection, specifically the fission product yield data via photofission experiments for the isotopes of 238U and 232Th. Photofission has been proposed as a less costly and more flexible production method for the valley isotopes that are representative of DT neutron fission product yield distributions, which would benefit nuclear forensics exercises in testing for analysis techniques. Natural uranium and thorium targets were irradiated with bremsstrahlung photons (nominal endpoint energies of 8, 14, and 20 MeV), generated via impingement of high-energy electrons on a tungsten electron-to-photon radiator. Gamma-ray spectroscopy was used to observe and identify the decay of the resulting fission products. The photofission yields of 102 fission products were measured for the set of the 238U and 232Th targets at the three bremsstrahlung X-ray endpoint energies 8, 14, and 20 MeV. These fragments included masses ranging from A=84 to A=144, with half-lives as short as 1.07 seconds.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2020

Materials and Fuels Complex Five-Year Investment Strategy (FY2020-FY2024)

The Department of Energy Office of Nuclear Energy (DOE-NE) mission is to provide the research, development, and demonstration (RD&D) foundation to extend the lives of the current operating reactor fleet, develop the next generation of nuclear reactors, and provide integrated nuclear fuel cycle solutions. The Materials and Fuels Complex (MFC) serves as the foundation of a nuclear RD&D enabling test bed at Idaho National Laboratory (INL) and is an integral part of a National Reactor Innovation Center (NRIC) strategy. MFC facilities focus on developing and maintaining RD&D capabilities that can increase research throughput, reduce barriers to deployment, and facilitate commercialization of new ideas and technologies for clean and secure sources of energy. This document is a complimentary document to the Materials and Fuels Complex Five-Year Mission Strategy FY-20 – FY-24 (INL/EXT-20-57224) that defines specific implementation strategies for increasing research capability and throughput through targeted investments in research facilities, research instruments, and research staff. Refer to INL/EXT-20-57224 for details about MFC, its capabilities, and the overall mission strategy for the Materials and Fuels Complex.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

Validation and Recalibration of the Solubility Models in Support of the Heater Test in Salt Formations

The US Department of Energy Office of Nuclear Energy is conducting a brine availability heater test to characterize the thermal, mechanical, hydrological and chemical response of salt at elevated temperatures. In the heater test, brines will be collected and analyzed for chemical compositions. In order to support the geochemical modeling of chemical evolutions of the brines during the heater test, we are recalibrating and validating the solubility models for the mineral constituents in salt formations up to 100°C, based on the solubility data in multiple component systems as well as simple systems from literature. In this work, we systematically compare the model-predicted values based on the various solubility models related to the constituents of salt formations, with the experimental data. As halite is the dominant constituent in salt formations, we first test the halite solubility model in the Na-Mg-Cl dominated brines. We find the existing halite solubility model systematically over-predict the solubility of halite. We recalibrate the halite model, which can reproduce halite solubilities in Na-Mg-Cl dominated brines well. As gypsum/anhydrite in salt formations controls the sulfate concentrations in associated brines, we test the gypsum solubility model in NaCl solutions up to 5.87 mol•kg –1 from 25 ° C to 50 ° C. The testing shows that the current gypsum solubility model reproduces the experimental data well when NaCl concentrations are less than 1 mol•kg –1 . However, at NaCl concentrations higher than 1, the model systematically overpredicts the solubility of gypsum. In the Na + —Cl – —SO 4 2– —CO 3 2– system, the validation tests up to 100 ° C demonstrate that the model excellently reproduces the experimental data for the solution compositions equilibrated with one single phase such as halite (NaCl) or thenardite (Na 2 SO 4 ), with deviations equal to, or less than, 1.5 %. The model is much less ideal in reproducing the compositions in equilibrium with the assemblages of halite + thenardite, and of halite + thermonatrite (Na 2 CO 3 •H 2 O), with deviations up to 31 %. The high deviations from the experimental data for the multiple assemblages in this system at elevated temperatures may be attributed to the facts that the database has the Pitzer interaction parameters for Cl – —CO 3 2– and SO 4 2– —CO 3 2– only at 25 ° C. Finally, in the Na + —Ca 2+ —SO 4 2– —HCO 3 – system, the validation tests also demonstrate that the model reproduces the equilibrium compositions for one single phase such as gypsum better than the assemblages of more than one phase.

