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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

DOC-WMF - A Generic Model for Assessment of Waste Disposal Facilities - 19278

Numerical models and computer codes play an important role in the safety assessment of nuclear waste disposal facilities, as well as in the regulatory review of proponents' safety assessments. An important part of a safety assessment is the radiation dose calculation. This paper presents a computer code called DOC-WMF for calculating radiation dose from nuclear waste disposal facilities. DOC-WMF is a generic code that could be used for dose calculations for deep geological or near surface disposal facilities of nuclear wastes. DOC-WMF is much faster than complicated 3-D numerical models in terms of computational speed since the model uses the commercial code GoldSim as the computation engine which is based on analytical solutions for contaminant transport. As a result, it enables a user to quickly assess the dose consequences of multiple options or scenarios for a nuclear waste disposal facility. DOC-WMF was adapted from, and an improvement to, the Scoping of Options and Analyzing Risk (SOAR) model developed by the United States Nuclear Regulatory Commission (US NRC). A key improvement over the SOAR model is that DOC-WMF allows the calculation of dose contributions from other pathways such as food ingestion and air inhalation in addition to the drinking water pathway which is the sole dose contribution considered in the SOAR model. DOC-WMF also allows for: the consideration of 55 radionuclides with option to add more if required; the choice of source release method; various pathways to the biosphere; and dose calculations including an extended array of exposure pathways. In addition, DOC-WMF can simulate more complex hydrogeological settings and waste forms than SOAR does. DOC-WMF was compared against known analytical solutions for verification. It was then applied to two test cases and the results were compared. The first was a hypothetical deep geological disposal repository. The second test case was a near-surface nuclear waste disposal facility. DOC-WMF independently modeled these two test cases and achieved similar results as the more complicated numerical models did. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2015

DRACS thermal performance evaluation for FHR

Direct Reactor Auxiliary Cooling System (DRACS) is a passive decay heat removal system proposed for the Fluoride-salt-cooled High-temperature Reactor (FHR) that combines coated particle fuel and a graphite moderator with a liquid fluoride salt as the coolant. The DRACS features three coupled natural circulation/convection loops, relying completely on buoyancy as the driving force. These loops are coupled through two heat exchangers, namely, the DRACS Heat Exchanger and the Natural Draft Heat Exchanger. In addition, a fluidic diode is employed to minimize the parasitic flow into the DRACS primary loop and correspondingly the heat loss to the DRACS during normal operation of the reactor, and to keep the DRACS ready for activation, if needed, during accidents. To help with the design and thermal performance evaluation of the DRACS, a computer code using MATLAB has been developed. This code is based on a one-dimensional formulation and its principle is to solve the energy balance and integral momentum equations. By discretizing the DRACS system in the axial direction, a bulk mean temperature is assumed for each mesh cell. The temperatures of all the cells, as well as the mass flow rates in the DRACS loops, are predicted by solving the governing equations that are obtained by integrating the energy conservation equation over each cell and integrating the momentum conservation equation over each of the DRACS loops. In addition, an intermediate heat transfer loop equipped with a pump has also been modeled in the code. This enables the study of flow reversal phenomenon in the DRACS primary loop, associated with the pump trip process. Experimental data from a High-Temperature DRACS Test Facility (HTDF) are not available yet to benchmark the code. A preliminary code validation is performed by using natural circulation experimental data available in the literature, which are as closely relevant as possible. The code is subsequently applied to the HTDF that is under construction at the Ohio State University.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Analytic Solution for Time-Dependent Neutron Transport with Delayed Neutrons via Singular Eigenfunction Expansion

