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At least 991 records · Page 55
All sourcesNASA NTRSDOE OSTI
DOE OSTI2017

A new probe concept for internal pipework inspection - 241

The interior visual inspection of nuclear pipework is a critical inspection activity required to ensure the continued safe, reliable operation of plant and thus avoid costly outages. Typically, the video output from a manually deployed probe is viewed by an operator online with the task of identifying and estimating the location of surface defects such as cracks, corrosion and pitting. However, it is very difficult to estimate the nature and spatial extent of defects from the often disorientating small field of view video of a relatively large structure. This work describes a new visual inspection system incorporating a wide field of view machine vision camera and additional sensors designed for inspecting 3 - 6 inch diameter pipes. The output of the system is a photorealistic model of the internal surface of the pipework. The generation of this model relies upon a core component of the system in the form of image feature extraction which estimates camera location. This paper considers the accuracy of this estimation as a function of the number and configuration of the extracted image features. (authors)

42 ENGINEERING↗
DOE OSTI2016

Accelerated Corrosion Tests to Evaluate Long-Term Performance of BORAL{sup R} in Spent Fuel Pools

Neutron absorber materials are used to increase storage capacity in spent fuel pools (SFPs) while maintaining criticality safety margins. BORAL{sup R} - a neutron absorber material commonly used in SFPs - is a metal matrix composite containing boron carbide (B{sub 4}C) and Al-1100 mixture. The aluminum (Al) cladding, which is on both sides of the core, serves as a protective barrier. Al cladding is susceptible to corrosion, which can lead to blisters between the core material and the cladding over time. Based on operational experience to date, pits also have been observed via visual examination and microscopy. EPRI has maintained a database containing coupon data spanning more than 25 years, and to date no loss of BORAL{sup R} efficacy has been observed. In the United States, about 50% of the plants use BORAL{sup R} as neutron absorber material in SFPs. BORAL{sup R} is also used in nuclear plants in Mexico, Korea, Taiwan, and other countries. Given that the plant lifetime is increasing-and, hence, the SFP lifetime-EPRI initiated a five-year accelerated corrosion test in 2013 to evaluate and determine the long-term performance of BORAL{sup R} in SFPs. This paper presents an overview of the accelerated corrosion test project. Results from the Year 1 and Year 2 test results are also presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Neutron-Gamma flux and Dose calculations in a Pressurized Water Reactor (PWR)

The Institute of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to the French nuclear safety authorities. In this context, one of its missions is to improve nuclear safety through the enhancement of knowledge regarding operating nuclear power reactors. The issues concerning the aging and the characterisation of the reactor vessel require knowledge of the neutron and gamma radiation environment on the vessel and beyond. Since few years, ENEA (Energia Nucleare e Energie Alternative, i.e. Nuclear Energy and Alternative Energy agency) and IRSN collaborate on Monte-Carlo simulations and Variance Reduction Techniques (VRT), which are used to determine neutron and gamma spectra in and out of the core. Additionally, IRSN participates in the DISCOMS project, which stands for 'Distributed Sensing for Corium Monitoring and Safety', aiming at examining the status of the third barrier of confinement, in order to define possible mitigation strategies in case of a severe accident, namely: i) reactor pressure vessel breakthrough and corium relocation outside the vessel, ii) reactor pit concrete erosion and iii) corium cooling. One of the IRSN tasks in the frame of the DISCOMS project is to provide the radiation level in terms of neutron and gamma doses and flux energy spectra at various locations. This paper summarizes the study performed on the radiation transport in the Tihange-I Belgian nuclear reactor and contributes to several issues requiring the knowledge of the radiation environment in and outside the reactor vessel. This paper gives an overview of the work presented at the ICRS-13 and RPSD-2016 conference (for more details see the proceedings paper)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

Criticality safety analysis of spent fuel storage or handling disposals using burnup credit - loading curve calculation - 14445

