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At least 955 records · Page 53
All sourcesNASA NTRSDOE OSTI
DOE OSTI2023

PARTITIONING AND TRANSMUTATION OF USED NUCLEAR FUEL IN SUPPORT OF GEOLOGICAL WASTE DISPOSAL

Nuclear energy is a low-carbon technology that generates bulk baseload electricity and supports long-lasting eco-friendly sustainability goals. However, it still faces challenges, such as the securing and managing of long-lived nuclear actinides and fission products—most notably the used fuel, which is considered high-level waste (HLW). Direct geological waste disposal would be the most economical option for HLW but requires highly qualified sites whose geophysical behavior will remain stable for 10,000 years. Partitioning and transmutation (P&T) of used fuel reduces radioactivity and decay heat generation, enabling efficient geological waste disposal. P&T research and development, along with the prospects of P&T techniques such as wet/dry separation and accelerator-driven systems, were reviewed to identify critical needs that must be met to foster their successful implementation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2023

TRANSPORTATION CONSIDERATIONS FOR USED FUEL AND ADVANCED NUCLEAR ENERGY TECHNOLOGIES

This study focuses on requirements and challenges pertaining to advanced transportable nuclear energy systems, which are currently in development and feature varying degrees of mobility. These reactors mostly use high-assay low-enriched uranium (HALEU) fuel, and only limited refueling is required during their lifecycles. Due to the smaller source terms, higher burnup fuel, and potentially higher thermal efficiency involved, the waste inventory will be less than that generated by conventional light-water reactors. However, the relevant regulatory requirements must be met when transporting HALEU fuel (whether fresh or used) inside the reactor vessel—mainly the fuel material transportation, siting, and physical protection requirements outlined in US Nuclear Regulatory Commission (NRC) regulations 10 CFR 71 and 73. Spent nuclear fuel (SNF), as well as irradiated fuel (used during any interval of reactor operation), contains highly radioactive materials that generate heat and necessitate special precautions during transportation. This study addresses the challenges that this incurs, and outlines potential solutions.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2012

Instrument development continues in Oak Ridge

Peer review panels composed of 80 external scientists recently visited Oak Ridge National Laboratory (ORNL) to review almost 700 proposals for experiments on 23 instruments at the Spallation Neutron Source (SNS) and High Flux Isotope Reactor (HFIR). These were proposed for the time period from January-June 2012. About 40% of the proposals were approved for beam time and 20% were placed on an alternate list if time becomes available. The Hybrid Spectrometer HYSPEC at SNS began its commissioning in September 2011. HYSPEC is otpimized for studying low energy dynamics in single-crystal samples using a broad variety of sample environments, and is equipped with a polarization analysis capability. It is expected to be available for users on a limited basis in the second half of 2012. The detector tank of CORELLI has been installed on beamline 9 at SNS. Now that the tank is in place, banks of neutron detectors and boron carbide shielding will be installed around the interior. CORELLI is optimized to probe complex disorder in crystalline materials through diffuse scattering from single-crystal samples. It will begin commissioning in 2014. CORELLI is one of four instruments being developed under the SING II (SNS Instruments Next Generation II) project. The others are the Macromolecular Neutron Diffractometer (MANDI), the Vibrational Spectrometer (VISION, scheduled to begin commissioning in 2012), and the Time of Flight Ultra Small Angle Neutron Scattering Instrument (TOF-USANS). The single crystal neutron diffractometer IMAGINE, was deliverd to HFIR in October 2011. Preliminary testing has been carried out. IMAGINE will provide atomic resolution information on chemical, organic, metallo-organic and protein single crystals that will enable their chemical, physical and biological structure and function to be understood. This instrument will benefit scientists with interests in pharmaceuticals, minerals and other inorganic crystals, small molecules, molecular organo-metallic crystals and metal-organic frameworks (MOFs) molecular crystal structures. The quasi-Laue geometry, combined with a large solid angle detector, will enable rapid data collection from crystals with volume < 1mm{sup 1} and unit cell < 100 {angstrom}. Construction and installation of the optical system is in progress. Commissioning is expected to start in April 2012.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor power operational fluctuations; - Study of gamma radiation effects only. The fission neutron spectrum can be limitative for some of the tests, as most neutrons are in the 1-2 MeV energy range. Significant progress has been made lately in compact neutron generators using D-D and D-T fusion reactions, achieving higher neutron fluxes and longer lifetime than previously available. The advantages of using compact neutron generators for testing of electronic components and circuits will be also discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2021