An important way to verify the proper functioning of a computer code, and its underlying numerical schemes, is by comparing numerical results generated with the code to an analytic solution. To this end, we present an analytic solution for time-dependent neutron transport with delayed neutrons, which can be used to verify computer codes for modeling reactor kinetics phenomena. To develop our analytic solution, we use the singular eigenfunction expansion method. Ganapol et al. have also developed a solution to this problem, but by using the multiple collision technique. Our approach has certain advantages with respect to this previous work, in particular that it yields a closed form (i.e., in terms of an integral rather than an infinite series) explicit expression for the angular flux (i.e., instead of just the scalar flux or the angular flux at a single spatial point). In addition, Su and Olson have presented an analytic solution for a very similar problem: non-equilibrium radiative transfer. Their approach involves a Laplace transform, which they invert by showing there are no singularities in the right complex plane of the Laplace transform parameter and then simply choosing the Bromwich contour as the imaginary axis. To extend this previous work to our case, we would further need to exclude supercritical problems in order to use the same contour. The resulting analytic solution would then be of limited interest. In contrast, our analytic solution is completely general with respect to sub, delayed super, or prompt supercriticality. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2016

Measurements of Neutron Importance Distribution in {sup 235}U Fast Critical Reactor

The reactor physics related integral experiments are significant to criticality safety analysis and the design of the next generation nuclear system. Measurements of parameters such as prompt neutron decay constant α, prompt neutron generation time Λ, effective delayed neutron fraction β{sub eff}, neutron importance distribution, etc., can provide data for the verification of calculation methods, computer codes and cross-section data in criticality safety analysis and test the design of new reactors. Among these parameters, neutron importance distribution has its unique value. Beside verification and validation of nuclear data, computer codes, and simulation methods, the experimental results of neutron importance distribution are also useful to the correcting of the spatial effects in Rossi-α measurement, which is based on point-reactor neutron kinetics. The relative neutron importance distribution experiments were performed on the {sup 235}U fast critical reactor of China Academy of Engineering Physics (CAEP). The key experimental objective was to obtain the importance distribution data for nuclear data validation and criticality safety analysis. We demonstrate in this paper the experiment and the data analysis. {sup 252}Cf neutron importance spatial distributions in the diametral hole of a depleted uranium reflected {sup 235}U fast critical reactor were measured. The evaluated uncertainties of the experiment data were varying with the source location but within 1.5% ranges in all. The experiment results can provide useful data for applications like nuclear data validation and the determination of rectification factor R of Rossi-a measurements. However, the experimental uncertainties are not satisfying and can be further reduced by using a {sup 252}Cf source with much higher intensity, recording more precisely the absolute position of the neutron source, and carrying out the experiment in more shallow subcritical states in which the system multiplication is larger. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Golden rule kinetics of transfer reactions in condensed phase: The microscopic model of electron transfer reactions in disordered solid matrices

The algorithm for a theoretical calculation of transfer reaction rates for light quantum particles (i.e., the electron and H-atom transfers) in non-polar solid matrices is formulated and justified. The mechanism postulated involves a local mode (an either intra- or inter-molecular one) serving as a mediator which accomplishes the energy exchange between the reacting high-frequency quantum mode and the phonon modes belonging to the environment. This approach uses as a background the Fermi golden rule beyond the usually applied spin-boson approximation. The dynamical treatment rests on the one-dimensional version of the standard quantum relaxation equation for the reduced density matrix, which describes the frequency fluctuation spectrum for the local mode under consideration. The temperature dependence of a reaction rate is controlled by the dimensionless parameter ξ{sub 0}=ℏω{sub 0}/k{sub B}T where ω{sub 0} is the frequency of the local mode and T is the temperature. The realization of the computational scheme is different for the high/intermediate (ξ{sub 0} < 1 − 3) and for low (ξ{sub 0}≫ 1) temperature ranges. For the first (quasi-classical) kinetic regime, the Redfield approximation to the solution of the relaxation equation proved to be sufficient and efficient in practical applications. The study of the essentially quantum-mechanical low-temperature kinetic regime in its asymptotic limit requires the implementation of the exact relaxation equation. The coherent mechanism providing a non-vanishing reaction rate has been revealed when T→ 0. An accurate computational methodology for the cross-over kinetic regime needs a further elaboration. The original model of the hopping mechanism for electronic conduction in photosensitive organic materials is considered, based on the above techniques. The electron transfer (ET) in active centers of such systems proceeds via local intra- and intermolecular modes. The active modes, as a rule, operate beyond the kinetic regimes, which are usually postulated in the existing theories of the ET. Our alternative dynamic ET model for local modes immersed in the continuum harmonic medium is formulated for both classical and quantum regimes, and accounts explicitly for the mode/medium interaction. The kinetics of the energy exchange between the local ET subsystem and the surrounding environment essentially determine the total ET rate. The efficient computer code for rate computations is elaborated on. The computations are available for a wide range of system parameters, such as the temperature, external field, local mode frequency, and characteristics of mode/medium interaction. The relation of the present approach to the Marcus ET theory and to the quantum-statistical reaction rate theory [V. G. Levich and R. R. Dogonadze, Dokl. Akad. Nauk SSSR, Ser. Fiz. Khim. 124, 213 (1959); J. Ulstrup, Charge Transfer in Condensed Media (Springer, Berlin, 1979); M. Bixon and J. Jortner, Adv. Chem. Phys. 106, 35 (1999)] underlying it is discussed and illustrated by the results of computations for practically important target systems.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2016

