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At least 757 records · Page 42

Comparing FEL Codes for Advanced Configurations

Various FEL codes employ different approximations and strategies to model the FEL radiation generation process. Many codes perform averaging procedures over various length scales in order to simplify the underlying dynamics. As FELs are developed in more advanced configurations beyond simple SASE, the assumptions of some codes may be called into question. Here, we compare the unaveraged code Puffin to averaged FEL codes including a new version of GENESIS in a variety of situations. In particular, we study a harmonic lasing setup, a High-Gain Harmonic Generation (HGHG) configuration modeled after the FERMI setup, and a potential Echo-Enabled Harmonic Generation (EEHG) configuration also at FERMI. Lastly, we find the codes are in good agreement, although small discrepancies do exist.

47 OTHER INSTRUMENTATION↗

High Temperature Gas Reactors: Assessment of Applicable Codes and Standards

Current interest expressed by industry in HTGR plants, particularly modular plants with power up to about 600 MW(e) per unit, has prompted NRC to task PNNL with assessing the currently available literature related to codes and standards applicable to HTGR plants, the operating history of past and present HTGR plants, and with evaluating the proposed designs of RPV and associated piping for future plants. Considering these topics in the order they are arranged in the text, first the operational histories of five shut-down and two currently operating HTGR plants are reviewed, leading the authors to conclude that while small, simple prototype HTGR plants operated reliably, some of the larger plants, particularly Fort St. Vrain, had poor availability. Safety and radiological performance of these plants has been considerably better than LWR plants. Petroleum processing plants provide some applicable experience with materials similar to those proposed for HTGR piping and vessels. At least one currently operating plant - HTR-10 - has performed and documented a leak before break analysis that appears to be applicable to proposed future US HTGR designs. Current codes and standards cover some HTGR materials, but not all materials are covered to the high temperatures envisioned for HTGR use. Codes and standards, particularly ASME Codes, are under development for proposed future US HTGR designs. A 'roadmap' document has been prepared for ASME Code development; a new subsection to section III of the ASME Code, ASME BPVC III-5, is scheduled to be published in October 2011. The question of terminology for the cross-duct structure between the RPV and power conversion vessel is discussed, considering the differences in regulatory requirements that apply depending on whether this structure is designated as a 'vessel' or as a 'pipe'. We conclude that designing this component as a 'pipe' is the more appropriate choice, but that the ASME BPVC allows the owner of the facility to select the preferred designation, and that either designation can be acceptable.

02 PETROLEUM↗

A Description of the Full Particle Orbit Following SPIRAL Code for Simulating Fast-ion Experiments in Tokamaks

The numerical methods used in the full particle-orbit following SPIRAL code are described and a number of physics studies performed with the code are presented to illustrate its capabilities. The SPIRAL code is a test-particle code and is a powerful numerical tool to interpret and plan fast-ion experiments in Tokamaks. Gyro-orbit effects are important for fast ions in low-field machines such as NSTX and to a lesser extent in DIII-D. A number of physics studies are interlaced between the description of the code to illustrate its capabilities. Results on heat loads generated by a localized error-field on the DIII-D wall are compared to measurements. The enhanced Triton losses caused by the same localized error-field are calculated and compared to measured neutron signals. MHD activity such as tearing modes and Toroidicity-induced Alfven Eigenmodes (TAEs) have a profound effect on the fast-ion content of Tokamak plasmas and SPIRAL can calculate the effects of MHD activity on the confined and lost fast-ion population as illustrated for a burst of TAE activity in NSTX. The interaction between Ion Cyclotron Range of Frequency (ICRF) heating and fast ions depends solely on the gyro-motion of the fast ions and is captured exactly in the SPIRAL code. A calculation of ICRF absorption on beam ions in ITER is presented. The effects of high harmonic fast wave heating on the beam-ion slowing-down distribution in NSTX is also studied.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

A Magnetic Diagnostic Code for 3D Fusion Equilibria

A synthetic magnetic diagnostics code for fusion equilibria is presented. This code calculates the response of various magnetic diagnostics to the equilibria produced by the VMEC and PIES codes. This allows for treatment of equilibria with both good nested flux surfaces and those with stochastic regions. DIAGNO v2.0 builds upon previous codes through the implementation of a virtual casing principle. The code is validated against a vacuum shot on the Large Helical Device (LHD) where the vertical field was ramped. As an exercise of the code, the diagnostic response for various equilibria are calculated on the LHD.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Development of ICF Mix Code

