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At least 433 records · Page 24
All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

STAT7 v1.2 User Guide: The STAT7 Code for Statistical Propagation of Uncertainties in Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for thermal hydraulics analyses of research and test reactors. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to the fit functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one latera stripe of a fuel element. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2020

Light Water Reactor Sustainability Program Materials Research Pathway Technical Program Plan

The Materials Research (MR) Pathway within the Light Water Reactor Sustainability (LWRS) Program is charged with performing the research and development (R&D) to develop the scientific basis for understanding and predicting long-term environmental degradation behavior of materials in nuclear reactors. Furthermore, it is essential to use the mechanistic understanding of degradation phenomena in materials to develop mitigation, repair, and new material alternatives for existing components. The work will provide data and methods to assess performance of systems, structures, and components (SSCs) essential to safe and sustained reactor operations. The R&D products developed from the LWRS Program will be used by utilities, industry groups, and regulators to inform operational and regulatory requirements for materials in reactor SSCs subjected to long-term operation conditions, providing key input to both regulators and industry. The intent of this research is to help in reducing the operating costs, which may be in the form of offset maintenance costs due to better predictive models for component lifetimes, improved analysis of materials through nondestructive evaluation, reduced costs for repairs, or extended performance of plants through the selection of improved replacement materials. To best achieve this, industry experience and guidance are important as is their role in coordinated or collaborative research projects. The objectives of this report are to describe the motivation and organization of the MR Pathway within the LWRS Program, provide details on the individual research tasks within the MR Pathway, describe the outcomes and deliverables of the MR Pathway, including recent technical highlights and progress, and list the requirements for performing the research.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2023

Collaborative Study on Weldability of Irradiated Stainless Steel 304 Materials Removed from The NRU Reactor

A collaborative research on developing advanced welding technologies for irradiated stainless steel 304 materials between CNL and ORNL has been established to support both U.S. and Canadian interests in evaluation of weld repair techniques on irradiated materials to support continuous operation of commercial nuclear power. The work utilizes unique material from the National Research Universal (NRU) reactor and the specialized welding hot cell facility at ORNL. The objective is to explore suitable welding technique and parameters and to determine the helium concentration limitation in terms of irradiated stainless-steel weldability. An additional objective is to develop the knowledge and understanding of cracking mechanisms induced by high concentration of helium in the process of weld repair of the irradiated structural alloys. This report summarizes the experimental welding evaluation on irradiated material from the NRU reactor containing helium concentrations ranging from 12 to 45 atomic parts per million (appm).

36 MATERIALS SCIENCE↗
DOE OSTI2022

Light Water Reactor Sustainability: Program Materials Research Pathway (FY22 Technical Program Plan)

The Materials Research (MR) Pathway within the Light Water Reactor Sustainability (LWRS) program is charged with performing R&D to develop the scientific basis for understanding and predicting the long-term environmental degradation behavior of materials in nuclear reactors. Furthermore, the mechanistic understanding of degradation phenomena in materials must be leveraged to develop mitigation and repair strategies and new material alternatives for existing components. This research will provide data and methods to assess the performance of systems, structures, and components essential to safe and sustained reactor operations. The R&D products developed from the LWRS program will be used by stakeholders—including utilities, industry groups, and regulators—to inform operational and regulatory requirements for materials in reactor systems, structures, and components subjected to long-term operation conditions, providing key inputs to regulators and industry. Therefore, the intent of this research is to provide options to reduce the operating costs, which may be in the form of offset maintenance costs due to better predictive models for component lifetimes, improved analyses of materials through nondestructive evaluation, reduced costs for repairs, or extended performance of plants through the selection of improved replacement materials. To provide the best options, industry experience and guidance are important because of their role in coordinated or collaborative research projects. The objectives of this report are to describe the motivation and organization of the MR Pathway within the LWRS program; provide details on the individual research tasks within the MR Pathway; describe the outcomes and deliverables of the MR Pathway, including recent technical highlights and progress; and describe the requirements for performing this critically important research.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

CRADA Number NFE-19-07859 with Purist, Inc. (CRADA Final Report)

Cooperative Research and Development Agreement (CRADA) NFE-19-07859 between Oak Ridge National Laboratory (ORNL) and Purist Inc. (Purist) focused on developing a technology for production of high purity radioisotopes using small-scale underutilized research nuclear reactors. Purist is developing a novel production mechanism that will produce, and efficiently isolate radioisotopes with high specific activity at small-scale research reactors. The goal of this technology is to enable underutilized smaller scale reactors, typically not capable of producing medical grade radioisotopes, to be utilized as a production supply source. Such radioisotope production facilities will complement production efforts of the few larger production facilities to reduce and prevent risk in the medical isotope supply chain. The work done under this CRADA explored target development, to maximize isotope production and separation using Purist’s technology to ultimately obtain a radioisotope product with high specific activity for use in medical applications. Furthermore, under this CRADA post production, capture, concentration and encapsulation of the radioisotope product was studied by developing post irradiation midi to microscale processes to capture, store and release the separated radioisotopes from large volumes of the capture matrix after production, via automated High-Pressure Ion Chromatography (HPIC) and microfluidic purification/separation systems. This will enable concentration of the radioisotope product obtained into small volumes, which will aid in packaging and transporting the final radioisotope product to users.