Severe Accident Research Directions after Fukushima

The accidents at the Fukushima Daiichi nuclear power station in March of 2011 renewed interest globally in the nature, causes, and health and economic consequences of severe accidents involving core damage and containment impairment. As a result of these accidents, research has been conducted into the modeling and phenomenological understanding of the accidents at Fukushima and the response and effectiveness of mitigating measures such as emergency water injection, primary system depressurization and containment venting actions. An important example of this is the jointly conducted OECD-NEA organized Benchmark of the Fukushima Accidents (BSAF project) where accident reconstruction analyses using computer codes such as MELCOR and MAAP are ongoing. Even prior to the Fukushima accidents, the Sandia/NRC sponsored State of Art Reactor Consequence Analyses (SOARCA) foreshadowed the accidents in Japan where long-term and short-term station blackout accidents in boiling water reactors with Mark-1 containment were analyzed showing striking similarities to the accidents at Fukushima. These analyses also highlighted important differences in our prior understanding, in particular the importance of suppression pool thermal stratification and the extraordinary operation of the Fukushima Unit-2 reactor core isolation cooling (RCIC) system, prompting addition research into both of these areas. Additionally, the focus of the world's major computer codes on modeling the Fukushima accidents in the BSAF effort, led to several 'Crosswalk' comparisons between code predictions, highlighting subtle but important modeling differences between the codes and identifying areas where modeling improvements can be made. Collectively, these research areas are aimed at improving our understanding of severe accident progression and real-world response of safety related systems and components, and should lead to improved accident response and mitigation capabilities in the future. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Group Structure Optimization Using the PyGroup Code

Group constants are used as coefficients in the energy domain discretized Linear Boltzmann Equation (LBE), in which the Continuous-Energy (CE) fluxes are also integrated over energy bins and transformed into group fluxes. This procedure, based on the Multi-Group (MG) theory, is widely applied in radiation transport and diffusion simulations Careful preparation of MG cross section data, or group constants, is thus a prerequisite for success in acquiring high-fidelity LBE solutions. Generation of MG cross section data itself is a complicated multi-step task, since group constants are indispensably coupled with the solution of the LBE, i.e. the flux spectrum in the problem to be solved. Theoretically, a MG cross section library is only valid for the problem that it is built upon. Nevertheless, in practice MG cross section libraries are prepared for similar type of problems, such as thermal reactors, fast reactors, etc., to facilitate the neutronic simulations. Generally, CE point-wise cross section data set, generated by both theoretical models and experimental data, is first converted to Fine-Group (FG) libraries using a pre-defined or precalculated flux spectra, then, Broad-Group (BG) libraries are collapsed from the fine-group libraries using problem-dependent flux spectra. Proper treatments of energy and spatial self-shielding and resonance cross section are required in the CE-to-MG converting process Previously, we have developed a computer code, YGROUP, for fine-to-broad group cross section collapsing. To maintain high accuracy of the resulting BG libraries, YGROUP provides users several options on collapsing strategies and weighting schemes for group binning and evaluation of BG cross section data in users' specific problems.. An approach based on Particle Swarm Optimization was applied to automatically determine the optimal group structure. We also developed a hybrid weighting scheme to further improve BG library performance for Special Nuclear Materials (SNM) detection applications. In this work, a new computer code, PyGroup, has been developed to implement the PSO approach with an improved simulation-driven fitness function design that includes both response and forward/adjoint flux errors. Together with YGROUP, and deterministic transport code TITAN/PENMSHXP, the PyGroup code package is tested on a Dual Range Coincident Counter (DRCC) model. In this paper, we demonstrated the capability of the PyGroup code for group structure optimization using the DRCC detector model. In the future, an iteration scheme can be implemented to determine the number of broad groups based on a user-defined tolerance. we will also apply PyGroup to more complicated reactor models, and investigate physics-based fitness function designs to further reduce computational cost. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2025