MILLENNIUM and REEL SAS Companies have shared their own experience to develop a neutron and criticality calculations chain in order to perform isotopic composition predictions of spent fuel after irradiation in a reactor and criticality analysis of fuel storages taking into account burnup credit. The whole calculation sequence is divided into three steps, all calling SCALE 6.1.2 modules. The first two steps are dedicated to the evaluation of isotopic compositions of burnt fuel. As transport and depletion calculations in three dimensions appears to be way too long for the whole core (about several months), dealing with each fuel assembly one by one is the alternative solution MILLENNIUM and REEL SAS commonly chose to study. Therefore, the first step aims at determining the axial burnup profile of a given fuel assembly using a 3-D model with infinite reflections on lateral axes. The T6-depl sequence of TRITON module coupling KENO-VI and ORIGEN-S was used in this step. Sensitivity studies confirmed that addition of the 8 closest neighbors of the studied fuel assembly in the model has an insignificant impact on the axial burnup profile. However, fuel assemblies in the peripheral zone of the core must be treated separately with a model including the nearest fuel assemblies and the effect of peripheral water. Then, the second phase evaluates the isotopic compositions of each fuel assemblies at the center of the active height with a 2-D model under conservative core conditions. The T-depl sequence of the TRITON module coupling NEWT to ORIGEN-S is called in this step. Each fuel assembly must be studied separately to consider loading pattern changes occurring at each cycle. An infinite model with reflections on lateral axes is used and the closest neighbors of the fuel assembly are limited to 2 pins - which is around the size of the neutrons mean free path in the core. This is an innovative physical trick aiming at compensating for a limited computing power. Finally, isotopic compositions calculated are combined with the axial burnup profile of the first step to extrapolate the concentrations along the whole active length. Data can then be added in any criticality model, such as a spent fuel pit or any disposal handling fuel assemblies. Criticality analyses of burnt fissile material using burnup credit can be managed within a reasonable amount of time, on condition that sufficient proofs of code qualification for the studied range of parameters are provided, especially on fission products isotopic compositions calculation. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

Evaluation of an unexpected reflector temperature effect for water-isolated array configurations - 14500

It has been conventional wisdom in the criticality safety discipline that for spent fuel systems like water pit storage and transportation, modeling water at its peak density will maximize reactivity. Thus, modeling water at 3.96 deg. C will produce a conservatively bounding k{sub eff} for that configuration. This assumption has been found to be incorrect for certain high-leakage water-isolated array configurations. In fact, temperatures up to the boiling point of water can in fact be more reactive, and the most reactive temperature is highly dependent upon the specific fuel configuration being analyzed. In light of the accident at Fukushima-Daiichi - where the spent fuel pool temperature increased - knowledge of this unexpected behavior is essential to ensuring that criticality safety evaluations are conservatively bounding. To illustrate the reflector temperature effect, MC21 calculations are provided for a sensitivity study based on LEU-COMP-THERM-044, which models the IPEN/MB-01 reactor at various temperatures between 3.9 deg. C and 93.3 deg. C. (authors)

42 ENGINEERING↗
DOE OSTI2015

Fissile material found in electrical junction box at nuclear fuel services, Inc. - 14569

On October 29, 2014 at Nuclear Fuel Services, Inc., while troubleshooting a failed heater control circuit, fissile material was found in a small electrical junction box that contained a Resistance Temperature Detector (RTD). The RTD was mounted inside a stainless steel thermowell located in a uranium recovery boildown loop. Over time, the end of the thermowell had become pitted allowing fissile solution to penetrate the thermowell and collect in the electrical junction box. Conduit connected the small electrical junction box to an unfavorable geometry electrical panel, providing a credible scenario where an unsafe volume of fissile liquid could potentially collect in an unfavorable geometry. This postulated upset scenario had not been analyzed by Nuclear Criticality Safety (NCS). Prior to restart of the operation, a corrective action was taken to replace the damaged thermowell. Additionally, a review of other fissile systems was completed to identify all unfavorable geometry electrical panels where the potential for a similar accident scenario existed. Drain holes were drilled in the bottom of all those identified. Analysis of the damaged thermowell is ongoing. Expanded details of the incident, the short term and planned long term corrective actions, and lessons learned are discussed. (authors)

42 ENGINEERING↗
DOE OSTI2015

Overview of zion comparative analysis project for assessment of Boral{sup R} neutron absorber material performance and monitoring in spent fuel pools - 14618