Our project over the summer of 2012 was to develop an ICF mix code. The code we developed is a simplified arbitrary Lagrange Eulerian (ALE) method written from scratch in C++ and is designed to run on unstructured grids. We have performed initial verification of our code, comparing to analytical solutions of a Sod-like shock tube problem and the Sedov blast wave problem. The code was then extended to 2D and verified with respect to the 1D solutions. We also show the capability of our code to handle turbulent instabilities, such as Richtmyer-Meshkov (RM), which are commonplace within ICF experiments. Finally we present an ICF-like initial condition setup in our 1D Lagrange Hydro code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Development and Benchmarking of a Hybrid PIC Code For Dense Plasmas and Fast Ignition (Final Report)

Radiation processes play an important role in the study of both fast ignition and other inertial confinement schemes, such as plasma jet driven magneto-inertial fusion, both in their effect on energy balance, and in generating diagnostic signals. In the latter case, warm and hot dense matter may be produced by the convergence of a plasma shell formed by the merging of an assembly of high Mach number plasma jets. This innovative approach has the potential advantage of creating matter of high energy densities in voluminous amount compared with high power lasers or particle beams. An important application of this technology is as a plasma liner for the flux compression of magnetized plasma to create ultra-high magnetic fields and burning plasmas. HyperV Technologies Corp. has been developing plasma jet accelerator technology in both coaxial and linear railgun geometries to produce plasma jets of sufficient mass, density, and velocity to create such imploding plasma liners. An enabling tool for the development of this technology is the ability to model the plasma dynamics, not only in the accelerators themselves, but also in the resulting magnetized target plasma and within the merging/interacting plasma jets during transport to the target. Welch pioneered numerical modeling of such plasmas (including for fast ignition) using the LSP simulation code. Lsp is an electromagnetic, parallelized, plasma simulation code under development since 1995. It has a number of innovative features making it uniquely suitable for modeling high energy density plasmas including a hybrid fluid model for electrons that allows electrons in dense plasmas to be modeled with a kinetic or fluid treatment as appropriate. In addition to in-house use at Voss Scientific, several groups carrying out research in Fast Ignition (LLNL, SNL, UCSD, AWE (UK), and Imperial College (UK)) also use LSP. A collaborative team consisting of HyperV Technologies Corp., Voss Scientific LLC, FAR-TECH, Inc., Prism Computational Sciences, Inc. and Advanced Energy Systems Inc. joined efforts to develop new physics and numerical models for LSP in several key areas to enhance the ability of LSP to model high energy density plasmas (HEDP). This final report details those efforts. Areas addressed in this research effort include: adding radiation transport to LSP, first in 2D and then fully 3D, extending the EMHD model to 3D, implementing more advanced radiation and electrode plasma boundary conditions, and installing more efficient implicit numerical algorithms to speed complex 2-D and 3-D computations. The new capabilities allow modeling of the dominant processes in high energy density plasmas, and further assist the development and optimization of plasma jet accelerators, with particular attention to MHD instabilities and plasma/wall interaction (based on physical models for ion drag friction and ablation/erosion of the electrodes). In the first funding cycle we implemented a solver for the radiation diffusion equation. To solve this equation in 2-D, we used finite-differencing and applied the parallelized sparse-matrix solvers in the PETSc library (Argonne National Laboratory) to the resulting system of equations. A database of the necessary coefficients for materials of interest was assembled using the PROPACEOS and ATBASE codes from Prism. The model was benchmarked against Prism's 1-D radiation hydrodynamics code HELIOS, and against experimental data obtained from HyperV's separately funded plasma jet accelerator development program. Work in the second funding cycle focused on extending the radiation diffusion model to full 3-D, continued development of the EMHD model, optimizing the direct-implicit model to speed up calculations, add in multiply ionized atoms, and improved the way boundary conditions are handled in LSP. These new LSP capabilities were then used, along with analytic calculations and Mach2 runs, to investigate plasma jet merging, plasma detachment and transport, restrike and advanced jet accelerator design. In addition, a strong linkage to diagnostic measurements was made by modeling plasma jet experiments on PLX to support benchmarking of the code. A large number of upgrades and improvements advancing hybrid PIC algorithms were implemented in LSP during the second funding cycle. These include development of fully 3D radiation transport algorithms, new boundary conditions for plasma-electrode interactions, and a charge conserving equation of state that permits multiply ionized high-Z ions. The final funding cycle focused on 1) mitigating the effects of a slow-growing grid instability which is most pronounced in plasma jet frame expansion problems using the two-fluid Eulerian remap algorithm, 2) extension of the Eulerian Smoothing Algorithm to allow EOS/Radiation modeling, 3) simulations of collisionless shocks formed by jet merging, 4) simulations of merging jets using high-Z gases, 5) generation of PROPACEOS EOS/Opacity databases, 6) simulations of plasma jet transport experiments, 7) simulations of plasma jet penetration through transverse magnetic fields, and 8) GPU PIC code development. The tools developed during this project are applicable not only to the study of plasma jets, but also to a wide variety of HEDP plasmas of interest to DOE, including plasmas created in short-pulse laser experiments performed to study fast ignition concepts for inertial confinement fusion.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Nonlinear to Linear Elastic Code Coupling in 2-D Axisymmetric Media.