NeuroCoreX: Brain-Inspired Computing from Code to Circuit

NeuroCoreX is an open-source codebase that enables the implementation of brain-inspired, energy-efficient neuromorphic computing models on FPGA hardware. Designed to support real-time learning, all-to-all neural connectivity, and flexible network architectures, NeuroCoreX offers a hands-on, accessible platform for exploring biologically inspired models of neural computation. It empowers researchers, students, and developers to implement and experiment with adaptive systems—bringing the power of neuromorphic computing to a broader community through a low-cost, scalable, and reconfigurable framework.

Gautam, Ashish [Oak Ridge National Laboratory (ORN↗
DOE OSTI2016

DSD2D-FLS 2010: Bdzil's 2010 DSD Code Base; Computing tb and Dn with Edits to Reduce the Noise in the Dn Field Near HE Boundaries

The full level-set function code, DSD3D, is fully described in LA-14336 (2007) [1]. This ASCI-supported, DSD code project was the last such LANL DSD code project that I was involved with before my retirement in 2007. My part in the project was to design and build the core DSD3D solver, which was to include a robust DSD boundary condition treatment. A robust boundary condition treatment was required, since for an important local “customer,” the only description of the explosives’ boundary was through volume fraction data. Given this requirement, the accuracy issues I had encountered with our “fast-tube,” narrowband, DSD2D solver, and the difficulty we had building an efficient MPI-parallel version of the narrowband DSD2D, I decided DSD3D should be built as a full level-set function code, using a totally local DSD boundary condition algorithm for the level-­set function, phi, which did not rely on the gradient of the level-­set function being one, |grad(phi)| = 1. The narrowband DSD2D solver was built on the assumption that |grad(phi)| could be driven to one, and near the boundaries of the explosive this condition was not being satisfied. Since the narrowband is typically no more than10*dx wide, narrowband methods are discrete methods with a fixed, non-­resolvable error, where the error is related to the thickness of the band: the narrower the band the larger the errors. Such a solution represents a discrete approximation to the true solution and does not limit to the solution of the underlying PDEs under grid resolution.The full level-­set function code, DSD3D, is fully described in LA-14336 (2007) [1]. This ASCI-­supported, DSD code project was the last such LANL DSD code project that I was involved with before my retirement in 2007. My part in the project was to design and build the core DSD3D solver, which was to include a robust DSD boundary condition treatment. A robust boundary condition treatment was required, since for an important local “customer,” the only description of the explosives’ boundary was through volume fraction data. Given this requirement, the accuracy issues I had encountered with our “fast-­tube,” narrowband, DSD2D solver, and the difficulty we had building an efficient MPI-parallel version of the narrowband DSD2D, I decided DSD3D should be built as a full level-­set function code, using a totally local DSD boundary condition algorithm for the level-­set function, phi, which did not rely on the gradient of the level-­set function being one, |grad(phi)| = 1. The narrowband DSD2D solver was built on the assumption that |grad(phi)| could be driven to one, and near the boundaries of the explosive this condition was not being satisfied. Since the narrowband is typically no more than10*dx wide, narrowband methods are discrete methods with a fixed, non-resolvable error, where the error is related to the thickness of the band: the narrower the band the larger the errors. Such a solution represents a discrete approximation to the true solution and does not limit to the solution of the underlying PDEs under grid resolution.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

