​​​​​​​​​Reports - Design, Methods and Va​lidation

63591 R1_AGC-4 Disassembly ReportDesign, Methods and Validation
The Advanced Reactor Terminology Graphite Research and Development program is currently measuring irradiated material property changes in several grades of nuclear graphite to predict behavior and operating performance within the core of these new high temperature reactor designs. The Advanced Graphite Creep (AGC) experiment, consisting of six irradiation capsules, will generate the irradiated graphite performance data for the Very High Temperature Reactor operating conditions. All six capsules in the experiment conducted at Idaho National Laboratory will be irradiated in the Advanced Test Reactor, disassembled in the Hot Fuel Examination Facility, and examined at the Idaho National Laboratory Research Center. This is the disassembly report describing the disassembly, shipment, post irradiation inspection, and storage of the graphite specimens contained within the AGC 4 irradiation test series capsule (the fourth irradiation capsule of the series).
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.
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.
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.
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.
Report on Year-2 of Water NSTF Matrix TestingDesign, Methods and Validation
This report serves as a summary of the testing activities conducted during the program’s second year of operation. Matrix testing began with establishing two-phase conditions that would serve as a common reference basis for nominal behavior, system parameters, and to monitor system repeatability. This two phase reference, or baseline test case, identified specific initial and boundary conditions that were derived from openly available literature of the full scale reactor concept.
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.
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.
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.