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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 253 records · Page 14
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Update on ORNL TRANSFORM Tool: Simulating Multi-Module Advanced Reactor with End-to-End I&C

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the fourth year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled reactor) concepts, including the use of multiple coupled reactors at a single site. The focus of this report is the development of a steam generator and drum system model that includes the complex dynamics of typical steam drum systems, the development of instrumentation and controls for the steam generator with drum system model, and the development of multi-reactor module models that reflect the full power reactor innovative small module design concept. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor models; ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface technical area; and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environment and suite of models are identified as the TRANSFORM tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the Advanced Reactors Technology program; (2) developing a library of baseline component modules that can be assembled into full plant models using available geometry, design, and thermal-hydraulic data; (3) defining modeling conventions for interconnecting component models; and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Status of the phenomena representation, 3D modeling, and cloud-based software architecture development

Early in 2013, researchers at the Idaho National Laboratory outlined a technical framework to support the implementation of state-of-the-art probabilistic risk assessment to predict the safety performance of advanced small modular reactors. From that vision of the advanced framework for risk analysis, specific tasks have been underway in order to implement the framework. This report discusses the current development of a several tasks related to the framework implementation, including a discussion of a 3D physics engine that represents the motion of objects (including collision and debris modeling), cloud-based analysis tools such as a Bayesian-inference engine, and scenario simulations. These tasks were performed during 2015 as part of the technical work associated with the Advanced Reactor Technologies Program.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2015

Reference Operational Concepts for Advanced Nuclear Power Plants

This report represents the culmination of a four-year research project that was part of the Instrumentation and Control and Human Machine Interface subprogram of the DOE Advanced Reactor Technologies program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Report on the completion of the procurement of the first heat of Alloy 709

This report provides details on the completion of the procurement of the first commercial-sized heat of Alloy 709. The report is a Level 3 deliverable in FY17 (M3AT- 17OR1602053), under the Work Package AT-17OR160205, “Advanced Alloy Development” performed by Oak Ridge National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Technologies (ART). This work project supports the fabrication scale up effort for Alloy 709 that was started in FY16. The effort culminated in the placement of a Purchase Order in August 2016 with a commercial vendor to melt an Alloy 709 heat using industrial melt practice. Four ingots, totaling about 45,000 lb, had been bottom-poured from the melt in September 2016. Two of the ingots were hot rolled into 1.2”x60”x155” and 1.1”x60”x100” plates using standard hot rolling process in FY17. Some small test pieces were cut from the asrolled plates and sent to Argonne National Laboratory (ANL) for archival. The plates were then heat treated and surface pickled by the vendor. The plates were subsequently delivered to ANL and Oak Ridge National Laboratory (ORNL). Properties and microstructure screening were performed on these plates upon delivery in February 2017 at ANL. Several samples were cut from the as-rolled and heattreated plates and were analyzed for their microstructures, hardness values, grain sizes, and room temperature tensile properties. The results indicate that the scaled-up heat of Alloy 709 fabricated using commercial practice exhibit tensile properties that exceed the minimum values specified in the ASME Code Case for commercial heat of NF709. These plates will be used to support base metal testing for the 650°C, 100,000-h Alloy 709 Code Case development, for fabrication of weldments, and for the NEUP projects.

36 MATERIALS SCIENCE↗
DOE OSTI2017

FY 2017 – Thermal Aging Effects on Advanced Structural Materials

This report provides an update on the evaluation of the effect of thermal aging on tensile properties of existing laboratory-sized heats of Alloy 709 austenitic stainless steel and the completion of effort on the thermal aging effect on the tensile properties of optimized G92 ferritic-martensitic steel. The report is a Level 3 deliverable in FY17 (M3AT-17AN1602081), under the Work Package AT-17AN160208, “Advanced Alloy Testing - ANL” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Understanding and Predicting Effect of Sodium Exposure on Microstructure of Grade 91 Steel

This report provides an update on the understanding of the effect of sodium exposures on microstructure and tensile properties of Grade 91 (G91) steel in support of the design and operation of G91 components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602018), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Report on Understanding and Predicting Effects of Thermal Aging on Microstructure and Tensile Properties of Grade 91 Steel for Structural Components

This report provides an update on understanding and predicting the effects of long-term thermal aging on microstructure and tensile properties of G91 to corroborate the ASME Code rules in strength reduction due to elevated temperature service. The research is to support the design and long-term operation of G91 structural components in sodium-cooled fast reactors (SFRs). The report is a Level 2 deliverable in FY17 (M2AT-17AN1602017), under the Work Package AT-17AN160201, “SFR Materials Testing” performed by the Argonne National Laboratory (ANL), as part of the Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