An approach to model reactor core nodalization for deterministic safety analysis

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Method for computing self-consistent solution in a gun code

Complex gun code computations can be made to converge more quickly based on a selection of one or more relaxation parameters. An eigenvalue analysis is applied to error residuals to identify two error eigenvalues that are associated with respective error residuals. Relaxation values can be selected based on these eigenvalues so that error residuals associated with each can be alternately reduced in successive iterations. In some examples, relaxation values that would be unstable if used alone can be used.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2014

A geophysical shock and air blast simulator at the National Ignition Facility

The energy partitioning energy coupling experiments at the National Ignition Facility (NIF) have been designed to measure simultaneously the coupling of energy from a laser-driven target into both ground shock and air blast overpressure to nearby media. The source target for the experiment is positioned at a known height above the ground-surface simulant and is heated by four beams from the NIF. The resulting target energy density and specific energy are equal to those of a low-yield nuclear device. The ground-shock stress waves and atmospheric overpressure waveforms that result in our test system are hydrodynamically scaled analogs of full-scale seismic and air blast phenomena. This report summarizes the development of the platform, the simulations, and calculations that underpin the physics measurements that are being made, and finally the data that were measured. Agreement between the data and simulation of the order of a factor of two to three is seen for air blast quantities such as peak overpressure. Historical underground test data for seismic phenomena measured sensor displacements; we measure the stresses generated in our ground-surrogate medium. We find factors-of-a-few agreement between our measured peak stresses and predictions with modern geophysical computer codes.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2015

Sorption Modeling and Verification for Off-Gas Treatment

The project has made progress toward developing a comprehensive modeling capability for the capture of target species in off gas evolved during the reprocessing of nuclear fuel. The effort has integrated experimentation, model development, and computer code development for adsorption and absorption processes. For adsorption, a modeling library has been initiated to include (a) equilibrium models for uptake of off-gas components by adsorbents, (b) mass transfer models to describe mass transfer to a particle, diffusion through the pores of the particle and adsorption on the active sites of the particle, and (c) interconnection of these models to fixed bed adsorption modeling which includes advection through the bed. For single-component equilibria, a Generalized Statistical Thermodynamic Adsorption (GSTA) code was developed to represent experimental data from a broad range of isotherm types; this is equivalent to a Langmuir isotherm in the two-parameter case, and was demonstrated for Kr on INL-engineered sorbent HZ PAN, water sorption on molecular sieve A sorbent material (MS3A), and Kr and Xe capture on metal-organic framework (MOF) materials. The GSTA isotherm was extended to multicomponent systems through application of a modified spreading pressure surface activity model and generalized predictive adsorbed solution theory; the result is the capability to estimate multicomponent adsorption equilibria from single-component isotherms. This advance, which enhances the capability to simulate systems related to off-gas treatment, has been demonstrated for a range of real-gas systems in the literature and is ready for testing with data currently being collected for multicomponent systems of interest, including iodine and water on MS3A. A diffusion kinetic model for sorbent pellets involving pore and surface diffusion as well as external mass transfer has been established, and a methodology was developed for determining unknown diffusivity parameters from transient uptake data. Two parallel approaches have been explored for integrating the kernels described above into a mass-transport model for adsorption in fixed beds. In one, the GSTA isotherm kernel has been incorporated into the MOOSE framework; in the other approach, a focused finite-difference framework and PDE kernels have been developed. Issues, including oscillatory behavior in MOOSE solutions to advection-diffusion problems, and opportunities have been identified for each approach, and a path forward has been identified toward developing a stronger modeling platform. Experimental systems were established for collection of microscopic kinetics and equilibria data for single and multicomponent uptake of gaseous species on solid sorbents. The systems, which can operate at ambient temperature to 250°C and dew points from -69 to 17°C, are useful for collecting data needed for modeling performance of sorbents of interest. Experiments were conducted to determine applicable models and parameters for isotherms and mass transfer for water and/or iodine adsorption on MS3A. Validation experiments were also conducted for water adsorption on fixed beds of MS3A. For absorption, work involved modeling with supportive experimentation. A dynamic model was developed to simulate CO 2 absorption with chemical reaction using high alkaline content water solutions. A computer code was developed to implement the model based upon transient mass and energy balances. Experiments were conducted in a laboratory-scale column to determine model parameters. The influence of geometric parameters and operating variables on CO 2 absorption was studied over a wide range of conditions. This project has resulted in 7 publications, with 3 manuscripts in preparation. Also, 15 presentations were given at national meetings of ANS and AIChE and at Material Recovery and Waste Forms Campaign Working Group meetings.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Computational efficiency improvement with Wigner rotation technique in studying atoms in intense few-cycle circularly polarized pulses