07 ISOTOPE AND RADIATION SOURCES↗
DOE OSTI2025

STAT7 v2.0 User Guide

The STAT7 software was developed to perform steady-state, single-phase thermal hydraulics analysis of plate-fueled reactors based on statistical propagation of uncertainties. Application of the software includes non-power research and test reactor analysis, and it has been used for the conversion to low- enriched uranium fuel of U.S. High Performance Research Reactors such as the Massachusetts Institute of Technology Research Reactor. Since it can be necessary to repeat reactor safety analysis, such as during fuel reloading, STAT7 accommodates flexibility in analyzing many practical aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model uncertainty in common fuel fabrication parameters and other key reactor operating parameters required for reactor thermal hydraulics analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux, or onset of flow instability. STAT7 supports water properties based on the IAPWS-IF97 functions (The International Association for the Properties of Water and Steam Industrial Formulation 1997 for the Thermodynamic Properties of Water and Steam) in addition to fitted functions. STAT7 predicts axial profiles of fuel, cladding, and coolant temperature along a lateral stripe that runs the full length of the fuel plate from the bottom to the top. STAT7 can simultaneously analyze every axial node in each lateral stripe of all fuel plates and coolant channels in every fuel element of an entire reactor core. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. In a single execution, STAT7 can be used to perform full core analysis by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each fuel element in the core.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2022

Multiphysics Analyses of the Bottom Components of the 3D Printed Transformational Challenge Reactor

This research presents multi-physics analyses on the bottom components of the Transformational Challenge Reactor (TCR) facility. These components include the bottom axial reflector and the steel exit cone. The bottom axial reflector is made of pure silicon carbide elements hosting helium cooling channels These elements are 3D printed and therefore can host any arbitrary shape of the helium cooling channels. The design of the bottom reflector considers the neutronics and thermo-fluid dynamics performances as well as the manufacturing process optimization. More precisely, the best design of the bottom reflector reduces neutron leakage by avoiding straight cylindrical helium channels that facilitate neutron leakage, minimizes the helium flow pressure drop, and reduces the number of 3D printed silicon carbide pieces. The exit cone steel structure collects the hot helium from the bottom fuel assemblies and channels the cold helium to the top of the fuel assemblies. The steel simultaneous contact with hot and cold helium flows sets a large thermal gradient. Different designs of the exit cone are proposed to reduce the steel equivalent stress from the helium thermal load. The multi-physics analyses have been performed using Ansys Fluent, Ansys Mechanical, STAR-CCM+, and Serpent computer programs.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
NASA NTRS1995

Safety analysis

We are engaged in a research program in safety-critical computing that is based on two case studies. We use these case studies to provide application-specific details of the various research issues, and as targets for evaluation of research ideas. The first case study is the Magnetic Stereotaxis System (MSS), an investigational device for performing human neurosurgery being developed in a joint effort between the Department of Physics at the University of Virginia and the Department of Neurosurgery at the University of Iowa. The system operates by manipulating a small permanent magnet (known as a 'seed') within the brain using an externally applied magnetic field. By varying the magnitude and gradient of the external magnetic field, the seed can be moved along a non-linear path and positioned at a site requiring therapy, e.g., a tumor. The magnetic field required for movement through brain tissue is extremely high, and is generated by a set of six superconducting magnets located in a housing surrounding the patient's head. The system uses two X-ray cameras positioned at right angles to detect in real time the locations of the seed and of X-ray opaque markers affixed to the patient's skull. the X-ray images are used to locate the objects of interest in a canonical frame of reference. the second case study is the University of Virginia Research Nuclear Reactor (UVAR). It is a 2 MW thermal, concrete-walled pool reactor. The system operates using 20 to 25 plate-type fuel assemblies placed on a rectangular grid plate. There are three scramable safety rods, and one non-scramable regulating rod that can be put in automatic mode. It was originally constructed in 1959 as a 1 MW system, and it was upgraded to 2 MW in 1973. Though only a research reactor rather than a power reactor, the issues raised are significant and can be related to the problems faced by full-scale reactor systems.