Creation of the NaSCoRD Database

This report was written as part of a United States Department of Energy (DOE), Office of Nuclear Energy, Advanced Reactor Technologies program funded project to re-create the capabilities of the legacy Centralized Reliability Database Organization (CREDO) database. The CREDO database provided a record of component design and performance documentation across various systems that used sodium as a working fluid. Regaining this capability will allow the DOE complex and the domestic sodium reactor industry to better understand how previous systems were designed and built for use in improving the design and operations of future loops. The contents of this report include: overview of the current state of domestic sodium reliability databases; summary of the ongoing effort to improve, understand, and process the CREDO information; summary of the initial efforts to develop a unified sodium reliability database called the Sodium System Component Reliability Database (NaSCoRD); and explain both how potential users can access the domestic sodium reliability databases and the type of information that can be accessed from these databases.

96 KNOWLEDGE MANAGEMENT AND PRESERVATION↗
DOE OSTI2017

FY 2017-Influence of Sodium Environment on the Tensile Properties of Advanced Alloys

This report provides an update on the understanding of the effects of sodium exposures on tensile properties of advanced alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY17 (M3AT-17AN1602093), under the Work Package AT-17AN160209, “Sodium Compatibility” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program. Three laboratory-size heats of Alloy 709 austenitic steel were investigated in liquid sodium environments at 550-650°C to understand its corrosion behaviour, microstructural evolution, and tensile properties. In addition, a commercial scale heat has been produced and hot-rolled into plates.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Final Project Report on Computational Modeling and Analysis of Air-Based NSTF

The Reactor Cavity Cooling System (RCCS) is a passive safety concept under consideration for the overall safety strategy of advanced reactors such as the High Temperature Gas-Cooled Reactor (HTGR). The air-cooled RCCS uses natural convection to drive the flow of air from outside the power plant to remove decay heat during normal operation and accident scenarios. The Natural convection Shutdown heat removal Test Facility (NSTF) at Argonne National Laboratory (“Argonne”) is a ½ - scale model of the primary features of one conceptual air-cooled RCCS design. The NSTF program performs ex-vessel, passive heat removal experiments in support of Department of Energy Office of Nuclear Energy’s Advanced Reactor Technology (ART) program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Implementation of Software QA for SAS4A/SASSYS-1 Rev. 1

In an effort to support closure of gaps in advanced reactor licensing pathways, a multi-year task to systematically develop sodium-cooled fast reactor (SFR) codes and methods for use in a licensing framework was launched as part of the DOE-NE Advanced Reactor Technologies (ART) Program’s Regulatory Technology Development Plan (RTDP). This report documents Phase 2 of this task, which was focused on the finalization and initial implementation of a software quality assurance (SQA) program for SAS4A/SASSYS-1 and completion of high-priority SQA-related technical activities. The ultimate goal of this ongoing RTDP task is to develop an SQA framework that satisfies the fundamental requirements of relevant SQA standards and guidance while still maintaining sufficient flexibility and efficiency such that the framework can be sustained with modest resources.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Molten Salt Reactor Campaign Modeling and Simulation Program Plan

The modeling and simulation activities within the Molten Salt Reactor Research and Development Campaign of the US Department of Energy Office of Nuclear Energy Advanced Reactor Technologies Program facilitate the development and application of analysis tools for the licensing and design of molten salt reactors.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Report on the initiation of planned FY18 short and intermediate term creep rupture tests and creep-fatigue tests on the first commercial heat of Alloy 709

This report provides an update on the initiation of planned FY18 short and intermediate term creep rupture tests and creep-fatigue tests on the first commercial heat of Alloy 709 in support of the design and operation of structural components in sodium-cooled fast reactors (SFRs). The report is a Level 3 deliverable in FY18 (M3NT-18AN050502031), under the Work Package NT-18AN05050203, “A709 Code Case Testing” performed by Argonne National Laboratory (ANL), as part of Advanced Reactor Technologies Program.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Report on the completion of the development of processing map from as-cast Alloy 709 materials