36 MATERIALS SCIENCE↗
DOE OSTI2025

Evaluation of an Accident Tolerant Fuel Leak in the Advanced Test Reactor

Accident Tolerant Fuels (ATF), which are nuclear fuel sources designed to withstand operational irregularities and incidents, have been a topic of interest in the nuclear industry for several decades. Interest in ATF technology surged following the 2011 accident at Fukushima Daiichi in Japan. At the Advanced Test Reactor (ATR), one of Idaho National Laboratory’s (INL) four operating nuclear reactors, the ATF program is a collaborative effort between the national laboratory and various stakeholders within the nuclear industry. This program focuses on the research and development of novel fuel compositions, cladding, and component materials with enhanced accident-resistant properties. During one of ATR’s 60-day operating cycles in 2024, the reactor experienced five unplanned shutdowns. Following the fifth shutdown, radiation monitors detected an increase in radiation levels coming from the loop piping. Subsequent water samples confirmed the cause was a leak of fission products from the ATF experiment, designated as ATF-2C. The source of the leak was identified as the instrumented section of the test train. The primary discussions in this presentation are 1) the design of the ATF test train, 2) the operating parameters leading up to and following the detection of the leak, and 3) the quantification and characterization of the released fission products.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

SCALE Code System

The SCALE Code System is a widely used modeling and simulation suite for nuclear safety analysis and design that is developed, maintained, tested, and managed by the Reactor and Nuclear Systems Division (RNSD) of Oak Ridge National Laboratory (ORNL). SCALE provides a comprehensive, verified and validated, user-friendly tool set for criticality safety, reactor and lattice physics, radiation shielding, spent fuel and radioactive source term characterization, and sensitivity and uncertainty analysis. Since 1980, regulators, licensees, and research institutions around the world have used SCALE for safety analysis and design. SCALE provides an integrated framework with dozens of computational modules, including three deterministic and three Monte Carlo radiation transport solvers that are selected based on the desired solution strategy. SCALE includes current nuclear data libraries and problem-dependent processing tools for continuous-energy (CE) and multigroup (MG) neutronics and coupled neutron-gamma calculations, as well as activation, depletion, and decay calculations. SCALE includes unique capabilities for automated variance reduction for shielding calculations, as well as sensitivity and uncertainty analysis. SCALE’s graphical user interfaces assist with accurate system modeling, visualization of nuclear data, and convenient access to desired results.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRSDate not supplied

Effects of Varying the Heat Transfer Coefficient on Engine Performance

A physics-based Nuclear Thermal Propulsion (NTP) Testing Reference Design (TRD) power balance model was coded in Simulink to investigate engine performance for various design and parameter modifications. Since the primary mode of heat transfer in NTP engines is convective, the convective heat transfer coefficient (HTC) is a key parameter that requires accurate representation. The industry standard Westinghouse correlation has an uncertainty of ±20% which was investigated in this study. The results showed that a 20% decrease in the HTC led to a 4.14% increase in maximum fuel temperature while a 20% decrease in the HTC led to a 1.81% decrease in maximum fuel temperature suggesting that narrowing the uncertainty of this correlation through experimental work would be a critical step in the development of NTP engines. Furthermore, a maximum fuel temperature relationship with specific impulse was developed for the TRD engine which showed potential engine operation between specific impulse values of 715 and 900 seconds with minimal changes to the engine design. This graph could be useful for high level vehicle performance estimations for fuel types with different maximum operating temperatures.

Heat Transfer Coefficient↗
NASA NTRSDate not supplied

Effects of Heat Transfer Coefficient Variation on Nuclear Thermal Propulsion Engine Performance

A physics-based Nuclear Thermal Propulsion (NTP) Testing Reference Design (TRD) power balance model was coded in Simulink to investigate engine performance for various design and parameter modifications. Since the primary mode of heat transfer in NTP engines is convective, the convective heat transfer coefficient (HTC) is a key parameter that requires accurate representation. The industry standard Westinghouse correlation has an uncertainty of ±20% which was investigated in this study. The results showed that a 20% decrease in the HTC led to a 4.14% increase in maximum fuel temperature while a 20% decrease in the HTC led to a 1.81% decrease in maximum fuel temperature suggesting that narrowing the uncertainty of this correlation through experimental work would be a critical step in the development of NTP engines. Furthermore, a maximum fuel temperature relationship with specific impulse was developed for the TRD engine which showed potential engine operation between specific impulse values of 715 and 900 seconds with minimal changes to the engine design. This graph could be useful for high level vehicle performance estimations for fuel types with different maximum operating temperatures.

Heat Transfer Coefficient↗
NASA NTRS2017

Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one to be used at the Device Assembly Facility (DAF) at the Nevada National Security Site. Fabrication, assembly, and testing of the depleted uranium core has allowed for higher fidelity system level testing at GRC, and has validated the fabrication methods to be used on the highly enriched uranium core that will supply heat for the DAF KRUSTY demonstration.

Briggs, Maxwell H.↗