Development of Bismuth and Platinum Bi-Metallic Nanoparticles to Enhance Melt Wire Temperature Resolution

Advanced manufacturing (AM) based on direct-write (DW) technologies has emerged as the predominant enabler for the fabrication of active and passive sensors to be deployed in harsh operating environments seen in a nuclear reactor. Recently, Idaho National Laboratory and Boise State University have recently established capabilities to incorporate advanced manufacturing (AM) methods to accelerate, modernize and enhance functionality of nuclear sensors and instrumentation to achieve the goal of enhancing the safety and efficiency of nuclear reactors. A significant thrust of this work includes the development of nuclear relevant feedstock materials compatible with a variety of direct-write processes for the development, fabrication and testing of AM sensors for peak temperature detection and neutron flux monitoring. For this report, bismuth and bismuth/platinum (BiPt) bi-metallic nanoparticles were synthesized using wet chemical approaches to develop bi-metallic feedstock materials to enhance the sensitivity of melt wires for peak temperature detection. Nanoparticle characterization was performed with Transmission Electron Microscopy (TEM) for nanoparticle composition and size, Transmission X-Ray Diffraction Fluorescence (TXRF) for nanoparticle composition, and Differential Scanning Calorimetry to elucidate the melting point of BiPt bi-metallic nanoparticles. Preliminary results indicate bi-metallic BiPt nanoparticles as a viable pathway for fabricating high resolution AM melt wires.

36 MATERIALS SCIENCE↗
DOE OSTI2018

Plant Modernization Pathway: FY 2019 External Review Summary Report

This report describes an external review conducted by the Light Water Reactor Sustainability Program Plant Modernization Pathway (referred to as Pathway) to solicit feedback on the topics and results of the ongoing Instrumentation and Control (I&C) research program. This review was held in conjunction with the Nuclear Energy Institute (NEI) Digital I&C Working Group meeting that was held at Idaho National Laboratory on October 16–17, 2018. Members of the Nuclear Regulatory Commission (NRC) staff working on digital I&C issues through their Integrated Action Plan also attended. The NEI and NRC agreed that the meeting participants would serve as an external review panel for the purpose of obtaining expert input on the value and timing of the research projects. Five major research areas were presented, along with demonstrations of new technologies in the Human Systems Simulation Laboratory and the Computer-Assisted Virtual Environment. Following the presentations, the external review was conducted by referring to a set of summary slides of the five research areas, and allowing each participant to enter ratings and comments on a survey form as each area was addressed. Fourteen forms were returned. These feedback forms were analyzed by the Pathway who compiled and responded to the data and comments, which are documented in this report. The participants also had opportunity to provide comments on the LWRS Program as a whole and to suggest additional research topics of value. The feedback provided by the external review participants is taken to be a strong endorsement of the types of projects being conducted by the Pathway, the value they hold for the operating nuclear plants, and the general timing of need. The feedback aligns well with the priorities, levels of efforts allocated for the research projects, and project schedules of the current Pathway research program. The feedback also provides realistic observations on the practicality of some aspects of implementing these technologies. The Pathway will consider all of the feedback and address the recommended research topics that have been identified by the external review participants. Pathway research plans will be adjusted as warranted.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Automated Fissile Mass Estimation Python Script for the Advanced Experimental Fuel Counter (AEFC)

The Advanced Experimental Fuel Counter (AEFC) is a non-destructive assay (NDA) instrument developed to quantify the residual fissile mass in research reactor spent fuel assemblies. The system is installed in a spent fuel pool and measurements are acquired underwater. A calibration curve relates the totals and coincidence counting signal from active interrogation of the spent fuel assemblies to the fissile mass remaining in the fuel.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Measurements of SNAC2 area dosimeters placed in different configurations around the PROSPERO reactor and comparison with TRIPOLI-4 calculations

