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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 739 records · Page 41
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Examination of Hybrid Metal Coatings for Mitigation of Fission Product Release and Corrosion Protection of LWR SiC/SiC

There is a need to increase the safety margins of current and future light water reactors (LWRs) due to the unfortunate events at Fukushima Daiichi Nuclear Plant. Safety is crucial to restore public confidence in nuclear energy, acknowledged as an economical, high-­density energy solution to climate change. The development of accident-­tolerant fuel (ATF) concepts is crucial to this endeavor. The objective of ATF is to delay the consequences of accident progression, being inset in high temperature steam and maintaining high thermomechanical strength for radionuclide retention. The use of advanced SiCf-­SiC composite as a substitute for zircaloy-­based cladding is being considered. However, at normal operations, SiC is vulnerable to the reactor coolant and may corrode at an unacceptable rate. As a ceramic-­matrix composite material, it is likely to undergo microcracking operation, which may compromise the ability to contain gaseous fission products. A proposed solution to both issues is the application of mitigation coatings for use in normal operations. At Oak Ridge National Laboratory (ORNL), three coating technologies have been investigated with industry collaborators and vendors. These are electrochemical deposition, cathodic arc physical vapor deposition (PVD hereafter) and vacuum plasma spray (VPS). The objective of this document is to summarize these processing technologies, the resultant as-­processed microstructures and properties of the coatings. In all processes, substrate constraint resulted in substantial tensile stresses within the coating layer. Each technology must mitigate this tensile stress. Electrochemical coatings use chromium as the coolant facing material, and are deposited on a nickel or carbon “bond coat”. This is economical but suffers microcracking in the chromium layer. PVD-­based coatings use chromium and titanium in both metallic form and nitrides, and can be deposited defense-­in-­depth as multilayers. This vapor method eliminated tensile stress during processing and coatings were up to ~30 μm thick without microcracking. VPS produced coatings based on Zircaloy-2, which has a proven reactor-­compatibility. The tensile stresses appearred to be partially mitigated by annealing. Analysis showed that VPS coatings required further optimizations to prevent adverse reactions with the substrate and need a minimum thickness of ~50 μm. In addition, development of coatings are constrained by neutronic depletion analysis, which clearly indicated enrichment as an issue if the coating is too thick. Based on the present work, the cathodic arc PVD technology was considered ready for the extensive testing and evaluation on cladding materials due to their ability to mitigate the excessive tensile stresses and the reasonable coating quality achieved. The VPS Zircaloy-­2 coating technology required additional development toward mitigation of issues related to the substrate reaction and porosity. In the future, PVD and VPS will have be scaled upon successful development and demonstration. Electrochemical coatings, which are proven scalability, currently require development to mitigate issues related to the tensile stress after deposition.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Discriminating Uranium Isotopes Using the Time-Emission Profiles of Long-Lived Delayed Neutrons

In nuclear nonproliferation and safeguards, detecting and accurately characterizing special nuclear material remains one of the greatest challenges. Uranium enrichment determination is typically achieved by measuring the ratio of characteristic γ -ray emissions from 235 U and 238 U. Fission also produces β -delayed neutrons, which have been used in the past to determine uranium enrichment from the time dependence of the long-lived delayed-neutron emission rate. Such measurements typically use moderated 3 He-tube detectors. We demonstrate an alternative measurement technique that employs a fast neutron active-interrogation probe and a scintillation detector to measure the enrichment of uranium using both the buildup and decay of β -delayed-neutron emission. Instead of 3 He tubes, a capture-based heterogeneous composite detector consisting of scintillating Li glass and polyvinyl toluene is constructed and used, offering a prospect to scale delayed-neutron measurements to larger detector sizes. Since the technique relies on the existing tabulated nuclear data, no calibration standards are required. It is shown that the buildup of delayed-neutron emission can be used to distinguish between uranium samples and infer the uranium enrichment level, with accuracy that rivals the method that employs the time-dependent decay of delayed-neutron emission.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2017

Effect of Delayed Energy Release on Power Normalization in Reactor Depletion Calculations