Explosions within the earth nonlinearly deform the local media, but at typical seismological observation distances, the seismic waves can be considered linear. Although nonlinear algorithms can simulate explosions in the very near field well, these codes are computationally expensive and inaccurate at propagating these signals to great distances. A linearized wave propagation code, coupled to a nonlinear code, provides an efficient mechanism to both accurately simulate the explosion itself and to propagate these signals to distant receivers. To this end we have coupled Sandia's nonlinear simulation algorithm CTH to a linearized elastic wave propagation code for 2-D axisymmetric media (axiElasti) by passing information from the nonlinear to the linear code via time-varying boundary conditions. In this report, we first develop the 2-D axisymmetric elastic wave equations in cylindrical coordinates. Next we show how we design the time-varying boundary conditions passing information from CTH to axiElasti, and finally we demonstrate the coupling code via a simple study of the elastic radius.

97 MATHEMATICS AND COMPUTING↗

Micromagnetic Code Development of Advanced Magnetic Structures Final Report CRADA No. TC-1561-98

The specific goals of this project were to: Further develop the previously written micromagnetic code DADIMAG (DOE code release number 980017); Validate the code. The resulting code was expected to be more realistic and useful for simulations of magnetic structures of specific interest to Read-Rite programs. We also planned to further the code for use in internal LLNL programs. This project complemented LLNL CRADA TC-840-94 between LLNL and Read-Rite, which allowed for simulations of the advanced magnetic head development completed under the CRADA. TC-1561-98 was effective concurrently with LLNL non-exclusive copyright license (TL-1552-98) to Read-Rite for DADIMAG Version 2 executable code.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

This report documents the verification and validation (V&V) of the PLTEMP/ANL Version 4.3 code. A brief description of the code and its history are given in Section 1.1. Section 1.2 provides an overview of PLTEMP/ANL verification and validation work done to date. Section 1.3 lists some publications by other authors on PLTEMP/ANL verification and validation. Section 1.4 describes a UNIX script developed to routinely verify the code after a code modification. The V&V work documented in this report was performed over several years, as code development progressed. In each verification/validation testing discussion, the version of PLTEMP/ANL which was initially used for the V&V work is referenced, but all of the testing presented in this report is applicable to current release version 4.3 of the code.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

SAM Code Validation using the Compact Integral Effects Test (CIET) Experimental Data

The primary objective of this study is to validate the system analysis code, SAM, using experimental data from the Compact Integral Effects Test (CIET) facility. The System Analysis Module (SAM) is an advanced and modern system analysis tool being developed at Argonne National Laboratory under the U.S. DOE Office of Nuclear Energy’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. Originally aiming at supporting sodium fast reactor (SFR) design, it is now extended to cover other non-light water reactor (non-LWR) concepts, such as high-temperature gas-cooled reactor (HTGR) designs, molten salt-fueled, and molten salt-cooled reactor designs. To support SAM code development for the wide range of non-LWR applications, it is of paramount importance to validate the code against experiments relevant to these reactor concepts. The Compact Integral Effects Test (CIET) facility, which was designed to reproduce the thermal-hydraulics response of fluoride salt-cooled high-temperature reactors (FHR) under both forced- and natural-circulation conditions, has been identified and selected as one of benchmark test facilities for SAM code validation. Among available CIET experiments, three sets of them with distinctive characteristics were selected for SAM code validation purpose. The three sets experiments are: 1) Power step change transient tests; 2) DHX-DRACS natural circulation tests; and 3) heater frequency response tests. For all tests, steady-state or transient, SAM-predicted results show very good agreement with experimental data. The successful validation of SAM against these selected CIET experiments demonstrates that the computer code is well suited for thermal-hydraulics analysis of FHR designs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Application of Performance Analysis Tools on SNL ASC Codes

