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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 793 records · Page 44
All sourcesNASA NTRSDOE OSTI
DOE OSTI2018

Software Requirements for a Point Kinetics Neutronics Code for Simulating TREAT Reactor Transients

This document provides the software requirements for a computer code to perform point kinetics-based simulations of nuclear reactor transients. These requirements address all aspects of code development and use, including input and output parameters and structure, software capabilities, and supporting documentation. This software is intended to replace the existing point kinetics code TREKIN, which is currently used for the analysis of the Transient Reactor Test Facility (TREAT).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Development of ASME Division 5 Code proposal on temperature limits for simplified design methods

This report establishes proposed upper temperature limits for the ASME BP&V Code Section III, Division 5, Nonmandatory Appendix HBB-T design by elastic analysis provisions for bounding ratcheting strain and creep-fatigue damage in Class A high temperature nuclear reactor components. Limitations on the use of these design options are required because the design by elastic analysis methods rely on bounding theories that assume a non-unified, decoupled model of creep-plasticity. However, at high temperatures creep and plastic deformation become coupled and bounding theorems relying a decoupled material response may fail. The report describes a method for selecting appropriate upper temperature limits, demonstrates directly through a comparison to full inelastic simulations that the existing Code provisions can be nonconservative at high temperatures, and develops the requisite Code language required to implement the temperature limits in Section III, Division 5.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Construction of an Object-Oriented Interface to an Enhanced EGS Monte Carlo Code

The version of SLAC’s EGS Code System (known as EGS4) was clearly recognized throughout the world as a “gold standard” for solving physics and engineering problems related to the transport of electrons and photons in any element, compound, or mixture. However, EGS4 was written in a computer language called Mortran3 and the learning curve for this rather obscure language was difficult and steep. One major objective of this project was to convert the EGS4 Mortran3 coding into Fortran77, making it much simpler for first-time users to understand and appreciate. In addition, since the introduction of EGS4 in 1985, a number of enhancements in electron-photon transport, as well as a growing need for additional physics processes, had been developed, and the new code (called EGS5) would incorporate more than 20 of these improvements. Of course, many key benchmark simulations had to be made during its development in order to prove that EGS5 was just as credible as its predecessor, EGS4. All of the above was tasked to SLAC, with outside support coming from colleagues at the University of Michigan (Nuclear Engineering Department) and at the High Energy Accelerator Research Organization (KEK) in Japan.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2018

MC 2 -3: Multigroup Cross Section Generation Code for Fast Reactor Analysis Nuclear

The MC 2 -3 code is a Multigroup Cross section generation Code for fast reactor analysis, developed by improving the resonance self-shielding and spectrum calculation methods of MC 2 -2 and integrating the one-dimensional cell calculation capabilities of SDX. The code solves the consistent P1 multigroup transport equation using basic neutron data from ENDF/B data files to determine the fundamental mode spectra for use in generating multigroup neutron cross sections.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

H2OI95: A Stand-Alone Fortran Code for Evaluating the IAPWS-95 Equation-of-State Model for Water

H2OI95 is a stand-alone Fortran code for evaluating the IAPWS-95 equation-of-state model (Wagner and Pruss, 2002) for the thermodynamic properties of water. It further evaluates the corresponding thermochemical properties of water consistent with the CODATA recommendations (Cox et al., 1989). The IAPWS-95 model is based on a model equation for the dimensionless Helmholtz energy for which the primary variables are the inverse reduced temperature (τ = T cr /T) and reduced density ( δ= ρ/ρ cr ). Here T is the absolute temperature (K), ρ is density (kg/m 3 ), and the subscript “cr” refers to the critical point of water (647.096 K and 22.064 MPa pressure in this model, for which ρ cr is 322 kg/m 3 ). The code solves four basic types of problems, distinguished by the specified inputs: 1, Temperature (K) and density (ρ kg/m 3 ) or reduced density (δ) 2. Temperature (K) and pressure (MPa). 3. Temperature (K) on the saturation (liquid-vapor equilibrium) curve 4. Pressure (MPa) on the saturation curve Each type of problem is run using a corresponding input (text) file. All but the first type of problem require iteration. For example, to solve for desired temperature and pressure, the reduced density must be adjusted to give the desired pressure. Iteration is accomplished using the Newton-Raphson method, though the secant method is also used in solving the fourth type of problem. H2OI95 has been used to conduct numerical studies of convergence and the problem of multiple numerical solutions, only some of which correspond to valid results. Obtaining valid results depends mainly on appropriate choice of starting values. The default values in H2OI95 appear to consistently lead to generally desired results. With modification (not addressed here), H2OI95 can be used to support SUPCRT92 (Johnson et al., 1992) and similar codes that compute chemical thermodynamic properties of species and reactions over a wide range of temperature and pressure (273.16-1273K and 0-1000 MPa).