Neutron absorber materials are used in spent fuel pool (SFP) storage racks to control reactivity. In the United States, about half of the operating plants currently use BORAL{sup R} neutron absorber material in SFP racks. Although operational experience spanning more than 25 years has not shown any loss of the BORAL{sup R} core material, corrosion effects, including blisters and pits, have been observed and reported. To improve understanding for the behavior of BORAL{sup R} in SFPs and evaluate the performance of monitoring approaches, EPRI initiated the Zion comparative analysis project in 2014. The Zion project involves three phases. During the first phase of the project, in situ measurements were performed on selected SFP rack panels in December 2014. Panels for in situ measurements were selected after an extensive analysis of each panel in the pool to ensure that the selected panels represented the pool in terms of variation in neutron and gamma dose values. As part of the second phase of the project, the remaining 10 coupons from the Zion SFP were shipped to a Penn State University (PSU) laboratory for analysis in January 2015. During the final phase of the project, a subset of the panels for which in situ measurements were performed will be harvested from the Zion SFP, cut into samples, and analyzed at PSU. The overall scope of this project will allow comparison of in situ measurements to real panel data as well as data obtained from coupon analysis. In this paper, an overview of the Zion comparative analysis project, the criteria and procedure for panel selection for in situ measurements and harvesting, preliminary coupon analysis results and in situ measurement results are presented. (authors)

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2024

Effect of Minor Ce Additions on Corrosion Behavior of Experimental Pipeline Steel

Minor additions of rare earth elements (REEs) have been shown to improve various steel properties. In particular, Cerium (Ce), as a relatively abundant and inexpensive element, has received attention as a potentially beneficial minor alloying addition for a variety of steel applications, including pipelines materials. The objective of this work is to investigate the effect of Ce on corrosion performance of experimental high-strength pipeline steel. Electrochemical corrosion experiments were carried out on steel without and with 0.006, 0.018, and 0.03 wt.% of Ce in 3.5 wt.% NaCl saturated with CO2 at 20 oC. For comparison purposes, X100 carbon steel specimens machined from an experimental pipe with a yield strength of 731 MPa were tested under the same conditions. The electrochemical techniques employed included: potentiodynamic polarization (PDP), linear polarization resistance (LPR), and electrochemical impedance spectroscopy (EIS). In addition, post-corrosion surface characterization was performed using scanning electron microscopy (SEM) equipped with energy dispersive X-ray spectroscopy (EDS), while crystalline phases were determined by X-ray diffraction (XRD). The corrosion of base metal without Ce was lower (0.15 mm/y) compared to carbon steel X100 (0.9 mm/yr). The electrochemical results showed that adding Ce accelerates corrosion rates compared to base metal without Ce. No pitting corrosion of tested steel was detected.

Belarbi, Zineb↗
DOE OSTI2024

Effect of Drying on Corrosion Mitigation of Hanford Transfer Lines

Radioactive waste is stored in underground, carbon-steel double-shell tanks at the Department of Energy Hanford site. The waste is transferred between the tanks and other assets using the transfer lines spanning throughout the various tank farms at Hanford. The transfer lines consist of a pipe-in-pipe design, small diameter pipes, and are not piggable. Recent inspection data of the transfer lines have shown areas with corrosion on both interior of the encasements and exterior of the primary pipes, with nearly 50 percent wall loss on the primary pipes and nearly 25% wall loss on the encasement pipes due to pitting corrosion. The visual inspections of the transfer lines have shown presence of corrosion products near the pipeline risers and beyond. It has been hypothesized that the corrosion is predominantly due to the high humidity conditions and in some cases is driven by the presence of residual hydrotest water in the encasement and the associated contact with the safety significant primary pipe. Therefore, drying of the transfer lines could lead to corrosion mitigation. Experimental studies are being conducted to understand the effect of environmental conditions, especially, relative humidity and temperature, on transfer line grade carbon steel corrosion and on mitigating corrosion. The experimental conditions are selected based on the seasonal temperature changes, and relative humidity conditions ranging from 30 to 100 percent. The experimental data will be used as guidance for maintaining a dry environment that will help mitigate the transfer-line corrosion caused by the high humidity conditions.

