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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 469 records · Page 26
All sourcesNASA NTRSDOE OSTI
NASA NTRS2014

A Guide to Axial-Flow Turbine Off-Design Computer Program AXOD2

A Users Guide for the axial flow turbine off-design computer program AXOD2 is composed in this paper. This Users Guide is supplementary to the original Users Manual of AXOD. Three notable contributions of AXOD2 to its predecessor AXOD, both in the context of the Guide or in the functionality of the code, are described and discussed in length. These are: 1) a rational representation of the mathematical principles applied, with concise descriptions of the formulas implemented in the actual coding. Their physical implications are addressed; 2) the creation and documentation of an Addendum Listing of input namelist-parameters unique to AXOD2, that differ from or are in addition to the original input-namelists given in the Manual of AXOD. Their usages are discussed; and 3) the institution of proper stoppages of the code execution, encoding termination messaging and error messages of the execution to AXOD2. These measures are to safe-guard the integrity of the code execution, such that a failure mode encountered during a case-study would not plunge the code execution into indefinite loop, or cause a blow-out of the program execution. Details on these are discussed and illustrated in this paper. Moreover, this computer program has since been reconstructed substantially. Standard FORTRAN Langue was instituted, and the code was formatted in Double Precision (REAL*8). As the result, the code is now suited for use in a local Desktop Computer Environment, is perfectly portable to any Operating System, and can be executed by any FORTRAN compiler equivalent to a FORTRAN 9095 compiler. AXOD2 will be available through NASA Glenn Research Center (GRC) Software Repository.

Gas Turbine Engines↗
NASA NTRS2015

TankSIM: A Cryogenic Tank Performance Prediction Program

Developed for predicting the behavior of cryogenic liquids inside propellant tanks under various environmental and operating conditions. Provides a multi-node analysis of pressurization, ullage venting and thermodynamic venting systems (TVS) pressure control using axial jet or spray bar TVS. Allows user to combine several different phases for predicting the liquid behavior for the entire flight mission timeline or part of it. Is a NASA in-house code, based on FORTRAN 90-95 and Intel Visual FORTRAN compiler, but can be used on any other platform (Unix-Linux, Compaq Visual FORTRAN, etc.). The last Version 7, released on December 2014, included detailed User's Manual. Includes the use of several RefPROP subroutines for calculating fluid properties.

Bolshinskiy, L. G.↗
DOE OSTI2017

Futility

The Futility package contains the following: 1) Definition of the size of integers and real numbers; 2) A generic Unit test harness; 3) Definitions for some basic extensions to the Fortran language: arbitrary length strings, a parameter list construct, exception handlers, command line processor, timers; 4) Geometry definitions: point, line, plane, box, cylinder, polyhedron; 5) File wrapper functions: standard Fortran input/output files, Fortran binary files, HDF5 files; 6) Parallel wrapper functions: MPI, and Open MP abstraction layers, partitioning algorithms; 7) Math utilities: BLAS, Matrix and Vector definitions, Linear Solver methods and wrappers for other TPLs (PETSC, MKL, etc), preconditioner classes; 8) Misc: random number generator, water saturation properties, sorting algorithms.

Collins, Benjamin S.↗
DOE OSTI2022

From PeleC to PeleACC, to PeleC++

PeleC is an Exascale Computing Project application for simulating compressible combustion in complex geometries. It has been built on top of the popular AMReX library. In the beginning of the Exascale Computing Project, PeleC was focused on KNL. It uses a mixture of C++, C, and kernels written in Fortran to obtain performance by focusing on vectorization. Recently we have taken two approaches in deciding PeleC's future for obtaining performance on exascale GPU machines. In the first programming model, we decorated the Fortran kernels with OpenACC directives. This expedited our ability to run at large scales on Summit's GPUs, where we achieved a significant speedup over the CPUs on Summit. The second programming model involved rewriting the Fortran kernels in C++ and using AMReX's Kokkos-like lambda abstractions for running on the GPU. This resulted in similar speedups on Summit's GPUs over merely utilizing the CPUs. Both approaches involved AMReX's management of memory transfers between the device and host. In this work, we compare and contrast the benefits and pitfalls to both programming approaches regarding performance, performance portability, and productivity. We also discuss advantages we have found in taking the time to modernize our code and why have chosen a specific pathway to prepare our code for the future DOE exascale machines.