37 INORGANIC, ORGANIC, PHYSICAL, AND ANALYTICAL CH↗
DOE OSTI2019

Relative Credits for Extra Efficiency Code Measures

Energy codes have traditionally contained mandatory and prescriptive items. Mandatory measures must be complied with in all situations while prescriptive measures can be traded with other efficiency measures by following a performance path. Recent editions of the International Energy Conservation Code (IECC) also contain additional efficiency packages, where one or more optional measures must be selected to satisfy an ‘additional’ efficiency requirement. Such a structure is currently employed in section C406 of the IECC, and similar approaches have been considered for ASHRAE Standard 90.1 as well as state building codes. This approach has the advantage of providing increasing levels of performance, while maintaining flexibility in allowing the designer to optimize the most appropriate technologies and efficiency measured based on the attributes of a particular project.

32 ENERGY CONSERVATION, CONSUMPTION, AND UTILIZATI↗
DOE OSTI2019

Codes and Standards Update: February 2019

The goal of the DOE OE Energy Storage System Safety Roadmap is to foster confidence in the safety and reliability of energy storage systems. There are three interrelated objectives to support the realization of that goal: research, codes and standards(C/S)and communication/coordination. The objective focused on C/S is "To apply research and development to support efforts that refocused on ensuring that codes and standards are available to enable the safe implementation of energy storage systems in a comprehensive, non-discriminatory and science-based manner." The following activities support that objective and realization of the goal: (a.) Review and assess C/S which affect the design, installation, and operation of energy storage systems(ESS). (b.) Identify gaps in knowledge that require research and analysis that can serve as a basis for criteria in those C/S. (c.) Identify areas in C/S that are potentially in need of revision or enhancement and can benefit from activities conducted under research and development. (d.) Develop input for new or revisions to existing C/S through individual stakeholders, facilitated task forces, or through laboratory staff supporting these efforts. The purpose of this Codes and Standards Update is to support the above activities by providing information on efforts being conducted by U.S. standards developing organizations(SDOs)and other entities that are focused on ESS safety. The information is organized relative to the scope of each document in relation to ESS from the "macro to the micro"(e.g., from overarching covering considerable scope ,to installation specific, to ESS and then ESS components). Note that more macro documents are also likely to adopt by reference more micro documents.

25 ENERGY STORAGE↗
DOE OSTI2018

RAZORBACK - A Research Reactor Transient Analysis Code (Version 1.0 - Volume 2: Software Design and Development Report)

RAZORBACK is a research reactor transient analysis computer code designed to simulate the operation of a research reactor (such as Sandia National Laboratories' Annular Core Research Reactor (ACRR)). The code provides a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, the equation of motion for fuel element thermal expansion, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This Software Design and Development Report describes how the code is designed to implement the functional requirements established in the Software Requirements Specification document for RAZORBACK. The description presents the physical model equations used in the design, and how these equations are set up to be solved numerically. This report also provides an evaluation of how the software design meets each of the functional requirements.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Energy Storage System Guide for Compliance with Safety Codes and Standards

Codes, standards and regulations (CSR) governing the design, construction, installation, commissioning and operation of the built environment are intended to protect the public health, safety and welfare. While these documents change over time to address new technology and new safety challenges there is generally some lag time between the introduction of a technology into the market and the time it is specifically covered in model codes and standards developed in the voluntary sector. After their development, there is also a timeframe of at least a year or two until they are adopted. Until existing model codes and standards are updated or new ones developed and then adopted, one seeking to deploy energy storage technologies or needing to verify an installation’s safety may be challenged in applying current CSRs to an energy storage system (ESS).