In the context of criticality accident alarm system tests, several experiments were carried out in 2013 on the PROSPERO reactor to study the response to neutron and gamma of different devices and dosimeters, particularly on the SNAC2 dosimeter. This article presents the results of this criticality dosimeter in different configurations, and compares the experimental measurements with the results of calculation performed with the TRIPOLI-4 Monte-Carlo Neutral Particles transport code. PROSPERO is a metallic critical assembly managed by the Criticality, Neutron Science and Measurement Department located at the French CEA Research Center of Valduc. The core, surrounded by a reflector of depleted uranium, is composed of 2 horizontal cylindrical blocks made of a highly enriched uranium alloy which can be placed in contact, and of 4 depleted uranium control rods which allow the reactor to be driven. This reactor, placed in a cell 10 m x 8 m x 6 m high, with 1.4-meter-thick concrete walls, is used as a fast neutron spectrum source and is operated at stable power level in delayed critical state, which can vary from 3 mW to 3 kW. PROSPERO is extensively used for electronic hardening or to study the effect of the neutrons on various materials. The SNAC2 criticality dosimeter is a zone dosimeter allowing the off line measurement of criticality accident neutron doses. This dosimeter consists of the pile up of seven activation foils embedded into a 23 mm diameter x 21 mm height cadmium container. The activation measurement of each foil, using a gamma spectroscopy technique, gives information about the neutron reaction rates. The SNAC2 software allows the spectrum unfolding from these values, taking into account the hypothesis of a particular spectrum shape, in three components: a Maxwell spectrum component for the thermal range, a 1/E component for the epithermal range, and a Watt spectrum component for the high energy range. Moreover, from the neutron spectrum, the SNAC software can calculate the neutron fluence integrated by the dosimeter and the neutron dose. During the 3 weeks measurement campaign many radioprotection devices were used. To modify the spectrum seen by these devices, several shields of various thicknesses made of concrete or polyethylene, with or without cadmium covers, were placed in the PROSPERO cell. These devices allow the study of criticality accident spectra in several environments: from metal to pseudo liquid. The fluxes measured by the SNAC2 devices were compared with TRIPOLI-4 calculations. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Simulation of differential die-away instrument’s response to asymmetrically burned spent nuclear fuel

Here, previous simulation studies of Differential Die–Away (DDA) instrument’s response to active interrogation of spent nuclear fuel from a pressurized water reactor (PWR) yielded promising results in terms of its capability to accurately measure or estimate basic spent fuel assembly (SFA) characteristics, such as multiplication, initial enrichment (IE) and burn-up (BU) as well as the total plutonium content. These studies were however performed only for a subset of idealized SFAs with a symmetric BU with respect to its longitudinal axis. Therefore, to complement the previous results, additional simulations have been performed of the DDA instrument’s response to interrogation of asymmetrically burned spent nuclear fuel in order to determine whether detailed assay of SFAs from all 4 sides will be necessary in real life applications or whether a cost and time saving single sided assay could be used to achieve results of similar quality as previously reported in case of symmetrically burned SFAs.

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2019

Thermal Hydraulic Experimental Test Article - Status Report for FY2019 (Rev. 1)

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. 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.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

We report the Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance-covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U 238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance-covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Practical applications of model checking in the finnish nuclear industry - 281

Model checking is a powerful, formal, computer-assisted verification method that can be used to prove that a model of a (hardware or software) system fulfills stated properties. When used right, model checking can prove the correctness of instrumentation and control (I and C) system application logic, be it a software or a field-programmable gate array (FPGA) based design. The verified properties can also address unwanted functionality, making model checking a very effective method for analyzing spurious actuation. Despite the benefits, model checking is still not widely adopted in the verification of I and C systems, one exception being the Finnish nuclear industry. Since 2008, VTT has applied model checking in practical customer work related to the Olkiluoto 3 EPR, Loviisa 1 and 2 VVER-440, and Hanhikivi 1 AES-2006 nuclear power plants, on commission from either the regulator or the utilities. In this paper, we look at how the method has been used in each of these three projects. We also introduce a user-friendly, graphical modelling tool called MODCHK. We then present an overall view of the design issues detected during nine years of customer work, and one practical example. The 100% coverage means that the design errors that are found often relate to scenarios that are otherwise hard to account for (e.g., exact timing of events, or improper operator actions). (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

A multi-detector, digitizer based neutron depth profiling device for characterizing thin film materials

Neutron depth profiling (NDP) is a mature, nondestructive technique used to characterize the concentration of certain light isotopes in a material as a function of depth by measuring the residual energy of charged particles in neutron induced reactions. Historically, NDP has been performed using a single detector, resulting in low intrinsic detection efficiency, and limiting the technique largely to high flux research reactors. In this work, we describe a new NDP instrument design with higher detection efficiency by way of spectrum summing across multiple detectors. Such a design is capable of acquiring a statistically significant charged particle spectrum at facilities limited in neutron flux and operation time.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

A reactor for high-throughput high-pressure nuclear magnetic resonance spectroscopy