exascale computing↗
DOE OSTI2010

Covariant spectator theory of np scattering: Effective range expansions and relativistic deuteron wave functions

We present the effective range expansions for the {sup 1}S{sub 0} and {sup 3}S{sub 1} scattering phase shifts, and the relativistic deuteron wave functions that accompany our recent high precision fits (with {chi}{sup 2}/N{sub data{approx_equal}}1) to the 2007 world np data below 350 MeV. The wave functions are expanded in a series of analytical functions (with the correct asymptotic behavior at both large and small arguments) that can be Fourier-transformed from momentum to coordinate space and are convenient to use in any application. A fortran subroutine to compute these wave functions can be obtained from the authors.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗
DOE OSTI2019

EGS Collab Testbed 1: Temperature profile

This submission includes an input file, plot file, Fortran conversion file, and 3D data file from the simulation of the temperature profile within the Test Bed #1 of the EGS Collab project. The simulation was executed with PNNL's STOMP-GT simulator, which reads the input file, and produces the STOMP-GT Simulated 2D Temperature Distribution Data (plot.001302 file). The Fortran conversion file, converts the STOMP-GT Simulated 2D Temperature Distribution Data (plot.001302) 2D results in local coordinates to 3D results in Homestake coordinates.

15 GEOTHERMAL ENERGY↗
DOE OSTI2011

A user's guide to the PLTEMP/ANL code.

PLTEMP/ANL V4.1 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of-Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst's time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2011

Software enhancements to the IVSEM model of the CTBTO IMS.

Sandia National Laboratories (SNL) developed the Integrated Verification System Evaluation Model (IVSEM) to estimate the performance of the International Monitoring System (IMS) operated by the Comprehensive Nuclear Test Ban Treaty Organization (CTBTO). IVSEM was developed in several phases between 1995 and 2000. The model was developed in FORTRAN with an IDL-based user interface and was compiled for Windows and UNIX operating systems. Continuing interest in this analysis capability, coupled with numerous advances in desktop computer hardware and software since IVSEM was written, enabled significant improvements to IVSEM run-time performance and data analysis capabilities. These improvements were implemented externally without modifying the FORTRAN executables, which had been previously verified. This paper describes the parallelization approach developed to significantly reduce IVSEM run-times and the new test setup and analysis tools developed to facilitate better IVSEM operation.

98 NUCLEAR DISARMAMENT, SAFEGUARDS, AND PHYSICAL P↗
DOE OSTI2012

Extension of the Supercritical Carbon Dioxide Brayton Cycle to Low Reactor Power Operation: Investigations using the Coupled ANL Plant Dynamics Code-SAS4A/SASSYS-1 Liquid Metal Reactor Code System