Shukla, Pavan K.↗
DOE OSTI2024

Feasibility Study for Cathodic Protection of Waste Tanks at Hanford

The Hanford Nuclear Reservation site contains approximately 211 million liters of radioactive and chemically hazardous waste arising from nuclear weapons production, beginning with World War II, and continuing through the Cold War. The waste is stored in 177 carbon-steel underground storage tanks, of which 149 are single-shell tanks (SSTs) and the remaining are double-shell tanks (DSTs). The mission of an ongoing River Protection Project is to retrieve the waste from the underground storage tanks and then treat and immobilize (i.e., vitrify) it for disposal. Waste from the older SSTs is being progressively retrieved into the newer DSTs for storage pending treatment, immobilization, and disposal. Waste chemistry controls are in place to mitigate general corrosion, pitting corrosion, and stress corrosion cracking (SCC) and minimize corrosion risks to the carbon steel tank liners of DST at the Hanford Site. The chemistry control program focuses on preserving liner integrity while balancing the need for caustic/nitrite additions that have large downstream impacts on vitrification. Some of the DSTs store salt cake or sludge waste layers contain out-of-specification interstitial liquids that contact the primary tank bottom or sidewall. Remediation of such layers through chemical additions is not practical. At the present time, the most significant leak integrity concern at Hanford is for corrosion of the tank bottoms of DSTs where the solids layer is out of specification because of hydroxide ion depletion and consequently, pH values that are less than 12. Cathodic protection (CP) has been proposed as a possible way to protect the liners of such tanks against corrosion. The objective of this project was to investigate the feasibility of applying CP to Hanford DSTs to mitigate internal corrosion of the carbon steel tank primary liners. The scope of the project included (1) a literature review on the application of CP to storage tank systems, in general, and to waste tanks at the Savannah River Site (SRS) and Hanford, (2) a review of the finite element analysis (FEA) model recently developed to evaluate potentials for a DST, and (3) the development of a boundary element analysis (BEA) model for CP for a DST. Tank AN-107, one of the Hanford DSTs, was selected for this work because (1) the bottommost waste layer in the tank is out-of-specification, and (2) electrochemical and resistivity data, based on recent waste sampling and testing, are available for this tank. The overarching conclusion of this study is that, based on the available information, it is feasible to apply CP to mitigate internal corrosion of the primary liners of the Hanford DSTs. However, significant additional information and work are required before a CP system can be designed. Recommendations for next steps in the design of the CP system will be presented.

WIERSMA, BRUCE↗
DOE OSTI2024

Study of Mechanical Grinding Effects on Niobium Surface

Mechanical grinding is commonly employed to eliminate surface defects such as scratches and pits from niobium cavity surfaces or sheets before cavity fabrication. Subsequently, chemically buffered polishing or electropolishing is often utilized to completely remove residues of the polishing media and any defects induced by mechanical grinding, ensuring a pristine surface. In this study, we conducted a systematic investigation to assess the influence of mechanical grinding using silicon carbide and aluminum oxide polishing media on niobium surfaces. Additionally, the study examines the effects of post-mechanical grinding chemical treatments on surface quality.

Chouhan, V.↗
DOE OSTI2025

FY24 Laboratory Directed Research and Development Annual Report

The Laboratory Directed Research and Development (LDRD) program yields foundational scientific research and development (R&D) essential to growing SRNL’s core competencies, in alignment with SRNL’s Strategic Plan to provide long-term benefits to the Department of Energy (DOE), the National Nuclear Security Administration (NNSA), and other customers and stakeholders. Five strategic goals are outlined in SRNL’s strategic plan: 1) Provide applied science and engineering for EM’s active clean-up sites and LM’s post closure management sites 2) Provide science-based solutions for gaps identified in nonproliferation strategic vision and support the government in activities impacting national security 3) Lead Science, Technology & Engineering as the central technical authority for processing tritium loaded reservoirs and support production of plutonium pits 4) Align science and energy security programs by focusing modern modeling, simulation, and data analytics tools on materials engineering and performance applications 5) Build a workforce for the future

