AGC-4 Experiment Irradiation Monitoring Data Qualification Final ReportGraphite R&RD
(December 2020) The Graphite Technology Development Program ran a series of six experiments to quantify the effects of irradiation on nuclear grade graphite. This report focuses on the fourth experiment, Advanced Graphite Creep 4 (AGC 4). The Advanced Reactor Development (ARD) Technology Development Office (TDO) Program for research and development activities require documentation of qualified monitoring data to design and license the first high-temperature reactor nuclear plant. Qualified data meets the requirements for use as described in the experiment planning and quality assurance documents. (
As-Run Physics Analysis for the AGC-4 Experiment Irradiated in the ATRGraphite R&RD
This ECAR documents the results of the ATR detailed physics analyses performed to calculate the displacements per atom and the fast neutron fluence (E > 0.1 MeV) of the Advanced Graphite Creep experiment, AGC-4, irradiated in the ATR East Flux Trap during ATR Cycle 157D, 158A, 162A, 162B, 164A, 164B, 166A, and Cycle 166B.
Advanced Microscopy Report on UCO Fuel Kernels from Selected AGR-1 and AGR-2 ExperimentsFuel Qualification
A variety of neutron irradiation experiments were completed at INL on TRISO coated particles contained in a graphitic matrix as part of the Advanced Gas Reactor (AGR) fuel development and qualification program. Although the initially advanced microscopy and microanalysis studies on AGR-1 particles focused predominantly on the SiC layer's role as the main fission-product containment, studies were further expanded to evaluate UCO kernels of selected AGR-1 and AGR-2 coated particles.
AGR-5/6/7 Data Qualification Report for ATR Cycles 162B through 168AFuel Qualification
This report provides the qualification status of experimental data for the AGR-5/6/7 fuel irradiation. AGR-5/6/7 was conducted in the ATR at INL in support of the development and qualification of TRISO low-enriched fuel for use in high-temperature gas-cooled reactors.
As-Run Thermal Analysis for the AGC-4 Experiment Irradiated in the ATRFuel Qualification
The Advanced Graphite Capsule (AGC) irradiation experiment will provide irradiation creep rate data for the new graphite proposed for the Next Generation Nuclear Plant (NGNP) program. The fourth experiment in the series (AGC-4) was designed to irradiate various types of graphite specimens at a temperature of 900 ºC and targeted displacements per atom (DPA) of 6.
Automated Internal Energy Calibration by OnTheFly for AGR-5/6/7Fuel Qualification
The software program, OnTheFly, was developed at INL to keep HPGe detectors energy calibrated during very long experiments. Over time, spectra produced from a HPGe detector will slowly stretch or contract, causing the energy calibration to change. If the energy calibration changes too much, it will cause energy lines to be misidentified.
Mechanical Properties of Aged A709High Temperature Materials R&D
A709 plate is in the process of being developed and qualified through a collaboration by Argonne National Laboratory (ANL), Idaho National Laboratory (INL), and Oak Ridge National Laboratory (ORNL). The goal is to qualify A709 plate in Section III, Division 5 of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (BPVC).
Status of INL Aged A709 Mechanical TestingFast Reactor Structural
A709 was selected to be qualified in Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code. Idaho National Laboratory (INL), Argonne National Laboratory and Oak Ridge National Laboratory are collaborating to develop and qualify A709 plate. Aging A709 to form beneficial precipitates prior to service is investigated. One tensile test, five creep-rupture tests, and eight cyclic tests have been completed on aged A709. Two creep-rupture tests are in progress.
Report Documenting Activity for Second Alloy 709 Commercial HeatHigh Temperature Materials R&D
Qualification of the advanced austenitic stainless steel material Alloy 709 for use in Section III Division 5 of the ASME Boiler and Pressure Vessel Code is being pursued by the DOE ART Program. Qualification of this alloy will allow its use in the design and construction of components for elevated temperature nuclear reactor systems.
AGR-2 Loose Particle Heating Tests in the Furnace for Irradiated TRISO TestingDesign, Methods and Validation
A furnace system was developed to thermally expose individual tristructural-isotropic (TRISO) coated particles to expand upon established postirradiation examination (PIE) safety testing that is used to explore particle failures and fission product release from compacts. Two systems were installed in the Irradiated Fuels Examination Facility at Oak Ridge National Laboratory.