Reactor depletion calculations typically assume that the thermal power of the reactor is directly proportional to the instantaneous fission rate within the core. However, approximately 7% of the sensible energy released per fission comes from the radioactive decay of fission products, which occurs alter the fission event. Thus, the assumed direct relationship between reactor power and fission rate implies that all fission products have reached secular equilibrium within the core. In situations where fission product concentrations have not reached equilibrium (e.g., shortly after reactor startup or a change in reactor power during operation), the delayed energy release due to fission product decay creates a nonlinear relationship between reactor power and fission rate. This non-linear relationship, in turn, affects the neutron flux levels and fuel consumption rate required to maintain the desired thermal power output for the reactor. For example, during initial startup of a commercial light water reactor the apparent energy release per fission is ∼7{sup %} lower than the typical value of ∼200 MeV/fission due to the potential energy stored in radioactive fission products. As a consequence, the flux level (and, therefore, uranium consumption rate) in the reactor must be correspondingly higher to preserve the desired reactor power. As the reactor operates and fission product inventories increase, the decay heat produced by the reactor will rise, producing a corresponding increase in the apparent energy release per fission. The characteristic time-dependent behavior of energy release due to the decay of fission products (i.e., decay heat) following reactor startup, shutdown, or changes in reactor power is well-known and widely-used in reactor safety analyses. However, to our knowledge, the effects of time-dependent decay heating rates on reactor depletion calculations have not been thoroughly investigated. In this paper, we present a simple methodology for explicitly accounting for delayed energy release due to the decay of fission products in reactor depletion calculations. In the proposed method, the balance equations for fission product decay energy are incorporated into the depletion coefficient matrix, using a strategy similar to that described by Isotalo in 2016. The modified system of differential equations can be solved via a matrix exponential approximation (or other suitable numerical method) to yield solutions for both nuclide densities and decay energy emission densities as a function of time. Numerical results from the method are presented for simplified (one-group point (0-D) and 1-D) representations of a commercial pressurized water reactor (PWR), in order to estimate the effect of delayed energy release on fuel consumption and flux level immediately following reactor startup.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Production of aromatics from di- and polyoxygenates

Methods, catalysts, and reactor systems for producing in high yield aromatic chemicals and liquid fuels from a mixture of oxygenates comprising di- and polyoxygenates are disclosed. Also disclosed are methods, catalysts, and reactor systems for producing aromatic chemicals and liquid fuels from oxygenated hydrocarbons such as carbohydrates, sugars, sugar alcohols, sugar degradation products, and the like; and methods, catalysts, and reactor systems for producing the mixture of oxygenates from oxygenated hydrocarbons such as carbohydrates, sugars, sugar alcohols, sugar degradation products, and the like. The disclosed catalysts for preparing the mixture of oxygenates comprise a Ni.sub.nSn.sub.m alloy and a crystalline alumina support.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Creating a Simulation Platform for Research and Development of Advanced Control Methods

Advanced nuclear reactors are essential to meet the changing energy requirements throughout both the United States and the rest of world. In addition to other features, they are designed to enable deployment in remote locations and operate in a fully (or near-fully) autonomous manner, which will require a new control paradigm. To realize autonomously operating reactors, the U.S. Department of Energy’s Nuclear Energy Enabling Technologies Advanced Sensors and Instrumentation (NEET ASI) program conducts research and development into the enabling technologies and methods needed, including digital twins, machine learning, and risk modeling, in addition to various types of control methods. These technologies and methods are the key foundations needed to achieve fully autonomous systems. To develop and evaluate the technologies and methods necessary for achieving autonomous operations, it is critical to identify a software tool capable of integrating all the required elements. In surveying the available solutions, no software platforms were identified that could accomplish what was needed without introducing drawbacks. This challenge was the motivation for the current effort: to develop a software platform that can seamlessly integrate autonomouscontrol-enabling technologies and methods, allowing for accelerated research and development and transfer of ideas. The resulting platform, known as the Control and Optimization Modular Modeling Application for Nuclear Deployment (COMMAND), is Python-based, and leverages open-source tools to provide flexibility and facilitate building upon prior research. It is designed to enable advanced reactor developers to deploy and test advanced control technologies and methods coupled with their own models, solutions, and hardware. Given the substantial undertaking of developing such a platform, the current effort focused on laying down scalable, flexible software foundations and infrastructure, then demonstrating the platform via a use case. These foundations included developing generic modules, which contain the base variable and system blocks (the information and functional building blocks, respectively, that can be used to design a simulation) and the data handling and storage blocks needed to exchange information between the various blocks; as well as enablingtechnology-specific modules. This platform was evaluated via a use case, which was to simulate and control a process for the Microreactor Automated Control System (MACS) test bed. While MACS is not currently directly coupled to any specific microreactor physics, it was initially developed in concert with the Microreactor Applications Research Validation and Evaluation (MARVEL) microreactor, and so the MARVEL physics are used here. As part of this use case, several enabling-technology-specific blocks within COMMAND were integrated, including a proportional integral derivative (PID) control block, a Reactor Excursion and Leak Analysis Program (RELAP5-3D) block, and an anomaly detection block. The COMMAND software platform was successfully demonstrated to achieve the scalability and flexibility objectives of this effort and will be leveraged by the program’s research efforts to advance state of the art control methodologies towards autonomous operations of advanced reactors. As new use cases are created and implemented, it is anticipated that COMMAND will continue to grow and evolve to meet new requirements.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2022