Clark, Sue [Savannah River National Laboratory (SR↗
DOE OSTI2026

MARVEL Reactor Fuel Performance Report

The Microreactor Applications Research Validation and EvaLuation (MARVEL) project is producing a high temperature liquid metal-cooled nuclear test bed at Idaho National Laboratory (INL) to ultimately improve the integration of microreactors to end-user applications. This ambitious effort seeks to design, authorize, construct, test, and operate the reactor within five years. In order to construct and operate the MARVEL reactor in a timely manner, the system will utilize materials and component designs which have already been used, qualified, or licensed from previous reactors. The MARVEL reactor will be located at the INL Transient Reactor Test (TREAT) facility in the north high-bay equipment pit and will use the existing 304 stainless steel-clad U-ZrH1.6 pin-type fuel system developed by General Atomics and purchased from TRIGA International. This fuel has been previously qualified under the United States Department of Energy’s (US DOE) Reduced Enrichment for Research and Test Reactors (RERTR) Program. Even though the regulator of the MARVEL reactor is the US DOE, the standards and overall approach recommended by the Nuclear Regulatory Commission is well-defined and utilized here. Following NUREG-1537 regulatory guidance, this report documents the authorization case for the MARVEL fuel system’s application to MARVEL and establishes stable and predictable fuel performance during the most thermophysically unfavorable conditions achievable in the MARVEL reactor. To that end, this report provides a comprehensive survey of the known thermophysical properties, performance, and quantitative relationships associated with the MARVEL reactor fuel element and uses this information to determine its mechanical integrity and risk of reaching unacceptable conditions during the most extreme accident scenarios predicted for the reactor using the most conservative assumptions available. The information contained herein is compiled from a combination of historical reports and peer reviewed scientific publication manuscripts. Known mechanisms under which the fuel is susceptible to failure are highlighted and compared to conditions that could exist in the MARVEL reactor during an unanticipated transient or accident scenario. The two scenarios considered for analysis in this report are (1) an unprotected loss of flow accident and (2) a hypothetical unprotected loss of coolant accident during the loss of flow accident. Preliminary 2D steady-state analyses herein indicate that both fuel-cladding chemical interactions and fuel-cladding mechanical interactions are negligible throughout the fuel’s operational cycle under both normal and high temperature accident scenario conditions. Although higher fidelity 3D time-dependent modeling and simulations are planned, the following may be concluded presently. The MARVEL fuel element maintains its geometric stability and structural integrity during the most extreme accident scenarios predicted for the MARVEL reactor. The hoop stress during the unprotected loss of flow accident reaches about -1.27 MPa; this negative stress indicates that it is compressive rather than tensile. The compressive stress is a result of the NaK pressure on the outside of the fuel element, caused by the restricted thermal expansion of the NaK coolant, exceeding the internal pressures generated inside of the fuel element. The hoop stress generated in the cladding during the unprotected loss of coolant accident reaches a maximum of about approximately 10 MPa, which is nearly an order of magnitude less than the predicted yield strength of the cladding under high-temperature accident scenario conditions. Calculations were compared with results from high performance computational simulations using BISON and are in very close agreement. A conservative MARVEL fuel meat peak temperature limit of 900 °C is recommended presently, which is about 180 °C higher than the peak fuel temperature predicted to occur during the most extreme accident. Based on the known properties and behavior of the MARVEL fuel element, the fuel successfully meets its design and safety requirements under normal and most extreme accident conditions with a large safety margin.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

ECAR-5363 Rev 0 MARVEL Environmental Assessment Inhalation Dose Consequence Calculations

The following dose consequence analysis is performed to support the Microreactor Applications Research Validation and Evaluation Project (MARVEL) environmental assessment. The purpose of the MARVEL project is to develop a nuclear microreactor applications test bed at Idaho National Laboratory (INL) to perform research and development on various operational features of microreactors to ultimately improve integration of microreactors to end-user applications. The MARVEL concept is a 100-kilowatt thermal (kWth) and approximately 20-kilowatt electric (kWe) generating microreactor. The MARVEL system will be located in the INL TREAT facility in the north high bay equipment pit. This ECAR analyzes a maximum reasonably foreseeable accident for MARVEL as required by “Recommendations for Analyzing Accidents under NEPA.” Consistent with the recommendations from this guidance, this ECAR considers the median air dispersion parameters for INL in evaluation accident releases from the MARVEL reactor concept and documents the dose from the maximum reasonably foreseeable accident.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2024

ECAR-5127 Rev 1 Evaluation of the MARVEL Reactor Inhalation Dose Consequences

The following dose consequence analysis is performed to support the preliminary Microreactor Applications Research Validation and Evaluation Project (MARVEL) design effort. The purpose of the MARVEL project is to develop a nuclear microreactor applications test bed at Idaho National Laboratory (INL) to perform research and development on various operational features of microreactors to ultimately improve integration of microreactors to end-user applications. The MARVEL concept is a 100 kilowatt (kW) thermal (kWth) and approximately 20-kW electric (kWe) generating microreactor. The MARVEL system will be located in the INL TREAT facility in the north high-bay equipment pit. This ECAR considers the bounding release fractions of the MARVEL reactor concept and documents the dose consequences to the public and collocated worker.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2023

ECAR-7448 Rev 0 MARVEL Post-Shutdown Dose Analysis in the Upper Confinement

This report documents a bounding post-shutdown dose rates over time in the upper confinement area and above the top of the TREXc Pit Shield at the end of two years of operation of the MARVEL microreactor. The dose rates take into consideration the doses from the core fission products, the sodium-potassium coolant (NaK), the steel components, the Stirling engines, and the gallium eutectic (EGaInSn) secondary coolant. The MCNP model in this analysis is based on the 90% design of the MARVEL reactor from ECAR- 7447. Neutron fluence used as inputs to this analysis were calculated in ECAR-7447.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2025