This milestone 1) exercised a broad set of performance profiling and analysis tools, including tools whose development has been promoted by the ASC program; 2) exercised the tools on two different SNL ASC codes, one Sierra code (Sierra/Aria, a C++ codebase) and one RAMSES code (ITS, a Fortran codebase); and 3) exercised the tools on multiple platforms, including the CTS-1 (e.g., Serrano) and ATS-1 Trinity (e.g., Mutrino) platforms. The milestone generated a plethora of strong and weak scaling, trend and profile data for multiple versions and problem cases for each of the two codes. A wealth of experience was gained with the various tools that included identification of problems, an improved understanding of feature sets, enhanced usage documentation, and insights for future tool-development. Results are provided from a large number and variety of performance analysis runs with the target codes, together with instructions for how to make use of the tools with the codes.

97 MATHEMATICS AND COMPUTING↗

Verification and Validation of the PLTEMP/ANL Code for Thermal-Hydraulic Analysis of Experimental and Test Reactors

The document compiles in a single volume the verification and validation works done for the PLTEMP/ANL code during the years of its development and improvement. The verification of sixteen capabilities of the PLTEMP/ANL Version 4.3 code that were identified by research reactor analysts as frequently used in their thermal-hydraulic analysis are described and documented. Each chapter of the document deals with the verification or validation of a specific model. The model verification is usually done by comparing the code with hand calculation, Microsoft spreadsheet calculation, or Mathematica calculation. The model validation is done by comparing the code with experimental data or a widely tested code like the RELAP5 code. In addition, some PLTEMP/ANL verification works that are available in the open literature are simply referenced and not included in the document. PLTEMP/ANL has been used in conversion safety analysis reports of several US and foreign research reactors that have been licensed and converted. A list of such reactors is given.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Oak Ridge National Laboratory Shutdown Dose Rate Code Suite

The Oak Ridge National Laboratory (ORNL) shutdown dose rate (SDDR) code suite calculates gamma dose rates at a location of interest due to activation of materials. The code suite is based on the rigorous two-step (R2S) method that involves two radiation transport calculations: (1) neutron radiation transport to determine neutron fluxes in the material for which activation calculations will be performed and (2) gamma radiation transport to determine dose rates at a location of interest, due to material activation. The ORNL SDDR code suite is used to determine an importance function that characterizes the neutron’s importance to the final SDDR when the Monte Carlo radiation transport method is used. This report demonstrates how the ORNL SDDR code suite is implemented through the following codes: Monte Carlo N-Particle (MCNP) code Version 5-1.60, Oak Ridge Isotope Generator (ORIGEN), MSX suite of utilities, Neutron Activation Gamma Source Sampler (NAGSS), and Automated Variance Reduction Generator (ADVANTG). For this demonstration, the International Thermonuclear Experimental Reactor (ITER) SDDR benchmark problem is used and the dose rate result is compared to previously published work.

61 RADIATION PROTECTION AND DOSIMETRY↗

MACCS (MELCOR Accident Consequence Code System) User Guide

The MELCOR Accident Consequence Code System (MACCS) is used by Nuclear Regulatory Commission (NRC) and various national and international organizations for probabilistic consequence analysis of nuclear power accidents. This User Guide is intended to assist analysts in understanding the MACCS/WinMACCS model and to provide information regarding the code. This user guide version describes MACCS Version 3.10.0. Features that have been added to MACCS in subsequent versions are described in separate documentation. This User Guide provides a brief description of the model history, explains how to set up and execute a problem, and informs the user of the definition of various input parameters and any constraints placed on those parameters. This report is part of a series of reports documenting MACCS. Other reports include the MACCS Theory Manual, MACCS Verification Report, Technical Bases for Consequence Analyses Using MACCS, as well as documentation for preprocessor codes including SecPop, MelMACCS, and COMIDA2. PAPERWORK REDUCTION ACT STATEMENT The NUREG does not contain information collection requirements and, therefore, is not subject to the requirements of the Paperwork Reduction Act of 1995 (44 USC 3501, et seq.). PUBLIC PROTECTION NOTIFICATION The NRC may not conduct or sponsor, and a person is not required to respond to, a request for information or an information collection requirement unless the requesting document displays a currently valid OMB control number. ACKNOWLEDGEMENTS Contributions to this User Guide were received from NRC and Sandia National Laboratories (SNL) project managers, technical experts, and code authors dedicated to the production of a valuable resource for the MACCS user community. Instructions and guidance included herein were developed over many years and include advancements in the code that provide users the ability to develop complex consequence modeling scenarios. WinMACCS and many of the early MACCS developments were due to vision of an earlier Project Manager, Jocelyn Mitchell. Jonathan Barr and AJ Nosek also contributed to the development of this report. The current NRC Project Manager, Salman Haq, provided the leadership to ensure this document was completed. Several other NRC and Sandia staff provided insights supporting development of the MACCS code and of this document.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Fast BLT Code