Significant progress has been made on the development of a control strategy for the supercritical carbon dioxide (S-CO 2 ) Brayton cycle enabling removal of power from an autonomous load following Sodium-Cooled Fast Reactor (SFR) down to decay heat levels such that the S-CO 2 cycle can be used to cool the reactor until decay heat can be removed by the normal shutdown heat removal system or a passive decay heat removal system such as Direct Reactor Auxiliary Cooling System (DRACS) loops with DRACS in-vessel heat exchangers. This capability of the new control strategy eliminates the need for use of a separate shutdown heat removal system which might also use supercritical CO 2 . It has been found that this capability can be achieved by introducing a new control mechanism involving shaft speed control for the common shaft joining the turbine and two compressors following reduction of the load demand from the electrical grid to zero. Following disconnection of the generator from the electrical grid, heat is removed from the intermediate sodium circuit through the sodium-to-CO 2 heat exchanger, the turbine solely drives the two compressors, and heat is rejected from the cycle through the CO 2 -to-water cooler. To investigate the effectiveness of shaft speed control, calculations are carried out using the coupled Plant Dynamics Code-SAS4A/SASSYS-1 code for a linear load reduction transient for a 1000 MWt metallic-fueled SFR with autonomous load following. No deliberate motion of control rods or adjustment of sodium pump speeds is assumed to take place. It is assumed that the S-CO 2 turbomachinery shaft speed linearly decreases from 100 to 20% nominal following reduction of grid load to zero. The reactor power is calculated to autonomously decrease down to 3% nominal providing a lengthy window in time for the switchover to the normal shutdown heat removal system or for a passive decay heat removal system to become effective. However, the calculations reveal that the compressor conditions are calculated to approach surge such that the need for a surge control system for each compressor is identified. Thus, it is demonstrated that the S-CO 2 cycle can operate in the initial decay heat removal mode even with autonomous reactor control. Because external power is not needed to drive the compressors, the results show that the S-CO 2 cycle can be used for initial decay heat removal for a lengthy interval in time in the absence of any off-site electrical power. The turbine provides sufficient power to drive the compressors. Combined with autonomous reactor control, this represents a significant safety advantage of the S-CO 2 cycle by maintaining removal of the reactor power until the core decay heat falls to levels well below those for which the passive decay heat removal system is designed. The new control strategy is an alternative to a split-shaft layout involving separate power and compressor turbines which had previously been identified as a promising approach enabling heat removal from a SFR at low power levels. The current results indicate that the split-shaft configuration does not provide any significant benefits for the S-CO 2 cycle over the current single-shaft layout with shaft speed control. It has been demonstrated that when connected to the grid the single-shaft cycle can effectively follow the load over the entire range. No compressor speed variation is needed while power is delivered to the grid. When the system is disconnected from the grid, the shaft speed can be changed as effectively as it would be with the split-shaft arrangement. In the split-shaft configuration, zero generator power means disconnection of the power turbine, such that the resulting system will be almost identical to the single-shaft arrangement. Without this advantage of the split-shaft configuration, the economic benefits of the single-shaft arrangement, provided by just one turbine and lower losses at the design point, are more important to the overall cycle performance. Therefore, the single-shaft configuration shall be retained as the reference arrangement for S-CO 2 cycle power converter preconceptual designs. Improvements to the ANL Plant Dynamics Code have been carried out. The major code improvement is the introduction of a restart capability which simplifies investigation of control strategies for very long transients. Another code modification is transfer of the entire code to a new Intel Fortran complier; the execution of the code using the new compiler was verified by demonstrating that the same results are obtained as when the previous Compaq Visual Fortran compiler was used.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2015

A User's Guide to the PLTEMP/ANL Code

PLTEMP/ANL V4.2 is a FORTRAN program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of ''PLTEMP'' codes in use at ANL for the past 20 years. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each with its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates/tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as Onset-of- Nucleate boiling (ONB), departure from nucleate boiling (DNB), and onset of flow instability (FI). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1966

An auto-diagrammer

Program to draw flowcharts automatically from PERT, Fortran II, Fortran IV, or FAP cards

PERT PROJECT↗
NASA NTRS2014

pFUnit 3.0 Tutorial Advanced

This tutorial will introduce Fortran developers to unit-testing and test-driven development (TDD) using pFUnit. As with other unit-testing frameworks, pFUnit, simplifies the process of writing, collecting, and executing tests while providing clear diagnostic messages for failing tests. pFUnit specifically targets the development of scientific-technical software written in Fortran and includes customized features such as: assertions for multi-dimensional arrays, distributed (MPI) and thread-based (OpenMP) parallellism, and flexible parameterized tests.These sessions will include numerous examples and hands-on exercises that gradually build in complexity. Attendees are expected to have working knowledge of F90, but familiarity with object-oriented syntax in F2003 and MPI will be of benefit for the more advanced examples. By the end of the tutorial the audience should feel comfortable in applying pFUnit within their own development environment.