This report provides details on the completion of the development of the processing map from the as-cast Alloy 709 materials. The report is a Level 4 deliverable in FY18 (M4NT18AN050502012), under the Work Package NT-18AN05050201, “A709 Development” performed by Argonne National Laboratory, as part of Advanced Structural Materials Program for the Advanced Reactor Technologies (ART). The report describes the development of a processing map to guide the development of hot work regimens for A709. The maps combine high-temperature flow stress measurements, described in a companion report by Oak Ridge National Laboratory, and metallographic observations, described here, into a single diagram that can be used to select hot working parameters to avoid regions of plastic instability and to avoid the development of unfavorable microstructures. This report describes the process of creating the map, provides the final map for A709, and uses the processing map to describe two potential hot work regimens that could be used in processing a DOE round ingot into bar stock.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2018

Multi-Physics Simulations for Molten Salt Reactor Evaluation: Chemistry Modeling and Database Development

To aid in design and licensing of molten salt reactors, a framework integrating the complex interaction of reactor neutronics, thermal hydraulics, and chemistry is being developed within the Department of Energy Advanced Reactor Technology Program’s Molten Salt Reactor (MSR) campaign. The challenges of integrating thermochemical and thermophysical behavior into a multi-physics reactor simulations include the following: (1) population of data needed for refinement of current models and development of nonexistent models through experimental measurements, first principles calculations, and development of a machine learning approach (2) thermochemical and thermophysical model development, (3) further development of Thermochimica, an open-source efficient equilibrium solver used to link thermochemical models to the multi-physics code, (4) a framework for integrating kinetic phenomena: nucleation, precipitation, mass/heat transport, and corrosion models, and (5) a computational environment to efficiently utilize the data and models within a multi-physics modeling tool. These challenges are being addressed through a collaboration among Oak Ridge National Laboratory, Argonne National Laboratory, the University of South Carolina, and the University of Ontario Institute of Technology.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Scoping Analysis of Sodium Cooled Fast Spectrum Test Reactor Cores

A team from the national laboratories conducted a study that assessed advanced reactor technology options and the irradiation needs of a broad group of stakeholders. This identified a gap in the fast neutron irradiation capability - the need for a test reactor that can deliver high neutron fluxes of fast neutrons. The study compared fast and thermal neutron irradiation technology options based on technology maturity and the ability to meet strategic objectives, along with pathways to deployment and tradeoffs in mission. It also provided a recommendation on specific technology options to address the current gap in fast neutron irradiation capabilities. In this context, a detailed scope of work was prepared that pertains to a trade study focusing on the core of a fast neutron spectrum test reactor. This report presents the main results obtained by a team of Argonne National Laboratory (ANL) and Idaho National Laboratory (INL) scientists. It aims at developing an overall understanding of how the various areas important to the design and successful operation of such a facility interact with one another. Developing a sense of the relevant sensitivities at a systems level creates an understanding of where and how margins may be available to improve performances (e.g., flux level, capacity factor, and thermal hydraulic limits). Emphasis is given to the reactor (core design options and safety) and its fuel cycle (fuel supply and fabrication, as well as spent fuel management).

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2019

Baseline Characterization Database Verification Report – IG-110 Billet 10X69

The purpose of this engineering calculations and analysis report (ECAR) is to present the data being collected in the Baseline Graphite Characterization program. This program is directly tasked with supporting Idaho National Laboratory’s (INL’s) research and development efforts in the Advanced Reactor Technologies (ART) program. This program populates a comprehensive database that will reflect the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical- and mechanical-property information being collected will be transferred to the Nuclear Data Management and Analysis System (NDMAS), and that database will help populate the handbook of property data available to member nations of the Generation-IV International Forum. The transfer of these data from the applicable technical lead to the dissemination databases available to other end users requires a full review of test procedures and data-collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents that analysis for IG-110, Billet 10X69 and facilitates the release of the associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2019

Baseline Characterization Database Verification Report – 2114 Billet A20570

The purpose of this engineering calculations and analysis report (ECAR) is to present data collected in the Baseline Graphite Characterization Program, which is directly tasked with supporting the Idaho National Laboratory’s (INL’s) research and development efforts on the Advanced Reactor Technologies (ART) Program. This program populates a comprehensive database that reflects the baseline properties of nuclear-grade graphite with regard to individual grade, billet, and position within individual billets. The physical- and mechanical-property information being collected will be transferred to the Nuclear Data Management and Analysis System (NDMAS), and that database will help populate the handbook of property data available to member nations of the Generation-IV International Forum. Transfer of these data from the applicable technical lead to the dissemination databases available to other end users requires a full review of the test procedures and data-collection efforts through an analysis of the multiple summary spreadsheets and values being collected. This report represents the analysis for 2114 Billet A20570 and facilitates release of associated data to the NDMAS custodians.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