We show that by introducing Wigner rotation technique into the solution of time-dependent Schrödinger equation in length gauge, computational efficiency can be greatly improved in describing atoms in intense few-cycle circularly polarized laser pulses. The methodology with Wigner rotation technique underlying our openMP parallel computational code for circularly polarized laser pulses is described. Results of test calculations to investigate the scaling property of the computational code with the number of the electronic angular basis function l as well as the strong field phenomena are presented and discussed for the hydrogen atom.

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2026

SMR safety through HTTF modeling and benchmark efforts for code validation for gas-cooled reactor applications

Accurate modeling and simulation tools for thermal-hydraulics calculations are a key element needed to design and license new advanced reactors including Small Modular Reactors (SMR) and Microreactors. Uncertainties in modeling and simulation can have significant safety and economic implications. The High Temperature Test Facility (HTTF) at Oregon State University (OSU) is a scaled integral effects experiment designed to investigate transient behavior in high-temperature gas-cooled prismatic-block nuclear reactors. High-quality measurement data is available from the HTTF that is suitable for a thermal-hydraulics code validation benchmark for gas-cooled reactor simulations. Here, this paper summarizes individual HTTF modeling efforts to date for tool validation at Idaho National Laboratory (INL), Argonne National Laboratory (ANL), Oregon State University (OSU) and Canadian Nuclear Laboratories (CNL) using system thermal-hydraulics codes, Computational Fluid Dynamics (CFD) codes and system-CFD code couplings. Also, the paper introduces the ongoing OECD Nuclear Energy Agency (NEA) High Temperature Gas Reactor Thermal-Hydraulics (HTGR T/H) benchmark that allows for better comparisons of results between different international modeling teams. The benchmark provides well defined computational problems that include code-to-code comparisons and comparisons to measured data. These problems provide an avenue for quantifying accuracy and identifying sources of uncertainty in thermal-hydraulics calculations, including in measured thermophysical properties, as part of validation for gas-cooled reactor simulation tools.

