Advanced Reactor Technologies (ART): Very High Temperature Reactor (VHTR) Research and Development (R&D) Quarterly Report
Highlights of ART TDO activities October through December 2017
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Results for “TRISO”
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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Conceptual Design Report for the Xe-6/7/8 Experiment in the NE Flux Trap
X energy, LLC has initiated conceptual design of a high-temperature, pebble-bed reactor and requested that Idaho National Laboratory evaluate the feasibility of irradiating several fuel pebbles in the Advanced Test Reactor. The proposed fuel for the X energy reactor utilizes the same tristructural-isotropic coated fuel particles as are used in the Department of Energy-sponsored Advanced Gas Reactor (AGR) fuel tests, which have been irradiated in the ATR since 2006. The X energy test has been denoted XE 6/7/8 based on nomenclature from the AGR fuel-test series. Three tests will be combined into one test train: XE 6 constitutes the fuel qualification test, whose aim is to demonstrate fuel performance at the upper end of prototypical reactor conditions; XE 7 is designated a “margin test,” and its purpose is to demonstrate fuel integrity at conditions substantially beyond reactor operating conditions; and XE 8 incorporates designed-to-fail particles in order to release fission products and help validate fission product release codes. The XE series of experiments would consist of multiple separate capsules that would be irradiated simultaneously. Each capsule would have its own inert sweep-gas atmosphere with individual on line temperature monitoring. The sweep-gas effluent from each capsule would also have online fission product monitoring to track performance of the fuel in each individual capsule during irradiation.
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AGR-5/6/7 Final Fuel Characterization
presentation for program review
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AGR-1 Daily As-run Thermal Analyses
This report documents the results of thermal analyses to predict the daily as-run temperatures for the AGR-1 experiment. Control gas gaps and compact-graphite holder gas gaps were modeled to change linearly with time. The control gas gap widths were determined by comparison to measured R/B data. The compact graphite holder gas gaps were determined from PIE measurements. Daily heat rates for each compact and component in the models were input from daily as-run physics analyses. Daily gas compositions and component fast neutron fluences were also input. Six different finite element models were created for the six different AGR-1 capsules. Each capsule had a different gas gap that was implemented to control the temperature of the experiment. Capsules on the top and bottom had larger gas gaps, while capsules in the middle were smaller.
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AGR-5/6/7 Operations Update for Technical Coordination Team (Apr 2020)
This is a powerpoint presentation designed to update the Technical Coordination Team (TCT) associated with the AGR irradiation experiment program on operations of the AGR-5/6/7 experiment over the past six months (since the last TCT meeting). The presentation primarily deals with operational challenges to failed fuel particles in capsule 1. Specifically, how to keep fission products from capsule 1 from entering the other four capsules in the experiment.
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Summary of BISON Milestones: NEAMS FY2020 Report
This summary report contains an overview of work performed under the work package entitled "FY2020 NEAMS Advanced Fuels Performance'', which is focused on the development and support of the fuel performance code BISON. The second chapter lists FY20 milestones titles, completion schedule, and milestone level. Subsequent chapters summarize and demonstrate completion of the milestones. The last chapter outlines FY21 proposed future work. In FY20, the NEAMS program emphasized development of BISON for its application to advanced reactors. While there are a variety of advanced fuel concepts, based on interaction with industry and the Nuclear Regulatory Commission, the fuel types we chose to develop were metallic fast reactor, UN/UC and particle fuels. The last chapter of this report documents proposed work for FY21. We plan to continue work on metallic and particle fuel in terms of developing/calibrating models and to begin rigorous validation/assessment for both fuel types. Due to the merger of the NEAMS and CASL programs, FY21 will see a return to light water reactor model development and simulation; this time focused on advanced technology fuels. Additionally, we seek to improve BISON, fundamentally. As such, we plan improvements to BISON and MOOSE in terms of algorithmic robustness, performance, ease-of-use, and quality assurance.
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Matrix Graphite Material Models In Pebbles and Compacts For Bison
The cores and reflectors in high-temperature gas-cooled reactors (HTGRs) are made of graphite materials, with the graphite acting as a moderator, a fuel host matrix, or the foundation for various structural components. This study aims to survey the models in the literature for graphite materials being used as host matrices in pebble/fuel compacts and to implement those surveyed models into Bison to conduct an early assessment of graphite's thermo-mechanical response under various reactor conditions. In this study, thermal (e.g., thermal conductivity, and specific heat capacity) and mechanical (e.g., elastic properties, thermal expansion, irradiation-induced dimensional changes, and irradiation-induced creep) material models for various graphite grades (e.g., H-451, IG-110, G-348, 2020, A3-3, and A3-27) are incorporated into Bison. Two benchmark problems are then exercised utilizing these new graphite-related capabilities: (1) modeling an Advanced Gas Reactor (AGR)-2 fuel compact, and (2) modeling the debonding of a particle-matrix interface.
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Status of Modifications to the AGR-3/4 Fission Product Transport Model
A one-dimensional (1D) finite-element model of the AGR 3/4 experiment fission product distributions based on the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework has been revised and implemented within BISON, with improvements which allow for detailed temperature histories to be used for analyses. Empirically determined concentration profiles and ring inventories from the AGR-3/4 experiment have been compiled for comparison purposes. Ongoing efforts to build fission product diffusion models which can describe the applicable physics for isotopic concentration profiles in each AGR-3/4 capsule are discussed, as well as the suitability of each isotopic concentration profile in each capsule for obtaining transport parameters.
