Results for “analysis codes” — Page 14

DOE OSTI2019

Sierra/SolidMechanics 4.54 Goodyear Specific

Sierra/SolidMechanics (Sierra/SM) is a Lagrangian, three-dimensional finite element analysis code for solids and structures subjected to extensive contact and large deformations, encompassing explicit and implicit dynamic as well as quasistatic loading regimes. This document supplements the primary Sierra/SM 4.54 User's Guide, describing capabilities specific to Goodyear analysis use cases, including additional implicit solver options, material models, finite element formulations, and contact settings.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2019

Sierra/SolidMechanics 4.52 Goodyear Specific

Sierra/SolidMechanics (Sierra/SM) is a Lagrangian, three-dimensional finite element analysis code for solids and structures subjected to extensive contact and large deformations, encompassing explicit and implicit dynamic as well as quasistatic loading regimes. This document supplements the primary Sierra/SM 4.52 User's Guide, describing capabilities specific to Goodyear analysis use cases, including additional implicit solver options, material models, finite element formulations, and contact settings.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2021

Sierra/SolidMechanics 5.0 Goodyear Specific

Sierra/SolidMechanics (Sierra/SM) is a Lagrangian, three-dimensional finite element analysis code for solids and structures subjected to extensive contact and large deformations, encompassing explicit and implicit dynamic as well as quasistatic loading regimes. This document supplements the primary Sierra/SM 5.0 User’s Guide, describing capabilities specific to Goodyear analysis use cases, including additional implicit solver options, material models, finite element formulations, and contact settings.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2020

Sierra/SolidMechanics 4.56 Goodyear User's Guide

Sierra/SolidMechanics (Sierra/SM) is a Lagrangian, three-dimensional finite element analysis code for solids and structures subjected to extensive contact and large deformations, encompassing explicit and implicit dynamic as well as quasistatic loading regimes. This document supplements the primary Sierra/SM 4.56 User’s Guide, describing capabilities specific to Goodyear analysis use cases, including additional implicit solver options, material models, finite element formulations, and contact settings.

42 ENGINEERING↗

DOE OSTI2025

Theory and User Manual for the Thermal Hydraulic Code SE2-ANL

A theory and user manual of thermal hydraulic code SE2-ANL was prepared in this work, to resolve the missing documentation. SE2-ANL, a special version of SUPERENERGY-2, is a multi-assembly steady-state Liquid Metal Reactor (LMR) core thermal hydraulic analysis code, and it was developed in the 1980s, but no complete documentation was ever made. This report describes the detailed methodology of the temperature calculation, inherited from (original) SUPERENERGY-2, and fuel and cladding temperature calculation capabilities as well as many other features which were incorporated in SE2-ANL. The input and output of SE2-ANL are also demonstrated in detail, using some simple models. The orificing zone capability as well as some other useful features of SE2-ANL are discussed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2018

Thermal Stratification Modeling and Analysis for Sodium Fast Reactor Technology

General discussion of thermal hydraulic phenomenon present in sodium fast reactors during known reactor transients. Specifically, this research focuses on presenting research focused on modeling thermal stratification in the upper plenum of a sodium fast reactor. An experimental facility was built to provide validation for both computational fluid dynamic and system code analysis.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2023

BISON analysis of FeCrAl and Zircaloy cladding deformation during simulated BWR cyclic dryout conditions

The BISON fuel performance analysis code has been used to study FeCrAl and Zircaloy cladding performance to better understand cladding deformation behavior in boiling water reactors during conditions caused by anticipated operational occurrences (AOOs) or by anticipated transients without scram (ATWS). BISON models of novel separate effects tests simulating ATWS events predicted mechanical responses to thermal cycling boundary conditions. Here, we assessed FeCrAl and Zircaloy deformation and burst models to determine those best suited for cyclic dryout event analysis. In addition, we compared diametral strain profiles of the post-experiment claddings to simulation results. The capability of the BISON code for modeling BWR AOO or ATWS events is promising.