Status of Eddy Current Developments at INL

Both standard and advanced Light Water Reactor (LWR) fuel systems are composed of fuel pellets enclosed in a metallic cladding. The fuel cladding represents the first containment of the fission products and the first safety barrier. Upon progression of irradiation, the cladding properties deteriorate, mainly due to the pick-up of hydrogen, which causes embrittlement, and the consequence corrosion that reduced the metallic substrate. Those changes in the properties combined with the harsh environment, e.g., fatigue and fretting occurring during operation, might eventually lead to loss of hermeticity. Therefore, it is important to be able to quantify the amount of corrosion taking place and to detect and size structural discontinuity that can promote cladding failure. Eddy current testing is one of the non-destructive methods employed for this purpose. Inspections are carried out using special probes, which are designed and optimized to inspect a tube with well-defined geometrical and physical characteristics, i.e. external diameter, wall thickness, material electric conductivity and magnetic permeability. The inspection goals are mainly twofold: first, the determination of the amount of corrosion by measurement of the external oxide thickness; second, the detection of flaws in the tube wall, ether on the external or the internal surface of the tube. This report summarizes the available capabilities for eddy current testing of LWR fuel claddings, identifies needs and are of developments for testing of ATF claddings.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Integrated simulation of U-10Mo monolithic fuel swelling behavior

Here, a separate computational branch has been implemented within the DART (Dispersion Analysis Research Tool) computational code to simulate the swelling behavior of U-10Mo monolithic fuel under the operating conditions of high-power research and test reactors (RTRs). The monolithic branch of the DART code implements a mechanistic rate-theory-based fission-gas-behavior model for the calculation of fission gas swelling, as well as a suite of thermal, physical, and mechanical models to take into account various processes occurring in RTR fuels during irradiation. In order to accurately simulate and eventually predict U-10Mo monolithic fuel irradiation behavior, the code uses materials properties calculated with lower length-scale computational methods, such as gas atom diffusivity and U-Mo surface energy from atomic simulations and grain-morphology-specific recrystallization kinetics (recrystallized fuel volume fractions vs. fission density) predicted using the phase-field method. The remainder of fission gas behavior parameters used in the model were calibrated with measured intergranular bubble size distributions. With this integrated simulation approach, the swelling behavior of U-10Mo monolithic fuel was simulated for various initial grain sizes at different operating conditions and compared with measured data. Furthermore, because limited experimental data exist for parameter calibration detailed sensitivity studies for the important parameters used in the fission gas behavior model were performed in order to examine their impact on both intergranular gas bubble morphology at low fission density, and on total porosity at high fission density.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2015

On the application of betweenness centrality in chemical network analysis: Computational diagnostics and model reduction