25 ENERGY STORAGE↗
DOE OSTI2019

Development and Validation of a Conjugate Heat Transfer Model for the Two-Phase CFD Code NEK-2P

A project is underway to develop, verify and validate an advanced two-phase flow modeling capability for the highly-scalable, high-performance CFD code NEK5000. The goal of the project is to develop a new two-phase version of the NEK5000 code, named NEK-2P, to simulate the two-phase flow and heat transfer phenomena that occur in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. The NEK-2P two-phase flow models follow the approach used for the Extended Boiling Framework (EBF) [2-3] previously developed at Argonne, but include more fundamental physical models of boiling phenomena and advanced numerical algorithms for improved computational accuracy, robustness, and computational speed. The development of the NEK-2P two-phase solver and the implementation of the Extended Boiling Framework two-phase models were initially supported by Argonne National Laboratory (Argonne) through a Laboratory Directed Research and Development (LDRD) project during FY2014-2016. The development and validation of the two-phase models through analyses of selected two-phase boiling flow experiments was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY2017-2019. This report focuses on the FY2019 development and validation of the Conjugate Heat Transfer (CHT) model for the NEK-2P Two-Phase, Computational Fluid Dynamics (CFD) code. The NEK-2P CHT model calculates the coupled behavior of the fluid and solid domains. In the fluid domain NEK-2P calculates the vapor and liquid phase temperatures, velocities and mass fractions. In the solid domain only the temperatures are calculated. The fluid and solid domains are coupled with a heat flux boundary condition at the fluid-solid interface. The heat flux and the wall temperatures at the fluid-solid interface are updated using the NEK-2P wall heat-flux partitioning model. A wall heat flux boundary condition is also applied to the exterior surface of the solid domain. The CHT model validation was illustrated through the analysis of three of the Becker Critical Heat Flux (CHF) tests. Reasonably good agreement with measured data was obtained in predicting the CHF location and post CHF wall temperature behavior illustrating the ability of the NEK-2P CHT model to simulate the CHF phenomena for a wide range of thermal-hydraulic conditions. Apart from the CHT model development and validation, initial simulations were performed for the Virginia Tech. (VT) 3x3 bare rod bundle experiments including both cold and boiling flow simulations.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Scan of an Unpublished Report: "Fission Product Behavior During Severe LWR Accidents: Recommendations for the MELCOR Code System"

This document provides a scanned version of a 1987 SAND report that was never formally published. However, this report was referenced within the MELCOR Reference Manual and, therefore, provides historical information and technical basis for the MELCOR code. This document is being made available to permit users of the MELCOR code access to the information. The title page has been edited to prevent any confusion with regards to the possible documentation identifiers, such as the SAND report number or the intended date of publication. Beyond these modifications, a cover, distribution, and back cover are prepended and appended to the document to conform to modern SAND report style guidelines. The first four chapters of this report were updated and released under the title "Fission Product Behavior During Severe LWR Accidents: Recommendations for the MELCOR Code System. Volume I" and were made available by the U.S. NRC through the Adams database under accession number ML19227A327. No prior release of the remaining content of this report has occurred.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2020

Report on FY 2020 Testing in Support of the Development and Extension of ASME Code Rules for Grade 91

This report summarizes the FY 2020 research and activities at Oak Ridge National Laboratory in support of the ASME code rule development and code extension for Grade 91 (Gr. 91) steel. Incorporation of the Gr. 91 temperature-dependent fatigue curves in Section III Division 5 was one major task, and the associated major aspects brought up during the course of the ASME balloting process are documented in this report. Experiments and analysis continued on the development of the alternative creep-fatigue evaluation based on elastic–perfectly plastic (EPP) plus simplified model test (SMT) methodology. The focus of the study in FY 2020 was the evaluation of the hold-time effect at high cycle and low strain range region. Baseline testing in support of the evaluation of hold time effects was started and will continue in FY 2021. Major on-going experiments and activities in support of the ASME code development and extension of Gr. 91 will be completed in FY 2021.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2020