Test Driven using pFUnit↗
DOE OSTI2010

An overview of the Morfeus project.

The objectives of this project are to: (1) move scientific programmers to higher-level, platform-agnostic yet scalable abstractions; (2) to demonstrate general OOD patterns and distill new domain-specific patterns from multiphysics applications in Fortran; and (3) to construct an open-source framework that encourages the use of the demonstrated patterns. Some conclusions are: (1) Calculus illuminates a path toward highly asynchronous computing that blurs the task/data parallel distinction; (2) Fortran 2003 appears to have the expressiveness to support the general GoF design patterns in multiphysics applications; and (3) several domain-specific and language-specific patterns emerge along the way.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2010

For Trilinos tutorial.

The objectives are: (1) To increase the adoption of Trilinos throughout DOE research communities that principally write Fortran, e.g. climate & combustion researchers; and (2) To maintain the OOP philosophy of the Trilinos project while using idioms that feel natural to Fortran programmers.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2016

Updating a PWR Simulator in Python

A nuclear reactor is a complicated system that is comprised of many systems and components. In order to fully model a reactor, several approximations and correlations must be used for a real-time simulation. In 1984 as part of a doctoral dissertation at Massachusetts Institute of Technology (MIT), a FORTRAN 77 program was developed to simulate a four-loop, Westinghouse pressurized water reactor (PWR). The program was later called PRISM and extended to include a comprehensive set of features accessible in a DOS-based user interface. This present work seeks to update the original PRISM program for cross platform deployment using the Python language. The goal of this present work is to update and improve a teaching tool for both students and reactor operators. Originally, PRISM was created as part of a PhD project in FORTRAN 77, but the code no longer works readily on modern operating systems. Python was chosen to rewrite the program because of the ease of use of the Python modules including Kivy as well as the portability of the program to various operating systems. This project aims to develop desktop and mobile versions of the software, allowing the user flexibility in choice of platform. This simulator is scheduled to be deployed in time for a live demonstration at the 2016 Annual ANS meeting. The project aims to produce a reactor simulator that will showcase the dynamics of a PWR system in real-time under various conditions. Some potential scenarios include startup conditions, scram operations, or failure conditions such as a break in a coolant leg. The user will be able to induce various failure scenarios and the program will simulate how the reactor will behave. The simulation will include a suite of monitoring functions that will allow automatic systems to attempt to adapt to the changing conditions, e.g., control rod movement. This program will provide an educational resource for anyone learning about nuclear reactor kinetics or dynamics of a reactor system. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2019

TableCalculator: a Transparent Public Tool to Replicate US NRC LLW Classification Table Calculations - 19395

Classification of Low-Level Waste (LLW) as Class A, B, C, or Greater than Class C (GTCC) is based on tables provided in Section 61.55 of Title 10 of the Code of Federal Regulations (CFR). Those tables were developed from model projections of dose to an inadvertent intruder in agricultural and construction scenarios at different times after closure of a LLW land disposal site. The calculations were performed with two FORTRAN codes, which are documented in US NRC guidance documents published in the early 1980's. Although the equations and parameter values used in the codes are publicly available, no modern, user-friendly implementation of the codes is known to US NRC staff. Because Very Low-Level Waste (VLLW) and GTCC waste disposal are related to the low end and high end of the existing LLW waste classifications, respectively, recent interest in options for disposal of VLLW and GTCC waste have generated public interest in the original assumptions made during the development of the LLW classification tables. Certain stakeholder analyses have recommended alternatives to the existing LLW concentration limits by calculating a projected intruder dose under different conditions than were evaluated in the original development of 10 CFR Part 61 (e.g., using more modern dosimetry or assuming deeper disposal) and comparing the radionuclide concentrations that result in a projected annual dose of 5 milliSieverts (mSv) (500 millirem (mrem)) total effective dose equivalent (TEDE) to the concentrations in the waste classification tables. However, beginning with the 10 CFR Part 61 LLW classification limits and modifying only one or two assumptions to generate a proposed concentration limit could lead to misleading results without a thorough analysis of whether other assumptions made in the original development of the waste classification tables remain valid. In particular, the US NRC made adjustments to the 10 CFR 61.55 table values during the rulemaking process based on qualitative considerations. Those adjustments mean that many of the values in the final tables cannot simply be assumed to quantitatively represent a 5 mSv (500 mrem) projected inadvertent intruder dose without an assessment of whether the technical bases used to develop the adjustments are applicable to the new analysis. Furthermore, the waste classification limits were based on the limiting critical organ dose rather than a TEDE. Therefore, any calculation based on the assumption that the waste classification limits correspond to 5 mSv (500 mrem) TEDE could benefit from consideration of how differences in exposure pathways could affect the limiting critical organ dose differently from the TEDE. To facilitate a more comprehensive understanding of the calculations used to develop the waste classification tables, the US NRC staff has developed a user-friendly tool to replicate the original calculations. Results were verified against output from the FORTRAN codes and the classification table values. The tool, TableCalculator, is expected to be made publicly available and will allow the user to trace the original calculations and to observe the effects of changes in disposal assumptions and other parameter values. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2017