36 MATERIALS SCIENCE↗

DOE OSTI2019

Validation of RELAP5-3D He-FUS3 Data

The RELAP5-3D computer code was validated using data from the HE FUS3 experimental facility. HE FUS3 is a helium-cooled, electrically-heated experimental facility that was used to generate experimental data for a benchmark exercise aimed at the validation of system transient analysis codes for very-high-temperature gas reactor applications. RELAP5-3D was validated using results from seven steady-state tests and one transient that simulated a loss-of-flow accident.

97 MATHEMATICS AND COMPUTING↗

DOE OSTI2016

Full Integration of the Fission Product Release Module STACY into the HTR Code Package (HCP) - 18661

The HTR Code Package is under development since 2010 at Forschungszentrum Juelich and at Aachen University. The HCP combines the capabilities of individual legacy codes developed during the German HTR program in a highly integrated manner and contains new features. Within the HCP, all aspects related to fission product release and the redistribution within the primary circuit are handled by the module STACY (Source Term Analysis Code system). STACY is the successor of the codes FRESCO-I/II (dealing with fission product release) and PANAMA (dealing with fuel performance issues). Compared to these predecessor codes, STACY can take amongst others a time-dependent nuclide inventory and a radial temperature profile within the fuel element into account. So far, this version of STACY performs an individual fission product release calculation for a representative number of tracer pebbles. While this version of STACY still relies on codes like VSOP and MGT-3D and contains features such as basic fuel management which are also contained in the before mentioned codes, STACY now will be coupled to forms an integral part of the HCP. Thus, fission product release calculations will be examined with one consistent reactor input model. This allows assessing the release of fission products already during the design phase, besides parameters like the maximum fuel temperature during a transient. Furthermore the total nuclide inventory of a fuel element is distributed among fuel kernels, coating layers, graphite grains and pores of the surrounding matrix graphite. Formerly, this total inventory calculated by the burnup code TNT was distributed according to the distribution of Uranium. As a new feature, STACY now keeps track of the distributions of nuclide inventories within the fuel element. Especially for nuclides mainly formed by activation and decay processes like Ag-110m, this approach is more accurate. Further, the knowledge about the actual nuclide inventories improves the accuracy of fuel performance calculations. In addition, the internal coupling allows a feedback from the fission product calculation to other code modules. Nuclides amounts being released can be considered in e.g. the burnup calculation while calculating the source term of for example tritium or within the calculation which takes care of the redistribution of species by the coolant. Within this paper, the application of the STACY to HTR cores with special emphasis on fission product release to reactor models such as the HTR-Modul and HTR-10 will be discussed. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2016

Comparison of Surrogate Models with Physical Models for Dynamic Probabilistic Risk Analysis Using the RAVEN Code

Dynamic probabilistic risk analysis (DPRA) has emerged in the nuclear industry as an effective tool to identify points of weakness in nuclear reactors and create a comprehensive approach to quantifying risk. While DPRA allows for detailed analysis, it usually requires thousands of runs due to the necessity of exploring low probability, high consequence events, which can be computationally prohibitively difficult to assess. Risk Analysis Virtual Environment (RAVEN) is a software package being developed under the Nuclear Energy Advanced Modeling and Simulation program at the Idaho National Laboratory. It is able to manage complex control logic to drive a RELAP-7 simulation and to handle sampling strategies including Monte Carlo, Latin Hypercube, Grid and Adaptive samplers. By performing a thorough investigation of the possibility space of an accident scenario with RELAP-7, RAVEN is able to identify consequences of accident scenarios that have not been encountered. The most computational resource intensive step of a RAVEN analysis is running the RELAP-7 code modeling the physical system behavior. Using surrogate models to represent the outcomes of more complex simulations rather than the full model can greatly reduce the computation effort required. In situations where the uncertainty distributions of a model are unknown, or when there are multiple sets of uncertainty distributions, DPRA generally requires rerunning the physical model for each set of uncertainty distributions used. By taking an unbiased sample of the physical model and creating a surrogate model from the results, the surrogate model can be sampled under a different set of uncertainty distributions to provide equivalent results. This paper presents an example of this approach. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2018