MSTAR2019 Code Description and User's Manual

This report describes an ideal cascade model of uranium enrichment that includes side feed and side product streams and flexible input options. It is based on the original MSTAR model developed by Ed Von Halle and implemented in a Visual Basic code. The current version allows the user to specify integer numbers of stages or the stage numbers of external flows instead of assays in those flows. The computational engine is written in FORTRAN-90 and is invoked by a user-friendly GUI written in C++. This version of MSTAR has been demonstrated to operate on Linux, Mac, and Windows platforms, and has undergone significant testing and quality analysis. A number of examples are presented to illustrate the operation of the code and guide the user. The code is extremely fast, and results are returned immediately. The input and output have been specially configured for analysis by the environmental sampling team of the International Atomic Energy Agency (IAEA). A number of additional output features are included to assist the user in visualizing computational results and downloading data to files for use in other analysis software.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2020

FY20 Status Report on Creep Rupture Testing to Support the Development of Alloy 709 Code Case

This report provides an update on the status of the creep rupture testing on the first commercial heat of Alloy 709 in support of the Alloy 709 Code Case development. A creep test laboratory has been set up at ANL that comprises of 15 indirectly loaded test frames capable of testing ASTM-size specimens of the commercial heat of Alloy 709. The systems are equipped with high temperature furnaces capable of operation up to 1000°C in an air environment. During FY18, uniaxial ASTM-sized creep specimens were fabricated from 9 Alloy 709 plates that were processed by AOD, ESR, and ESR-homogenized routes and subsequently solution annealed at 1050, 1100, and 1150°C, and creep-rupture tests were performed at 330MPa and 600°C. The test results led to the down selection of the ESR-1100°C condition to support the code case. In FY19, 15 creep-rupture tests of the ESR-1100°C material, requiring the use of Type K thermocouples, were initiated to support the 100,000-hour Code Case. 8 tests were completed in FY19. In FY20, 6 tests were completed at the time of writing this report. In addition, 1 more test were loaded in FY20, using Type S thermocouples, and was completed. This report presents the data from the completed tests in FY20 and also the microstructural characterization of the fractured specimens. The creep data enabled the preliminary analysis of the creep behavior of this material. The characterization showed that in brittle fractures, the pores were narrow and thin, and elongated in the direction perpendicular to the loading direction, while in ductile fractures, the pores were much rounder. Dense precipitates were observed in all the samples, the size of which grew with test temperature. Some of the precipitates could be identified to be M23C6 based on their cuboid shape.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Collaborative Research: Improved Efficiency and Coupling of the Radiation Code in the ACME Earth System Model. Final Report

This final report details all work performed on the project by both project partners. This project provided support to properly couple RTE+RRTMGP, a high-performance broadband radiation code, within DOE’s Energy Exascale Earth System Model (E3SM). RTE+RRTMGP is a successor to the RRTMG radiation code, which has been widely accepted for its speed and accuracy by the global modeling community, and has been in use in the NCAR CESM for many years and was implemented in the initial version of E3SM. However, the computational cost of RRTMG remains high relative to other components in part due to its complexity and to its inefficient use of modern optimization strategies, issues that were rectified by the development of RTE+RRTMGP. Many of the accomplishment in this project necessitated significant collaboration with the E3SM development team. One focus of the project was to enhance the code’s optimization on the limited number of emerging computing systems on which the model is expected be used, including Many Integrated Core (MIC) architectures and Graphics Processing Unit (GPU) hardware. We also developed additional capabilities for RTE+RRTMGP that E3SM scientists identified as important for the planned applications of the model. The result of our project was optimization of a key physical component (radiative transfer calculations) of E3SM, directly supporting E3SM’s overarching global modeling objectives. More broadly, this project provided overall advancements in the use of radiative transfer calculations in atmospheric modeling and simulation, particularly for climate.