Here, the concept of shortest path and betweenness centrality (BC) is introduced in combustion systems, with its application demonstrated in chemical network analysis and skeletal mechanism generation. After establishing the graphical representation of a chemical network for given pressure, temperature, and species concentrations, a metric BC is defined to rank the shortest paths passed by other nodes between the pair of source and target node, and as such captures the dominant indirect kinetic pathways between any pair of nodes in the network for computational diagnostics. As a result, the controlling pathways are retained by collecting the species with larger BC values. Such a notion then indicates that the index of importance of species could then be assigned based on their BC values to further guide skeletal mechanism generation. Unlike existing methods, the betweenness centrality approach takes into account of both the fluxes between species and their relative positioning within the chemical network. To demonstrate its potential utility to combustion studies, the approach was applied to the GRI-3.0, LLNL and USC-Mech II mechanisms to identify the important pathways in the chemical network at each local reaction state, and develop skeletal mechanisms from all reaction state samples in auto-ignition and perfectly stirred reactor (PSR) simulations. The performance of the BC ranking is compared to the methods of directed relation graph (DRG), DRG with error propagations (DRGEP) and sensitivity analysis (SA), and is shown to possess sufficient utility in producing skeletal mechanisms with good accuracy and flexibility for the cases studied.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2023

Novel machine-learning method for spin classification of neutron resonances

The performance of nuclear reactors and other nuclear systems depends on a precise understanding of the neutron interaction cross sections for materials used in these systems. These cross sections exhibit resonant structure whose shape is determined in part by the angular-momentum quantum numbers of the resonances. The correct assignment of the quantum numbers of neutron resonances is, therefore, paramount. In this project, we apply machine learning to automate the quantum number assignments using only the resonances' energies and widths and not relying on detailed transmission or capture measurements. The classifier used for quantum number assignment is trained using stochastically generated resonance sequences whose distributions mimic those of real data. Here we explore the use of several physics-motivated features for training our classifier. These features amount to out-of-distribution tests of a given resonance's widths and resonance-pair spacings. We pay special attention to situations where either capture widths cannot be trusted for classification purposes or where there is insufficient information to classify resonances by the total spin J. We demonstrate the efficacy of our classification approach using simulated and actual 52 Cr resonance data.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2014

Optimized Uncertainty Quantification Algorithm Within a Dynamic Event Tree Framework

Methods for developing Phenomenological Identification and Ranking Tables (PIRT) for nuclear power plants have been a useful tool in providing insight into modelling aspects that are important to safety. These methods have involved expert knowledge with regards to reactor plant transients and thermal-hydraulic codes to identify are of highest importance. Quantified PIRT provides for rigorous method for quantifying the phenomena that can have the greatest impact. The transients that are evaluated and the timing of those events are typically developed in collaboration with the Probabilistic Risk Analysis. Though quite effective in evaluating risk, traditional PRA methods lack the capability to evaluate complex dynamic systems where end states may vary as a function of transition time from physical state to physical state . Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. A limitation of DPRA is its potential for state or combinatorial explosion that grows as a function of the number of components; as well as, the sampling of transition times from state-to-state of the entire system. This paper presents a method for performing QPIRT within a dynamic event tree framework such that timing events which result in the highest probabilities of failure are captured and a QPIRT is performed simultaneously while performing a discrete dynamic event tree evaluation. The resulting simulation results in a formal QPIRT for each end state. The use of dynamic event trees results in state explosion as the number of possible component states increases. This paper utilizes a branch and bound algorithm to optimize the solution of the dynamic event trees. The paper summarizes the methods used to implement the branch-and-bound algorithm in solving the discrete dynamic event trees.

42 ENGINEERING↗
DOE OSTI2022

Reduced Order Models Generation for HTGRs Pebble Shuffling Procedure Optimization Studies

This report provides an initial study for producing reduced-order models (ROMs) of pebble-bed high temperature gas reactor (HTGR) models for the purposes of design optimization. As an initial study, this work is meant to be exploratory---identifying useful workflows and methods for ROM generation---and not meant to be a catch-all analysis of HTGR ROM generation and usage for optimization. This report summarizes three tasks performed in Fiscal Year 2022: 1) the creation of HTGR model, 2) the sensitivity analysis of model design parameters, and 3) an introduction to ROM generation techniques. The representative HTGR model created in this work is a multiphysics equilibrium-core using the BlueCRAB (comprehensive reactor analysis bundle) reactor analysis application, coupling four physical phenomena: neutronics, streamline depletion, porous flow thermal hydraulics, and pebble heat conduction. Part of the model creation was identifying some design parameters and quantities of interest that are relevant in an optimization analysis and adjustable in the model. The sensitivity analysis utilized a polynomial chaos expansion methodology to compute global sensitivity metrics. This analysis showed that thermal hydraulics parameters and quantities of interest had a relatively small impact on simulation results. Finally, the ROM generation work involved exploring three different ROM methodologies: polynomial regression, a Gaussian process, and artificial neural networks. Using a cross-validation technique to characterize ROM performance, the Gaussian process and single-layer artificial neural networks showed the most promising results. Overall, this study was insightful and the lessons learned will be invaluable for the eventual development of an HTGR design optimization workflow.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