Radial Deconsolidation and Leach-Burn-Leach of AGR-3/4 Compact 1-4 and 10-4Fuel Qualification
The Advanced Gas Reactor (AGR) Fuel Development and Qualification Program third and fourth irradiation experiments (AGR-3/4), originally planned as separate tests, were combined in one test train for irradiation in the Advanced Test Reactor at Idaho National Laboratory (INL).
Comparison of Fission Product Release Predictions using PARFUME with Results from the AGR-2 Irradiation ExperimentFuel Qualification
The PARFUME (PARticle FUel ModEl) code was used to predict fission product release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification Program.
Safety Testing and Destructive Examination of AGR-2 UO2 Compact 3-1-1Fuel Qualification
Post-irradiation examination and elevated-temperature safety testing are being performed on compacts from the Advanced Gas Reactor Fuel Development and Qualification Program’s AGR-2. The compacts in the AGR-2 irradiation experiment held either tristructural isotropic (TRISO)-coated particles containing uranium oxide fuel kernels or TRISO-coated particles containing fuel kernels with both uranium carbide and uranium oxide phases.
AGR-5/6/7 Post-Irradiation Examination PlanFuel Qualification
This plan describes the PIE activities to be carried out on AGR-5/6/7 fuel and the non-fuel components of the AGR-5/6/7 irradiation test train. PIE of AGR-5/6/7 fuel and test train components will provide data on fuel performance over a range of irradiation conditions (e.g., burnup, neutron fluence, and temperature), test fuel under postulated accident conditions, and support development of fuel performance and fission product transport models.
Comparison between PARFUME and Bison Using the AGR-2 Irradiation ExperimentFuel Qualification
This report documents comparisons between Fuel Model (PARFUME) model predictions versus Bison for selected compacts from the second irradiation test of the Advanced Gas Reactor (AGR) program that occurred from June 2010 to October 2013 in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL). PARFUME is a fuel performance analysis and modeling code, used for evaluating gas-reactor tristructural isotropic (TRISO) coated particle fuel for prismatic, pebble bed, plate, and cylindrical type fuel geometries.
NDMAS System and Process DescriptFuel Qualification
The DOE has made a significant investment in research to develop the next generation of reactor technologies as well as to improve the performance and lengthen the life cycle of existing nuclear reactors. Data collected to demonstrate new concepts may also be used in the future to support licensing of these technologies. Provenance of these data must be preserved. The NDMAS was established to manage and preserve data collected by fuels and materials research conducted by the high-temperature, gas-cooled reactor program.
AGR-5/6/7 Experiment Monitoring and Simulation ProgressFuel Qualification
Advanced Gas Reactor (AGR)--5/6/7 is the last of a series of AGR experiments conducted in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) in support of development and qualification of tri-structural isotropic (TRISO) low-enriched fuel for use in high temperature gas-cooled reactors. The test train contains five separate capsules that are independently controlled and monitored. Each capsule contains multiple 12.51 mm long compacts filled with low-enriched uranium carbide/oxide (UCO) TRISO fuel particles.
Technical Program Plan for INL Advanced Reactor Technologies Advanced Gas Reactor Fuel Development and Qualification ProgramHigh Temperature Test Facility
High-temperature gas cooled reactors (HTGRs) are graphite moderated nuclear reactors cooled with helium. Their high outlet temperatures and thermal energy conversion efficiency enable efficient and cost effective integration with non electricity generating applications. These applications include process heat and hydrogen production for petrochemical and other industrial processes that require operating temperatures between 300 and 900°C. HTGRs will supplement the use of premium fossil fuels such as oil and natural gas, improve overall energy security in the United States by reducing dependence on foreign fuels, and reduce carbon dioxide (CO2)/greenhouse gas emissions.
FY-20 GIF-RSWG and WGSAR Status ReportDesign, Methods and Validation
This report provides an end of year summary that reflects the progress and status of United States (U.S.) activities supporting the Global International Forum’s (GIF) Risk & Safety Working Group (RSWG) and the international regulators associated with the Working Group on the Safety of Advanced Reactors (WGSAR). This effort has commenced the development of a technology neutral, risk-informed approach for the selection of licensing basis events (LBE) and the assessment of defense in depth for Gen IV reactor technologies.
Effects of Notches on the Intermediate Creep-Rupture Life of Alloy 617 WeldmentFast Reactor Structural
A Code Case has recently been completed that adds Alloy 617 to Section III, Division 5 of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, which covers high temperature nuclear components. However, additional information is needed to address concerns raised by the Nuclear Regulatory Commission that are not covered by the Code Case.