Analysis for Building Envelopes and Mechanical Systems Using 2012 CBECS Data

This report describes the aggregation and mapping of certain building characteristics data available in the most recent Commercial Building Energy Consumption Survey (CBECS) (DOE EIA 2012) to describe most typical building construction practices. This report provides summary data for potential use in the support of modifications to the Pacific Northwest National Laboratory’s commercial building prototypes used for building energy code analysis. Specifically, this report outlines findings and most typical design choices for certain building envelope and heating, ventilating, and air-conditioning (HVAC) system choices.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗

DOE OSTI2018

Validation of RELAP5-3D Using HE-FUS3 Data

The RELAP5-3D computer code was validated using data from the HE-FUS3 experimental facility. HE-FUS3 is a helium-cooled, electrically-heated experimental facility that was used to generate experimental data for a benchmark exercise aimed at the validation of system transient analysis codes for very-high-temperature gas reactor applications. RELAP5-3D was validated using results from seven steady-state tests and one transient that simulated a loss-of-flow accident.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗

DOE OSTI2020

ADETS User Manual

ADETS (Automated DEpletion Transport System) calculates coupled neutronic/isotopic results for nuclear systems and produces a large number of criticality and burnup results based on various material feed/removal specifications, power(s), and time intervals. ADETS is a fully automated tool that links the LANL MCNP Monte Carlo transport code with the SCALE system, specifically ORIGEN-ARP (radioactive decay and burnup code). Additionally, ADETS can compute the dose at various distances of the reactor using a point gamma source, whose characteristics are computed by ORIGEN-ARP. In addition, ADETS is fully coupled with the Uncertanty Quantification, Probabilistic Risk Assessment, Parameter Optimization and Data Analysis code RAVEN (developed at the Idaho National Laboratory as well). RAVEN is released with ADETS but MCNP and SCALE (ORIGEN-ARP) are not included.

42 ENGINEERING↗

DOE OSTI2025

Preliminary Design of Engineering-Scale Salt Accident Analysis Facility to Support Molten Salt Reactor Licensing

Systems-level nuclear accident analysis codes for reactor licensing must be validated using experimental data that represent behaviors expected during actual full-scale accidents. Some behaviors may arise from coupled processes and only manifest at large scales. This report presents the preliminary design of the Salt Accident Analysis Facility (SAAF, pronounced “safe”), which is an experimental test facility to be constructed at Argonne that can be used to conduct integrated salt accident tests at an engineering scale. The measurement capabilities of the SAAF are based on previously developed methods and will provide the representative datasets that are needed to support molten salt reactor (MSR) licensing. Details of the design, the processes to be quantified, the measurement techniques for quantifying the processes, the variables that can be adjusted to simulate different accident scenarios, and operational considerations are presented herein. This report provides stakeholders the opportunity to give feedback on the test facility capabilities and planned analyses before it is constructed.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

DOE OSTI2020

The FEFF code

FEFF is a real-space multiple-scattering Green’s function code for simulating a variety of X-ray spectroscopies as well as ground-state and excited-state electronic structure. It is perhaps best known for its use in extended X-ray absorption fine-structure (EXAFS) calculations, particularly within various analysis codes such as DEMETER and LARCH. In addition to EXAFS, FEFF can also be used to calculate X-ray absorption near-edge structure (XANES), X-ray emission spectroscopy (XES), electron energy loss spectra (EELS) and many other spectroscopic quantities. Here, the basic approach used for the calculations is described, focusing on the most important approximations and aspects of the simulations. Here, the capabilities of the FEFF code, including those of the Java-based graphical user interface JFEFF, are also briefly described.