A LIBS method for simultaneous monitoring of the impurities and the hydrogenic composition present in the wall of the TJ-II stellarator

The study of plasma-wall interactions and impurity transport in the plasma fusion devices is critical for the development of future fusion reactors. An experiment to perform laser induced breakdown spectroscopy, using minor modifications of our existing laser blow-off impurity injection system, has been set up thus making both experiments compatible. The radiation produced by the laser pulse focused at the TJ-II wall evaporates a surface layer of deposited impurities and the subsequent radiation produced by the laser-produced plasma is collected by two separate lens and fiber combinations into two spectrometers. The first spectrometer, with low spectral resolution, records a spectrum from 200 to 900 nm to give a survey of impurities present in the wall. The second one, with high resolution, is tuned to the wavelengths of the Hα and Dα lines in order to resolve them and quantify the hydrogen isotopic ratio present on the surface of the wall. The alignment, calibration, and spectral analysis method will be described in detail. First experimental results obtained with this setup will be shown and its relevance for the TJ-II experimental program discussed.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Ultrasonic imaging in liquid sodium

The fourth generation of nuclear reactor can use liquid sodium as the core coolant. When the reactor is operating, sodium temperatures can reach up to 600 deg. C. During maintenance periods, when the reactor is shut down, the coolant temperature is reduced to 200 deg. C. Because molten sodium is optically opaque, ultrasonic imaging techniques are developed for maintenance activities. Under-sodium imaging aims at i) checking the health of immersed structures. It should also allow ii) to assess component degradation or damage as cracks and shape defects as well as iii) the detection of lost objects. The under-sodium imaging system has to sustain high temperature (up to 300 deg. C) and hostility of the sodium environment. Furthermore, specific constraints such as transducers characteristics or the limited sensor mobility in the reactor vessel have to be considered. This work focuses on developing a methodology for detecting damages such as crack defects with ultrasound devices. Surface-breaking cracks or deep cracks are sought in the weld area, as welds are more subject to defects. Traditional methods enabled us to detect emerging cracks of submillimeter size with sodium-compatible high-temperature transducer. The presented approach relies on making use of prior knowledge about the environment through the implementation of differential imaging and time-reversal techniques. Indeed, this approach allows to detect a change by comparison with a reference measurement and by focusing back to any change in the environment. It is a means of analysis and understanding of the physical phenomena making it possible to design more effective inspection strategies. Difference between the measured signals reveals the acoustic field scattered by a perturbation (a crack for instance), which may occur between periodical measurements. The imaging method relies on the adequate combination of two computed ultrasonic fields, one forward and one adjoint. The adjoint field, which carries the information about the defects, is analogous to a time-reversal operation. One of the interests of the presented method is that the time-reversal operation is not done experimentally but numerically. Numerical simulations have been carried out to validate the practical relevance of this approach. The preliminary numerical results show a nice agreement between the guessed and the actual positions of the defect. After water-tests, in sodium-tests must be done in order to validate the water/sodium transposition. For this purpose, an under-sodium device is under development, which can move the transducers with four degrees of freedom in a 1.5 m{sup 3} sodium pot. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Multi-Physics Simulation of The TRIGA Reactor

In nuclear engineering, the description of phenomena occurring in the core involves a coupling between neutronic and thermal hydraulics through the so called reactivity or thermal feedback. In water cooled reactor, thermal feedback and temperature coefficients of reactivity are important, when temperature changes, microscopic cross sections change as well impacting directly the neutron flux and so the thermal power. Other feedback, such as the variation of the macroscopic cross section due to a modification of the density can also have a significant impact on the core performance of the reactor. Those reasons were naturally leading in the development of coupling method in order to predict correctly the behaviour of the core in normal operation and accidental situation. This work describes the coupling scheme developed for the description of the TRIGA mark II reactor at JSI. The demonstration case presented here, the reactor is operating in steady state configuration at several power levels. The aim of our calculation was, for a given thermal power of the reactor, to numerically reproduce fuel temperature value monitored during reactor operation. The expected impact of the control rod on the neutron flux distribution was also verified. The predictions of the coupling scheme seem to be correct according to the few temperature measurements available and the expected neutron flux behaviour expecting from a control rod motion. We are considering precise measurement of the fuel temperature at more positions but also in the coolant channels between the fuel elements to complete the experimental validation phase. An upgrade of the coupling scheme into the transient mode is also considered to observe self-stabilisation of the power level, after a control rod perturbation, due to the negative temperature reactivity coefficient of the fuel. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Operation and Optimization of Microwave-Heated Continuous-Flow Microfluidics