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The Use of the High-density UN Fuel in Heat-pipe Microreactors
Heat-pipe microreactors (HPMRs) are very small-scale nuclear reactors that employ heat pipes for heat removal. HPMRs can be easily integrated with other forms of renewable energies, used for emergency responses to disaster relief zones, deployed in remote locations not connected to the grid, and removed from sites and replaced by new ones. HPMRs can also be used for space missions as heat pipes do not rely on gravity for heat transfer. Conventional fuel materials, such as UO 2 and UCO, are currently considered in most existing HPMR designs, but ceramic UN fuel that has high uranium density, high thermal conductivity, and a high melting point may become a better fuel candidate.
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High-Fidelity Neutronics Model of a Realistic Heat Pipe Microreactor Report
The creation of a digital twin requires two parallel products: a virtual model and a physical asset. This report details the development the high-fidelity reactor physics model of a realistic heat pipe micro reactor for the virtual model. The realistic heat pipe micro reactor provides a generic model that research can utilize to make safeguards considerations. The model will be leveraged for capturing diversion and misuse scenarios expected to take place in a microreactor. A python wrapper has been developed to rapidly explore various acquisition pathways by manipulating the reactor physics model on-the-fly.
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Low-Power Nuclear Heating Tests of UN Fuel Particles in Hydrogen Environment for Nuclear Thermal Propulsion
Recently, there has been a reemergence of interest within the space-exploration community regarding nuclear thermal propulsion (NTP) for interplanetary travel. NTP, as conceptualized, is a compact propulsion source, is relatively light because there is no need for an oxidizing agent and has a very high specific impulse compared to traditional forms of propulsion. These factors would reduce first-leg orbital-rocketry costs (by the source’s low weight and volume) and decrease overall travel time for astronauts with higher transit speeds. Idaho National Laboratory was provided bare, uncoated, and tungsten-coated uranium nitride fuel particles as part of a preliminary investigation. The fuel was fabricated by BWX Technologies and irradiated at the Massachusetts Institute of Technology research reactor. X-ray computed tomography and scanning electron microscopy and associated chemical analysis techniques were performed on these particles at Idaho National Laboratory. Both coated and uncoated forms were examined in the as-fabricated and as-irradiated conditions. The tungsten coating was found to be intact following the irradiation, indicating positive attributes for performance under these low-power-irradiation conditions. The irradiation of uncoated particles resulted in nitrogen enrichment at the edge of the particle.
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Development and Demonstration of a Risk Assessment Approach for Approval of a Transportation Package of a Transportable Nuclear Power Plant for Domestic Highway Shipment
The U.S. Department of Defense (DoD) Strategic Capabilities Office (SCO) has tasked PNNL to address the regulatory challenges associated with confirming the safe transport of Transportable Nuclear Power Plants (TNPPs) containing irradiated nuclear fuel. A previous report—Proposed Risk-Informed Regulatory Framework for Approval of Microreactor Transportation Packages (PNNL-31867)—determined that the expected radioactive inventory in the irradiated fuel of a TNPP would likely require shipment in an NRC-approved Type B package (or spent nuclear fuel cask) but that a TNPP “package” is unlikely to meet the entire suite of NRC requirements set forth in Part 71 of Title 10 of the Code of Federal Regulations (CFR) for a Type B package. It was therefore concluded that shipment of this initial TNPP transportation package, as well as possibly others, under existing regulations would likely require NRC approval using the 10 CFR 71.12 (“Specific exemptions”) process that relies on risk-informed decision making supported by quantitative risk assessment (i.e., Probabilistic Risk Assessment).
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Simulations of Deimos High Assay Low Enriched Uranium Critical Experiment [Slides]
The Deimos experiment went critical on Sept 18, 2024 and was a successful demonstration for HALEU nuclear data validation at the NCERC facility. Much design and engineering effort was required for the experiment and was successfully completed.
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Improved Fuel Particle and Compact Matrix Fission Product Release Prediction with Improved Mass Transfer Model
This study was motivated by the need to incorporate more physics-based mass transfer models in BISON in order to improve fuel particle and compact matrix fission product release prediction. This document details the integration of new modeling capabilities in BISON, including (1) development of sorption mass transfer models; improvement of thermal contact model by adding neon to the gas gap inventory, (3) implementation of PCEA and IG-110 graphite thermal models, (4) validation using Advanced Gas Reactor (AGR)-3/4 compact and capsule rings data. These new capabilities have been shown to enforce the desired physics with satisfactory accuracy. BISON’s predictions of fission product release of AGR-3/4 compacts compare favorably with PARFUME and the concentration profiles across capsule rings show reasonable trend to the experiment measurement.
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Nuclear Forensics-by-Design for Material Security: Research Strategies and Outcomes
Explore the source record for details and available documents.
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CLIMATE CHANGE 2023 Synthesis Report Summary for Policymakers
This Synthesis Report (SYR) of the IPCC Sixth Assessment Report (AR6) summarises the state of knowledge of climate change, its widespread impacts and risks, and climate change mitigation and adaptation. It integrates the main findings of the Sixth Assessment Report (AR6) based on contributions from the three Working Groups, and the three Special Reports. The summary for Policymakers (SPM) is structured in three parts: SPM.A Current Status and Trends, SPM.B Future Climate Change, Risks, and Long-Term Responses, and SPM.C Responses in the Near Term. This report recognizes the interdependence of climate, ecosystems and biodiversity, and human societies; the value of diverse forms of knowledge; and the close linkages between climate change adaptation, mitigation, ecosystem health, human well-being and sustainable development, and reflects the increasing diversity of actors involved in climate action. Based on scientific understanding, key findings can be formulated as statements of fact or associated with an assessed level of confidence using the IPCC calibrated languages.
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