21 - SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLAN↗
DOE OSTI2017

Computational Thermodynamics of Materials Zi-Kui Liu and Yi Wang

This authoritative volume introduces the reader to computational thermodynamics and the use of this approach to the design of material properties by tailoring the chemical composition. The text covers applications of this approach, introduces the relevant computational codes, and offers exercises at the end of each chapter. The book has nine chapters and two appendices that provide background material on computer codes. Chapter 1 covers the first and second laws of thermodynamics, introduces the spinodal as the limit of stability, and presents the Gibbs-Duhem equation. Chapter 2 focuses on the Gibbs energy function. Starting with a homogeneous system with a single phase, the authors proceed to phases with variable compositions, and polymer blends. The discussion includes the contributions of external electric and magnetic fields to the Gibbs energy. Chapter 3 deals with phase equilibria in heterogeneous systems, the Gibbs phase rule, and phase diagrams. Chapter 4 briefly covers experimental measurements of thermodynamic properties used as input for thermodynamic modeling by Calculation of Phase Diagrams (CALPHAD). Chapter 5 discusses the use of density functional theory to obtain thermochemical data and fill gaps where experimental data is missing. The reader is introduced to the Vienna Ab Initio Simulation Package (VASP) for density functional theory and the YPHON code for phonon calculations. Chapter 6 introduces the modeling of Gibbs energy of phases with the CALPHAD method. Chapter 7 deals with chemical reactions and the Ellingham diagram for metal-oxide systems and presents the calculation of the maximum reaction rate from equilibrium thermodynamics. Chapter 8 is devoted to electrochemical reactions and Pourbaix diagrams with application examples. Chapter 9 concludes this volume with the application of a model of multiple microstates to Ce and Fe3Pt. CALPHAD modeling is briefly discussed in the context of genomics of materials. The book introduces basic thermodynamic concepts clearly and directs readers to appropriate references for advanced concepts and details of software implementation. The list of references is quite comprehensive. The authors make liberal use of diagrams to illustrate key concepts. The two Appendices at the end discuss software requirements and the file structure, and present templates for special quasi-random structures. There is also a link to download pre-compiled binary files of the YPHON code for Linux or Microsoft Windows systems. The exercises at the end of the chapters assume that the reader has access to VASP, which is not freeware. Readers without access to this code can work on a limited number of exercises. However, results from other first principles codes can be organized in the YPHON format as explained in the Appendix. This book will serve as an excellent reference on computational thermodynamics and the exercises provided at the end of each chapter make it valuable as a graduate level textbook. Reviewer: Ram Devanathan is Acting Director of Earth Systems Science Division, Pacific Northwest National Laboratory, USA.

36 MATERIALS SCIENCE↗
DOE OSTI2019

CIAE-ANL CRADA for SFR SAFETY ANALYSIS: Final CRADA Report

The purpose of this project is joint verification and validation of system analysis code capabilities for Sodium-cooled Fast Reactor (SFR) design and licensing. As the researchers and code developers at the Argonne National Laboratory (ANL) and China Institute of Atomic Energy (CIAE) independently build their own SFR system analysis capabilities, this collaboration facilitated exchange of information on, and evaluation of alternative approaches for, modeling the key phenomena (such as decay heat removal and natural circulation) and the safety-related components (such as pumps, heat exchangers, and outlet plenum) common in SFR systems. The analysis at ANL and CIAE also applied their respective computational capabilities and methods (in SAS4A/SASSYS-1 computer code at ANL and FR-Sdaso computer code at CIAE) on jointly identified benchmark problems for verification, and publicly available integral tests for validation, of their results. ANL team also shared the expertise on the established software quality assurance practices and the techniques for component reliability estimates to support probabilistic risk assessments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Validation Study in SAS4A Code in Simulated Mild TOP Condition

Model improvements and validation efforts for SAS4A code, which is the most advanced computer code for simulating the the primary phase of the Core Disruptive Accident (CDA) of MOX-fueled Sodium-cooled Fast Reactors, have been performed utilizing the experimental data from the CABRI programs by JAEA (Japan Atomic Energy Agency), IRSN (Institut de Radioprotection et de Surete Nucleaire), CEA (Commissariat a l'Energie Atomique et aux Energies Alternatives) and KIT (Karlsruhe Institute of Technology). In order to confirm validity of improved models of SAS4A, a systematic and comprehensive validation plan was defined. It was based on the PIRT (Phenomena Identification and Ranking Table) result for CDA of MOX-fueled fast reactors. The fuel failure position and time are the most important parameters, which dominate the analytical results in a reactor calculation. The failure position will affect the reactivity feedback due to the fuel distribution after failure and consequently would have an effect on the energy release in the initiating phase. This paper describes the preliminary result of validation study for SAS4A code in simulated Transient Over-Power (TOP) condition. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