Knight, John C.↗
DOE OSTI2024

STAT7 v1.2 Verification and Validation Report

This report documents the software testing which has been performed for the STAT7 Version 1.2 (hereafter STAT7 v1.2) software, a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The good agreement achieved in this testing campaign, within the acceptance criteria of a 1% relative difference, also demonstrates that these capabilities have been implemented in STAT7 v1.2 correctly.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2025

STAT7 v2.0 Verification and Validation Report

This report documents the software testing which has been performed for STAT7 Version 2.0 (hereafter STAT7 v2.0), a code for the statistical propagation of uncertainties in steady-state thermal hydraulics (TH) analysis of plate-fueled research and test reactors. The key capabilities addressed in this testing were selected based on the use of the software for analysis work supporting the conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors (USHPRR) such as the Massachusetts Institute of Technology Research Reactor (MITR-II). This testing scope includes comparison of the code’s statistical processing capabilities using statistical tests, comparison of coolant property evaluations to NIST data tables, and systematic comparison of key TH solver capabilities to the results of hand calculations and to predictions from PLTEMP/ANL Version 4.4. The

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2024

C.9 Scribe NMAC Layer Video Development

This task incorporates NMAC insider threat principles and good practices into video format for aiding in training activities. It was labeled as Task C.9 under the FY24 Annex 1 for NA-211 SSM A&I Pillar. For activity development, the setting of a research and test reactor, RTR, was selected as the focus of this fiscal year. More specifically, a TRIGA reactor was selected for facility and reactor inspiration. FY 23/24 developed materials Research Reactor Facility Measures against the Insider Threat, LLNL-MI-839144, was used as a reference for selecting relevant, suggested mitigation measures to incorporate into the script and subsequent videos. Throughout the initial phases of development, discipline SMEs met with Scribe SMEs from Sandia National Laboratory to create the initial story arc and assess the limits of Scribe/Odin technologies available for the project. As a result, the decision for a two-fold deliverable was developed and utilized as the basis for script development efforts over the first half of the fiscal year.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2023

The STAT7 Code for Statistical Propagation of Uncertainties in Steady State Thermal Hydraulics Analysis of Plate Fueled Reactors

The STAT7 software was developed to perform steady-state thermal hydraulic analyses. Application of the software is for non-power research and test reactors, including conversion to low-enriched uranium fuel of U.S. High-Performance Research Reactors such as MITR-II. Since it can be necessary to repeat analysis during fuel reloading, STAT7 accommodates flexibility in analyzing many realistic aspects of reactor fuel management. STAT7 uses a Monte Carlo approach to model common fabrication parameters and other key reactor analysis uncertainties required for research and test reactor thermal hydraulic analyses. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings on avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one lateral stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis can be performed by analyzing the margin to onset of nucleate boiling and onset of flow instability for each axial node of each stripe of each plate of each element in the core.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2021

Regulatory Research Planning for Micro-Reactor Development

A limited-scope Regulatory Research Development Plan (RRDP) is being developed to link important advanced non light water nuclear microreactor technology research and development (R&D) activities sponsored by the U.S. Department of Energy’s Office of Nuclear Energy (DOE NE), Science, and Technology to regulatory requirements and critical licensing needs likely to impact the deployment of a domestic commercial microreactor fleet. This document is a precursor to that forthcoming limited scope microreactor RRDP. This report is being published to aid in the identification of key fiscal year (FY) 2022 topical research opportunities (as conveyed to the RRDP development team) by respondent microreactor developers and suppliers in consideration of micro-reactor regulatory needs and challenges noted by industry, DOE, and Nuclear Regulatory Commission (NRC) stakeholders. Although this document does not contain the regulatory effects analysis and R&D recommendations that is expected in the final RRDP, this document is being released now to assist DOE researchers and investigators in FY22 research planning activities.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
NASA NTRS1976

Circulation system for flowing uranium hexafluoride cavity reactor experiments

Research related to determining the feasibility of producing continuous power from fissile fuel in the gaseous state is presented. The development of three laboratory-scale flow systems for handling gaseous UF6 at temperatures up to 500 K, pressure up to approximately 40 atm, and continuous flow rates up to approximately 50g/s is presented. A UF6 handling system fabricated for static critical tests currently being conducted is described. The system was designed to supply UF6 to a double-walled aluminum core canister assembly at temperatures between 300 K and 400 K and pressure up to 4 atm. A second UF6 handling system designed to provide a circulating flow of up to 50g/s of gaseous UF6 in a closed-loop through a double-walled aluminum core canister with controlled temperature and pressure is described. Data from flow tests using UF6 and UF6/He mixtures with this system at flow rates up to approximately 12g/s and pressure up to 4 atm are presented. A third UF6 handling system fabricated to provide a continuous flow of UF6 at flow rates up to 5g/s and at pressures up to 40 atm for use in rf-heated, uranium plasma confinement experiments is described.