54 ENVIRONMENTAL SCIENCES↗
DOE OSTI2020

Development and Validation of a Two-Phase Thermal-Hydraulic CFD Code NEK-2P

A project is underway to develop, verify and validate an advanced two-phase flow modeling capability for the highly-scalable, high-performance Computational Fluid Dynamics (CFD) code NEK5000. The goal of this work is to verify and validate the two-phase version of the NEK5000 code, named NEK-2P, to simulate the two-phase flow and heat transfer phenomena that occur in a Boiling Water Reactor (BWR) fuel bundle under various operating conditions. The NEK-2P two-phase flow models follow the approach used for the Extended Boiling Framework (EBF) previously developed at Argonne but include more fundamental physical models of boiling phenomena and advanced numerical algorithms for improved computational accuracy, robustness, and computational speed. The development of the NEK-2P two-phase solver and the implementation of the Extended Boiling Framework two-phase models were initially supported by Argonne National Laboratory (Argonne) through a Laboratory Directed Research and Development (LDRD) project during FY2014-2016. The development and validation of the two-phase models through analyses of selected two-phase boiling flow experiments was supported by the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program in FY2017-2020. This report focuses on verification and validation of the water-steam boiling model NEK-2P Two-Phase, CFD code. The NEK-2P was validated with Nuclear Power Engineering Corporation (NUPEC) Pressurized Water Reactor (PWR) Sub-channel and Bundle Test (PSBT) void distribution benchmark. Three different simulations were performed and analyzed for various operating conditions such as wall-heat flux and sub-cooled inlet temperatures. Reasonably good agreement with measured data was obtained in predicting the measured void distributions. Simulations were performed for Virginia Tech. (VT) 3x3 rod bundle geometry with and without spacers. The preliminary results were presented for Simplified Spacer Grid (SSG). In addition, the implementation of interface reconstruction model was tested with one of the Becker benchmark Critical Heat Flux (CHF) experiments.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Economic Model For GDP And Tangible-Assets Loss Estimation In The MACCS Offsite Consequence Analysis Code

The MELCOR Accident Consequence Code System (MACCS) code is the NRC code used to perform probabilistic health and economic consequence assessments for atmospheric releases of radionuclides. MACCS is used by U.S. nuclear power plant license renewal applicants to support the plant specific evaluation of severe accident mitigation alternatives (SAMA) analyses as part of an applicant's environmental report for license renewal. MACCS is also used in severe accident mitigation design alternatives (SAMDA) and severe accident consequence analyses for environmental impact statements (EISs) for new reactor applications. The NRC uses MACCS in its cost-benefit assessments supporting regulatory analyses that evaluate potential new regulatory requirements for nuclear power plants. NRC regulatory analysis guidelines recommend the use of MACCS to estimate the averted "offsite property damage" cost (benefit) and the averted offsite dose cost elements.

29 ENERGY PLANNING, POLICY, AND ECONOMY↗
DOE OSTI2019

Multigroup Neutron Cross Section Generation for the SCEPTRE Code

Multigroup neutron cross sections were generated for the deterministic radiation transport code SCEPTRE. ENDF/B-VII.1 nuclear data files were downloaded from Los Alamos National Laboratory (LANL), processed with the LANL cross-section preparation code NJOY-2012, and post-processed to produce a SCEPTRE-formatted cross section file. A simple radiation transport problem was used to compare results calculated with MCNP, a continuous-energy radiation transport code from LANL, to results calculated with SCEPTRE using the NJOY-2012-produced multigroup cross sections. This problem was used to debug the python scripts used to post-process the NJOY-2012 output and to assess the accuracy of the multigroup cross sections. These comparisons demonstrate that the multigroup cross sections generated in this work are accurate for most elements but yielded inaccurate results for several common transition metals. This discrepancy appears to result from poor treatment of the resolved resonance region of the continuous-energy cross sections. Further work is recommended to investigate alternative methods to treat these resonances with NJOY-2012.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