Microwave (MW) technology can be powerful for electrification and process intensification but limited fundamental understanding of scalability and design principles hinders its effective use. In this work, we build a continuous-flow microreactor inside a commercial single-mode MW applicator and the corresponding computational fluid dynamics model to simulate the temperature profile. The model is in good agreement with experiments for various microreactor dimensions and operating conditions. The model indicates that MW heating is greatly influenced by reactor geometry as well as the operating parameters. We observe a strong correlation between parameters and develop a gradient boost regression tree model to predict the outlet temperature accurately. This model is then applied to optimize the dimensions and operating conditions to maximize the outlet temperature and energy efficiency, resulting in a Pareto optimal. We demonstrate computationally and experimentally that it is possible to surpass the Pareto optimal and achieve an energy efficiency of ~90% or greater at temperatures relevant for liquid-phase chemistry via salting of the solvent. The present methodology can be applied to other complex MW reactors. Lastly, the combined numerical and experimental approach provides insights into and a framework for scale-up and optimization.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2017

Effects of GdCl{sub 3} on U Electrochemical Properties in LiCl-KCl-UCl{sub 3}-GdCl{sub 3} Salt

Studies on pyro-processing technology have been stimulated in several countries including the United State, Russia, France, Japan, South Korea, and India. The main purpose of this technology is to treat the irradiated nuclear fuel, reduce volume of the nuclear waste, recycle actinides, and close the fast reactor fuel cycle. Therefore, ongoing researches and developments are expected. In pyro-processing, the most important process is an electro-refiner (ER) where uranium and minor actinides are separated on electrode surface while rare earth fission products are accumulated in LiCl- KCl eutectic salt. Due to the fact that the recovery of the U and actinide (An) materials are conducted via an electrolysis, many researchers have been investigating fundamental properties of U in LiCl- KCl system by using various electrochemical techniques. The resulting datasets are useful for effective system operation and development of safeguards in ER system. However, these values have been measured in pure UCl{sub 3} system in LiCl- KCl salt, which may not be practical to the real ER system owing to the fact that a variety of other fission products including An and lanthanide (Ln) are coexisting in the ER system. Therefore, in this work, gadolinium (Gd) was selected as a surrogate material for An or Ln materials in ER system. This is because that Gd has similar physical characteristics to Ln and An elements such as ionic size and reduction potential. By conducting electrochemical methods including cyclic voltammetry (CV), open circuit potential (OCP), and electrochemical impedance spectroscopy (EIS), effects of Gd coexistence on the electrochemical, thermodynamic, and kinetic properties of U were observed. It is hoped that this study will help us further understanding fundamental effect of the Ln/An effect on LiCl- KCl-UCl{sub 3}, yielding closer to the practical ER unit operation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Calculation of Prompt Fission Neutron Lifetimes by the Monte Carlo Method

An algorithm for calculation of prompt fission neutron lifetimes in a nuclear reactor by the Monte Carlo method is described. Evaluation of the importance function is carried out with solution of the neutron transport equation without solving the adjoint equation. The results of the prompt neutron lifetime calculations performed within some critical experiments are presented and compared with the experimental results.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Uncollided-Flux Treatment for Discrete-Ordinate Radiation Transport Solutions in the Rattlesnake Code System

We present the uncollided-flux treatment implemented in the Rattlesnake code system of the Idaho National Laboratory (INL). The ray-tracing method employed is capable of handling various types of finite-element meshes. Uncollided-flux techniques are necessary to capture streaming in low-density regions, such as the ones found in the TREAT test reactor at INL. Numerical results are presented.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