The design of a reactor for operando nuclear magnetic resonance (NMR) monitoring of high-pressure gas-liquid reactions is described. The Wisconsin High Pressure NMR Reactor (WiHP-NMRR) design comprises four modules: a sapphire NMR tube with titanium tube holder rated for pressures as high as 1000 psig (68 atm) and temperatures ranging from −90 to 90 °C, a gas circulation system that maintains equilibrium concentrations of dissolved gases during gas-consuming or gas-releasing reactions, a liquid injection apparatus that is capable of adding measured amounts of solutions to the reactor under high pressure conditions, and a rapid wash system that enables the reactor to be cleaned without removal from the NMR instrument. The WiHP-NMRR is compatible with commercial 10 mm NMR probes. Reactions performed in the WiHP-NMRR yield high quality, information-rich, and multinuclear NMR data over the entire reaction time course with rapid experimental turnaround.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Improving Predictions Under Uncertainty of Material Plasma Device Operations

Understanding the properties of materials when exposed to various plasma temperatures and fluxes is essential to the building and operating of fusion reactors. The Material Plasma Exposure eXperiment (MPEX) is an instrument currently being developed by the Department of Energy (DOE) for this purpose. MPEX is expected to come online in stages over the next five years. Proto-MPEX, the predecessor to MPEX, operated from 2014 to 2021, and was designed to understand the generation of plasma temperatures and fluxes at orders of magnitude below what will be obtained by MPEX. This work uses the recently developed stochastic neural network (SNN), a machine learning technique capable of operating under uncertainty to provide a surrogate model for the Proto-MPEX device. We demonstrate that SNN outperforms Bayesian neural network (BNN), a standard in the field of machine learning with uncertainty. The development of a robust surrogate of the Proto-MPEX will aid in the commissioning and operation of the MPEX device.

Archibald, Rick↗
DOE OSTI2015

Fluorophore-based sensor for oxygen radicals in processing plasmas

A high concentration of radicals is present in many processing plasmas, which affects the processing conditions and the properties of materials exposed to the plasma. Determining the types and concentrations of free radicals present in the plasma is critical in order to determine their effects on the materials being processed. Current methods for detecting free radicals in a plasma require multiple expensive and bulky instruments, complex setups, and often, modifications to the plasma reactor. This work presents a simple technique that detects reactive-oxygen radicals incident on a surface from a plasma. The measurements are made using a fluorophore dye that is commonly used in biological and cellular systems for assay labeling in liquids. Using fluorometric analysis, it was found that the fluorophore reacts with oxygen radicals incident from the plasma, which is indicated by degradation of its fluorescence. As plasma power was increased, the quenching of the fluorescence significantly increased. Both immobilized and nonimmobilized fluorophore dyes were used and the results indicate that both states function effectively under vacuum conditions. The reaction mechanism is very similar to that of the liquid dye.

42 ENGINEERING↗
NASA NTRS1981

Proceedings of the Conference on High-temperature Electronics

The development of electronic devices for use in high temperature environments is addressed. The instrumentational needs of planetary exploration, fossil and nuclear power reactors, turbine engine monitoring, and well logging are defined. Emphasis is place on the fabrication and performance of materials and semiconductor devices, circuits and systems and packaging.

Source record↗
DOE OSTI2022

Snowmass 2021 Scintillating Bubble Chambers: Liquid-noble Bubble Chambers for Dark Matter and CE$\nu$NS Detection

The Scintillating Bubble Chamber (SBC) Collaboration is developing liquid-noble bubble chambers for the quasi-background-free detection of low-mass (GeV-scale) dark matter and coherent scattering (CE$\nu$NS) of low-energy (MeV-scale) neutrinos. The first physics-scale demonstrator of this technique, a 10-kg liquid argon bubble chamber dubbed SBC-LAr10, is now being commissioned at Fermilab. This device will calibrate the background discrimination power and sensitivity of superheated argon to nuclear recoils at energies down to 100 eV. A second functionally-identical detector with a focus on radiopure construction is being built for SBC's first dark matter search at SNOLAB. The projected spin-independent sensitivity of this search is approximately $10^{-43}$ cm$^2$ at 1 GeV$/c^2$ dark matter particle mass. The scalability and background discrimination power of the liquid-noble bubble chamber make this technique a compelling candidate for future dark matter searches to the solar neutrino fog at 1 GeV$/c^2$ particle mass (requiring a $\sim$ton-year exposure with non-neutrino backgrounds sub-dominant to the solar CE$\nu$NS signal) and for high-statistics CE$\nu$NS studies at nuclear reactors.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