This note discusses numerical algorithmic software design considerations and performance estimates for a fast BLT coupling code [1] written in c++. The aim of this code is to conduct faster parameter studies over line orientations. The original matlab code was written by Mike Rivera. Art Barnes ported this code to julia. I rewrote portions of the code for speed improvement, mainly to eliminate some redundant computation when calculating many line orientations. But this code is still far from optimal.

97 MATHEMATICS AND COMPUTING↗

SAM Code Development for Transient Safety Analyses of Fluoride-salt-cooled High-temperature Reactors

The System Analysis Module (SAM) is under development at Argonne National Laboratory as a modern system-level modeling and simulation tool for advanced non-light water reactor safety analyses. It utilizes the object-oriented application framework MOOSE to leverage the modern software environment and advanced numerical methods. The capabilities of SAM are being extended to enable the transient modeling, analysis, and design of various advanced nuclear reactor systems. The molten-salt-cooled pebble-bed reactor, or pebble-bed FHR (PB-FHR) is a promising candidate among advanced nuclear reactor concepts with its improved passive safety characteristics and high thermal efficiency. To support the development and utilization of the SAM code for PB-FHR safety analysis, activities on SAM code enhancements, reference plant model developments, and code validations have been performed in the past a few years to support near-term industry and NRC needs. This report summarizes recent progress under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation program in SAM code development and demonstration for transient safety analysis of Fluoride-salt-cooled High-temperature Reactors. SAM capabilities has been significantly enhanced over the years to add FHR specific modeling features, including salt freezing and thawing, spherical core channel and pebble bed core modeling, solid-fluid thermal radiation, tritium transport and general species transport in fluids and solids, and the general code enhancements on solver schemes of point kinetics module and reactivity feedback models. A reference PB-FHR model is developed, based on publicly available information from Kairos Power’s generic FHR design and the University of California, Berkeley (UCB) Mk1 design. A reference reactor model is foundational to the methodologies employed by NRC to verify the adequacy of computer codes and evaluation models. The reference FHR model was utilized for a number of selected FHR design basis accidents, including station blackout, loss of heat sink, loss of flow, transient overpower, and overcooling events.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

ACES: Infrastructure As Code. Model Optimization and Performance Capability

Infrastructure as Code (IaC) refers to managing infrastructure (networks, physical/virtual machines, storage, and connection topology) in a descriptive model/language, rather than configuring it manually or using interactive configuration tools. Just like source code can be compiled to generate the same binary code, IaC enables generating the same environment every time it is applied. IaC is a key DevOps practice and is generally used in conjunction with continuous integration (CI) and continuous delivery (CD). In CI, all code changes are merged into a mainline branch and validated multiple times a day as developers check in their changes to the source code. In CD on the other hand, code changes are automatically packaged for a new release-to-production on a regular basis. This typically enables teams to deliver software changes much more quickly and often. Developing and managing the ACES platform using (IaC) is vital for the robust deployment and continued sustainment of this foundational computing capability. IaC and DevOps practices will help us solve many of the common challenges often encountered in developing and maintaining compute infrastructure. First, it will make the provisioning, deployment, and maintenance of the compute infrastructure across multiple environments much more efficient. Second, these processes help make the overall system much more stable by continuously testing new changes as they are introduced to the system. Third, it allows us to be much more confident of the security controls in place since they can be tested as part of the CI process and all new changes can be audited and tracked. Finally, IaC enables the ACES Platform to be adaptable to the emerging technologies due to its ability to spin up different test beds to evaluate and incorporate these technologies. This document addresses common infrastructure-management challenges, describes what happens if they are not addressed, and highlights the value of utilizing IaC to tackle them. Finally, we will provide a high-level overview of the IaC and DevOps practices being utilized by the ACES Platform team.

97 MATHEMATICS AND COMPUTING↗