Jaminet, J. F.↗
NASA NTRS1991

Vapor core propulsion reactors

Many research issues were addressed. For example, it became obvious that uranium tetrafluoride (UF4) is a most preferred fuel over uranium hexafluoride (UF6). UF4 has a very attractive vaporization point (1 atm at 1800 K). Materials compatible with UF4 were looked at, like tungsten, molybdenum, rhenium, carbon. It was found that in the molten state, UF4 and uranium attacked most everything, but in the vapor state they are not that bad. Compatible materials were identified for both the liquid and vapor states. A series of analyses were established to determine how the cavity should be designed. A series of experiments were performed to determine the properties of the fluid, including enhancement of the electrical conductivity of the system. CFD's and experimental programs are available that deal with most of the major issues.

Diaz, Nils J.↗
NASA NTRS1990

Soil bed reactor work of the Environmental Research Lab. of the University of Arizona in support of the research and development of Biosphere 2

Research at the Environmental Research Lab in support of Biosphere 2 was both basic and applied in nature. One aspect of the applied research involved the use of biological reactors for the scrubbing of trace atmospheric organic contaminants. The research involved a quantitative study of the efficiency of operation of Soil Bed Reactors (SBR) and the optimal operating conditions for contaminant removal. The basic configuration of a SBR is that air is moved through a living soil that supports a population of plants. Upon exposure to the soil, contaminants are either passively adsorbed onto the surface of soil particles, chemically transformed in the soil to usable compounds that are taken up by the plants or microbes as a metabolic energy source and converted to CO2 and water.

Frye, Robert↗
DOE OSTI2023

Fiscal Year 2023 Software Quality Assurance Activities for the ARC Software

The Argonne Reactor Code (ARC) software suite [1-17] has been developed by Argonne researchers for fast reactor design and analysis since the 1970s. With the ARC software suite, a user can quickly build a model of a proposed or existing fast spectrum reactor and carry out fuel cycle, nominal thermal analysis and flow requirements, and assess, as is appropriate, whether the core design and constraint system yield an acceptable mechanical behavior. For transient reactor analysis with SAS4A [18], the ARC software suite can be used to generate reactivity coefficients and kinetics parameters at any modeled fuel cycle time point which forms part of the input to SAS4A. The ARC suite was consistently being developed until the 1990s and followed a software QA program which was an appropriate standard for the time. In the 1990s, the DOE funding to fast reactor research and development was all but eliminated and the ARC software was put into maintenance mode. In the early 2000s, the software quality assurance (SQA) program for ARC was still in place to define an official version, but by 2005 it all but was abandoned as there were insufficient staff to fill the work roles. Since 2005, there has been a considerable increase in research and design work on fast spectrum reactors. The ARC software as a whole has since been exported to many universities and commercial companies and ANL support has been given to the various projects over the years [19-23]. Further, MC 2 -3, PERSENT, and DASSH were all developed after 2005 without any adherence to a software standard. In recent time, the DOE VTR project [22] paid for verification work to be done on the ARC software as part of the goal of making it NQA-1 complaint. The VTR project was not considered the appropriate pathway to fund and maintain a SQA program for the ARC software and while software developments (DASSH) were made and several manuals were updated and software verification work was carried out, the ARC software is not NQA-1 compliant. More recently the Advanced Reactor Development Program (ARDP [23]) has funded the creation of manuals for some ARC utility programs and funded additional software verification work on DIF3D [6, 7] and MC 2 -3 [2-5] for the purpose of commercial grade dedication. Because of the VTR and ARDP projects, software verification work was completed on MC 2 -3 and DIF3D, and detailed reports were created for each piece of software, which discuss the inputs and outputs from the codes that are covered by the verification work and link various analytic, code-to-code, and hand calculation based verification work presented in the report with verification test problems provided with the software. This is a key part of the commercial grade dedication work and constitutes the bulk of the cost to get the ARC software to commercial grade. The ARC software suite is a valuable asset as a fast reactor design and analysis tool set that has been reasonably well verified and validated with various fast reactor benchmark problems and experiments over decades. Some or all of the ARC software suite has been utilized for designing the IFR [20], PGSFR [21], VTR [22], and Natrium [23] reactors and we can expect it to continue to be used for advanced fast reactor design and/or confirmatory calculation purposes in the future. Due to increased interest by commercial companies and regulatory bodies, it is becoming more important to make the ARC software suite complete and ready-to-use in terms of its SQA pedigree and commercial grade dedication needs. This report discusses the achievements made towards building a new SQA program for the ARC software and dealing with outstanding identified QA gaps.

97 MATHEMATICS AND COMPUTING↗