Analysis of a nuclear hybrid energy system using absorption chillers and stratified chilled-water storage with an mPower reactor - 187

The intermittency of renewable energy technology, such as wind and solar power, puts strain on electric grids, often forcing carbon-based and nuclear sources of energy to operate in a load follow mode. Operating nuclear reactors in a load-follow fashion is undesirable due to the associated thermal and mechanical stresses placed on the fuel and reactor components. Various methods of Thermal Energy Storage (TES) can be coupled to nuclear (or renewable) power sources to help absorb grid instabilities caused by daily load demand changes and renewable intermittency. This is known as a Nuclear Hybrid Energy System (NHES). Previous research has found that a stratified chilled-water storage tank can help displace peak cooling loads to off-peak hours. In this study, a dynamic single effect, lithium bromide absorption chiller model was developed as a callable FORTRAN subroutine. Cooling towers, a stratified chilled-water storage tank, and cooling loads characteristic of a large campus were included to form a comprehensive TES model. This TES model was coupled to a high-fidelity mPower sized Small Modular Reactor (SMR) FORTRAN code that provides the transient behavior of primary and secondary-side components. Control algorithms based on excess capacity and conditions in the stratified chilled-water storage tank were used to determine when to divert low-pressure steam from the turbines to the absorption chillers in order to keep reactor power near steady-state. Simulation results depict how the TES system and SMR interact, and demonstrate that stratified chilled-water storage is a viable TES technology for an SMR in a NHES. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2014

Eikonal Corrections for Nucleus-Nucleus Reaction Cross Sections - Paper 90

The goal of this research is to modify a Fortran computer code, which uses an optical potential approximation to quantum mechanical multiple-scattering theory to compute total and reaction cross sections for nucleon-nucleus and nucleus-nucleus collisions for use in space radiation protection and shielding applications. These nuclear collision cross sections are utilized in space radiation transport codes that describe the passage of solar and galactic cosmic radiation fields through matter. In the code being modified by this work, an Eikonal approximation, in analogy with the complex index of refraction used to describe absorption in optics, is used to formulate the nuclear scattering amplitudes that describe the total and reaction cross sections in the formalism. Previous work found that the cross sections calculated with this model sometimes underestimated the experimental data. It has been noted that the Eikonal phase functions, which are used to calculate these cross sections and which contain all of the scattering information in the model, can be improved by incorporating higher order corrections In this work, a previously-developed first-order correction to the Eikonal phase function is coded in Fortran and incorporated into the optical model computer code that calculates reaction cross sections for nucleon-nucleus and nucleus-nucleus collisions at energies relevant to space radiation protection and shielding. The main thrust of this work is to test the use of this higher order correction by calculating reaction cross sections for a variety of nuclear collisions and comparing the results to experimental data. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