75 CONDENSED MATTER PHYSICS, SUPERCONDUCTIVITY AND↗

DOE OSTI2026

Calibration and Data Analysis of a Frequency Selectable Laser Source for CMB Detector Characterization

Cosmic Microwave Background (CMB) experiments study faint radiation left over from the early universe. The CMB was created when the universe became cool enough for light to travel freely through space, and today it gives scientists one of the earliest images of the universe. One important goal of modern CMB experiments is to measure this radiation with higher precision in order to search for evidence that supports the theory of cosmic inflation. To do this, scientists use extremely sensitive detectors that must be calibrated accurately. The Frequency Selectable Laser Source, or FLS, is a new calibration tool that can send selected frequencies to detectors and help measure their response. During my internship, I worked on the FLS after it returned to Fermilab from Chile, where it had been used to characterize detectors at the Simons Observatory. The system came back in parts, so the first part of my project was helping rebuild the optical and mechanical setup. After the system was rebuilt, we performed alignments to maximize the receiver photocurrent. We then collected calibration measurements over different frequency ranges, including 543 GHz to 568 GHz, 740 GHz to 766 GHz, and 60 GHz to 500 GHz. These measurements were used to check waterline calibration and reflectivity features and compare new data with previous data. Another major part of my project was learning Python so I could understand previous analysis code, modify it for new files, and write my own code to compare the mean response between datasets. The results showed that the new data was close to previous measurements and that waterline features near 556 GHz and 752 GHz were found within less than 1.5 GHz of the expected values. I also completed the reflectivity analysis for five prisms in two polarization orientations. In the original orientation, the results were consistent between the five prisms and close to values measured on a different system at the University of Chicago. I then collected a second set of measurements on my own with the polarization of the laser rotated by 90 degrees and compared them with the original data using the same Python workflow. The measured reflectivity increased for all five prisms in the new orientation, showing that the prism reflectivity depends on polarization. Future work will focus on using the FLS to characterize real CMB detectors.

Pumarino, Rafael [Unlisted]↗

DOE OSTI2017

Developmental study of advanced HIS design method for digital I and C+HMIT of PWR plant - 277

A new conceptual frame of how to design and validate a digital HIS (human interface system) on an innovative numerical simulation basis is proposed for the support of plant operators' supervisory control of various types of automated complex NPPs (nuclear power plants). The proposed conceptual framework utilizes the object-oriented software for plant DiD (defense-in depth) risk monitor with the combination of nuclear reactor accident simulation by an advanced nuclear safety analysis code RELAP5/MOD4. The conceptual frame proposed in this paper will be applied for an example practice for the SBLOCA (small break loss of coolant accident) case of passive safety PWR (pressurized water reactor) AP1000. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

DOE OSTI2026

Calibration and Data Analysis of a Frequency Selectable Laser Source for CMB Detector Characterization

Cosmic Microwave Background (CMB) experiments study faint radiation left over from the early universe. The CMB was created when the universe became cool enough for light to travel freely through space, and today it gives scientists one of the earliest images of the universe. One important goal of modern CMB experiments is to measure this radiation with higher precision in order to search for evidence that supports the theory of cosmic inflation. To do this, scientists use extremely sensitive detectors that must be calibrated accurately. The Frequency Selectable Laser Source, or FLS, is a new calibration tool that can send selected frequencies to detectors and help measure their response. During my internship, I worked on the FLS after it returned to Fermilab from Chile, where it had been used to characterize detectors at the Simons Observatory. The system came back in parts, so the first part of my project was helping rebuild the optical and mechanical setup. After the system was rebuilt, we performed alignments to maximize the receiver photocurrent. We then collected calibration measurements over different frequency ranges, including 543 GHz to 568 GHz, 740 GHz to 766 GHz, and 60 GHz to 500 GHz. These measurements were used to check waterline calibration and reflectivity features and compare new data with previous data. Another major part of my project was learning Python so I could understand previous analysis code, modify it for new files, and write my own code to compare the mean response between datasets. The results showed that the new data was close to previous measurements and that waterline features near 556 GHz and 752 GHz were found within less than 1.5 GHz of the expected values. I also completed the reflectivity analysis for five prisms in two polarization orientations. In the original orientation, the results were consistent between the five prisms and close to values measured on a different system at the University of Chicago. I then collected a second set of measurements on my own with the polarization of the laser rotated by 90 degrees and compared them with the original data using the same Python workflow. The measured reflectivity increased for all five prisms in the new orientation, showing that the prism reflectivity depends on polarization. Future work will focus on using the FLS to characterize real CMB detectors.

Pumarino Meza, Rafael [Unlisted, US; Fermilab]↗