Single-Bar SMT Testing on Alloy 617 using Software ControlsFast Reactor Structural
NL has performed simplified model testing using a variable in the software control program to compare with the testing performed at ORNL, which uses hardware to modify the control signal. The INL tests compared well with the ORNL tests, showing similar stress and strain values, as well as similarly shaped hysteresis loops. This demonstrates that possibility of performing simplified model testing without any hardware modifications.
AGC 3 Irradiated Material Properties AnalysisGraphite R&RD
This report documents the analysis of the irradiated material property data from the Advanced Graphite Creep (AGC)-3 graphite specimens. This is the third in a series of six irradiation test trains planned as part of the AGC experiment to fully characterize the neutron irradiation effects and radiation creep behavior of current nuclear graphite grades.
HDG-1 Graphite Preirradiation Data Package ReportGraphite R&RD
This report documents all pre-irradiation examination material-property measurement data for graphite specimens that are going to be used within the first high dose graphite (HDG) -1 irradiation capsule. The two new HDG capsules signify a major change to the AGC Experiment. HDG-1 and HDG-2 will replace the last two AGC capsules (AGC-5 and AGC-6) which were designed to irradiate graphite at the extreme upper operational temperatures for a VHTR design, 1100°C.
Tritium Transport Phenomena in Molten-Salt ReactorsDesign, Methods and Validation
In this work, we review phenomena relevant to tritium transport in molten-salt reactors, which produce tritium from lithium and beryllium salts at significantly higher levels than other reactor types. Modeling of such phenomena began following MSRE operations, and these early models attempted to predict measured tritium distributions in the MSRE, accounting for turbulent mass-transport processes (using established heat transfer correlations), permeation through a variety of metal structures such as heat exchanger tubes, and transport to and from bubbles introduced into the salt by gas sparging.
Establishing Regulatory Jurisdictional Boundaries at Collocated Advanced-Reactor FacilitiesDesign, Methods and Validation
This white paper discusses establishing and applying nuclear facility jurisdictional boundaries at advanced-nuclear-reactor facilities. It was written for industry review and evaluation with possible consideration by the U.S. Nuclear Regulatory Commission for subsequent regulatory action. The paper proposes a regulatory basis for establishing jurisdictional boundaries at operational advanced reactor installations at sites proximate to and sharing systems with non-NRC regulated facilities.
Status Report on ASME Code Development for Nonmetallic Core Components in 2020Graphite R&RD
The purpose of this report is to provide a status update on the progress and ongoing activities of the current ASME Boiler Pressure Vessel (BPV) Code Section III, Division 5, on nonmetallic core components and assemblies which includes the design rules for graphite and ceramic composite materials.
USDOE ART Graphite – Selection and Acquisition StrategyGraphite R&RD
The previous Selection and Acquisition Strategy document dates from 2007 . Since then, significant changes have occurred with reactor vendors and with the graphite manufacturers. There have also been changes in the membership of the Gen IV International Forum. These factors were considered when formulating a revised selection and selection strategy. In summary, the new graphite selection and acquisition plan is that no graphite billets will be purchased until there is new graphite production (NBG-18, PCEA). Existing supplies of NBG-17, NBG-18, PCEA, IG-110, 2114 are considered enough for future work.
Initial High Temperature Inelastic Constitutive Model for Alloy 617Design, Methods and Validation
This report describes progress on a Alloy 617 inelastic constitutive model slated for inclusion in the ASME Boiler & Pressure Vessel Code in Code Case N-898 “Use of Alloy 617 (UNS N06617) for Class A Elevated Temperature Construction, Section III, Division 5” as part of an appendix on inelastic modeling.
An Initial Assessment of the Design Margins of Different ASME Section III, Division 5 Design RulesDesign, Methods and Validation
This report develops a method for assessing the design margin of the ASME Section III, Division 5 Class A design rules, focusing on a definition of margin as the ratio between the actual, expected component life and the ASME design life. Full inelastic finite element simulations with a complete damage model capturing the available creep, creep-fatigue, fatigue, and tension test failure data produce the expected component lives, which can then be compared to corresponding ASME design calculations.
Preliminary design analysis workflow for Division 5 HHA-3200 requirements for graphite core componentsDesign, Methods and Validation
This report presents a design analysis workflow for graphite core components and assemblies, based on the design rules of ASME Boiler Pressure and Vessel Code, Section III, Division 5, Article HHA-3000. The workflow contains three stages: developing the design of the graphite core component, modeling the component with the finite element software MOOSE, and assessing if the component passes/fails the criteria of the HHA-3000 design rules.
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