SPC-71260 Rev 0 MARVEL Heat Extraction Subsystem Secondary Coolant Equipment (SCE) Design/Build

A. The Microreactor Applications Research Validation and Evaluation (MARVEL) reactor will offer experimental capabilities that are not currently available at DOE’s national laboratories. Idaho National Laboratory (INL), operated for the U.S. Department of Energy (DOE) by Battelle Energy Alliance, LLC (BEA) (Contractor hereafter) is procuring services for the design, analysis, fabrication, testing and delivery of a Secondary Coolant Equipment system (SCE). This specification contains the requirements for design, analysis, fabrication, testing and delivery of the SCE as described herein. The MARVEL reactor is a microreactor which uses eutectic sodium-potassium alloy (NaK) as a primary coolant. The primary coolant is circulated through four primary loops by natural convection of the coolant. In each loop is a closed well which will accommodate an intermediate heat exchanger (IHX) for extracting heat from the loop. These wells will be referred to in this specification as the “IHX wells.” It is intended for the IHX containment to also be filled with NaK. The MARVEL design team has determined that a Heat Extraction System (HES) using pumped NaK will be used to extract heat from the IHXs and deliver it to a downstream system for power generation or alternate process heat users. This Heat Extraction System will enable MARVEL operations including the ability to test, demonstrate, and address issues related to installation, startup, and operations. In addition, it will allow down-stream utilization of process heat for various uses. The objective of this specification is to develop the final design for the HES Secondary Coolant Equipment system (SCE) that will be used as the core of the HES. This system provides control of the NaK circulation between the MARVEL reactor and the subsequent process heat utilization systems. It does not include design of the Intermediate Heat Exchangers and piping inside the T-REXc pit in which the reactor is located. B. The MARVEL microreactor will be installed in the Transient Reactor Test Facility (TREAT) building in the Transient Reactor Test (TREAT) Micro-Reactor Experiment Cell (T-REXc) C. An INL Subcontractor has developed a conceptual design for this system per SPC-71145, referred to in that specification as the Process Heat Extraction System. SPC-71260 is based on the pumped NaK loop concept developed under SPC-71145. D. The SCE system design and (as option scope) fabrication shall be provided by the awardee of the subcontract (Subcontractor hereafter) pertaining to this Specification. Prior to shipment, the SCE will be fabricated, assembled, and tested at the Subcontractor’s facility. After successful completion of acceptance testing, the SCE and associated equipment will be shipped to the Materials and Fuels Complex (MFC) at the INL (Contractor’s Facility hereafter) to be installed by others in TREAT/T-REXc.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2018

Ion Dynamics in Ionic-Liquid-Based Li-Ion Electrolytes Investigated by Neutron Scattering and Dielectric Spectroscopy

A detailed understanding of the diffusion mechanisms of ions in pure and doped ionic liquids remains an important aspect in the design of new ionic-liquid electrolytes for energy storage. Here, to gain more insight into the widely used imidazolium-based ionic liquids, the relationship between viscosity, ionic conductivity, diffusion coefficients, and reorientational dynamics in the ionic liquid 3-methyl-1-methylimidazolium bis(trifluoromethanesulfonyl)imide (DMIM-TFSI) with and without lithium bis(trifluoromethanesulfonyl)imide (Li-TFSI) was examined. The diffusion coefficients for the DMIM + cation and the role of ion aggregates were investigated by using the quasielastic neutron scattering (QENS) and neutron spin echo techniques. Two diffusion mechanisms are observed for the DMIM + cation with and without Li-TFSI, that is, translational and local. The data additionally suggest that Li + ion transport along with ion aggregates, known as the vehicle mechanism, may play a significant role in the ion diffusion process. These dielectric-spectroscopy investigations in a broad temperature and frequency range reveal a typical α–β-relaxation scenario. The α relaxation mirrors the glassy freezing of the dipolar ions, and the β relaxation exhibits the signatures of a Johari–Goldstein relaxation. In contrast to the translational mode detected by neutron scattering, arising from the decoupled faster motion of the DMIM + ions, the α relaxation is well coupled to the dc charge transport, that is, the average translational motion of all three ion species in the material. Finally, the local diffusion process detected by QENS is only weakly dependent on temperature and viscosity and can be ascribed to the typical fast dynamics of glass-forming liquids.

25 ENERGY STORAGE↗