Advanced Reactor Development: Gas-Cooled Reactor First Quarter Report
Significant and major accomplishments of the Advanced Reactor Development: Gas-Cooled Reactor Program for the first quarter of 2021.
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Results for “Nuclear Graphite”
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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Towards a Benchmark Experiment with the Compact Nuclear Power Source
The Compact Nuclear Power Source (CNPS) was a high-assay low enriched uranium (HALEU) tristructural isotropic (TRISO)-fueled, graphite-moderated microreactor constructed in 1987 at Los Alamos National Laboratory. The reactor, conceived as a power source for short-range radar stations, was designed to be "walk-away safe," and was cooled by heat pipes and ambient air. Though the project was formally cancelled after the fuel and moderator material had been received, its potential to meaningfully advance the body of critical and integral data was evident, and critical experiments with a mock-up of the reactor proceeded at TA-18 until its disassembly in 1991. The reflector and some components of the core would later go on to see service as part of the New Production Reactor Modular High-Temperature Gas-Cooled Reactor (NP-MHTGR) critical experiments. In light of the progress made by groups like Westinghouse and X-Energy towards contemporary graphite-moderated microreactors, the system remains an attractive candidate for the basis of a benchmark experiment even today. Uncertainties remain, however, in some features of the system–inconsistencies in the dimensions and composition of the reactor’s components as described in literature and implemented in current computational models. The identification, quantification, and to the extent possible, the minimization of these uncertainties is a crucial task on the path towards creating a benchmark based on the CNPS system. The present work seeks to initiate this process with a survey of available documentation and improvements to the accuracy of CNPS neutronics models.
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AGC Experiment Status
AGC Experiment Status to include: History and status of the AGC-4, Review of the AGC experiment, AGC graphite grades and samples, The AGC-4 capsule and specimens, and HDG-1 and HDG-2. Description of the Experiment, AGC graphite grades and samples, AGC-1 Test Train, AGC-1 Test Train Design Features, New AGC Irradiation Schedule (2018) and update in schedule, Irradiation and disassembly history, present and future status, and shipment and initial PIE. Initial PIE strategy, recovered specimens, PIE results, and status of experiments.
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Design and Analysis of a 250 MW Plate-Fuel Reactor for Nuclear Thermal Propulsion
Nuclear thermal propulsion (NTP) is a viable option for deep space missions due to its high thrust and lightweight system. Promising fuel options considered for the high operating temperature of NTP include graphite composite fuel and tungsten cermet fuel. As there remain uncertainties in the fabrication and performance of these two fuel elements, alternative designs using fuel plates and more typical fuels based on proven nuclear technologies are also being considered. This paper provides a summary of thermal hydraulic and neutronic analyses of the straight-plate fuel assembly design used for a 250 MW NTP core. The summary of technical findings from a series of scoping studies might be useful for future NTP engine designs.
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ZiaCore Critical Experiment Demonstrates Key Technologies for Nuclear Energy Systems
ZiaCore is a LANL Laboratory Directed Research and Development (LDRD) project focused on developing and demonstrating key technologies for future nuclear energy systems. The project itself was split into three tasks: 1) Design of the ZiaCore Reactor, a UO2 fueled, graphite and zirconium-hydride (ZrH) moderated, heat pipe cooled micro-reactor 2) Development of the ZrH and heat pipes components 3) Performance of a critical experiment with a representative portion of the ZiaCore reactor incorporating the ZrH and heat pipes developed and made at LANL.
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Ceramic Composite Inert Matrix Fuel Forms in High-Temperature Gas-Cooled Microreactors
Here, this work optimizes micro-prismatic high-temperature gas reactor (HTGR) designs to reduce the energy-normalized mass of spent nuclear fuel (SNF) and high-level waste (HLW) produced. The optimization was performed for the current graphite moderator and an inert matrix fuel (IMF) concept employing different composite moderators in a prismatic design architecture. The fuel matrix is magnesium oxide (MgO) with entrained tristructural-isotropic (TRISO) fuel. The moderator materials, including beryllium oxide (MgO-BeO) and beryllium (MgO-Be) at 40 vol % loading and yttrium hydride (MgO-YH x=1.9 ) and zirconium hydride (MgO-ZrH x=1.9 ) at 15 vol % loading, were entrained within the MgO host matrix. A generic graphite micro-prismatic HTGR is used as the baseline point design where the external dimensions are held constant. The composite moderator designs use 19.9% enriched uranium nitride TRISO fuel and hexagonal assemblies. For each IMF concept, an optimization study was performed to maximize the discharge burnup of the fuel by varying the TRISO packing fraction and the lattice pitch of the assemblies. The mass of SNF and HLW, other waste metrics, fuel cost, environmental impact metrics, and the activity of the SNF and HLW at 100 years and 100 000 years were calculated for the optimized IMF and graphite reference designs. The IMF results were subsequently compared to those of the graphite reference and the values for a light water reactor (LWR) and a small modular LWR. For the SNF and HLW, all the IMF concepts and the graphite reference produced less waste compared to the traditional LWR designs. However, the IMF concepts outperformed the graphite reference regarding the mass of SNF and HLW. For the other waste metrics, the IMF concepts showed reductions in fuel cost with improved environmental metrics relative to the graphite reference. Overall, the IMF concepts significantly reduced the SNF and HLW produced per unit of energy generated compared to traditional LWR designs.
TRISO↗
Purification Procedures for Single-Wall Carbon Nanotubes
This report summarizes the comparison of a variety of procedures used to purify carbon nanotubes. Carbon nanotube material is produced by the arc process and laser oven process. Most of the procedures are tested using laser-grown, single-wall nanotube (SWNT) material. The material is characterized at each step of the purification procedures by using different techniques including scanning electron microscopy (SEM), energy-dispersive X-ray spectroscopy (EDS), transmission electron microscopy (TEM), Raman, X-ray diffractometry (XRD), thermogravimetric analysis (TGA), nuclear magnetic resonance (NMR), and high-performance liquid chromatography (HPLC). The identified impurities are amorphous and graphitic carbon, catalyst particle aggregates, fullerenes, and hydrocarbons. Solvent extraction and low-temperature annealing are used to reduce the amount of volatile hydrocarbons and dissolve fullerenes. Metal catalysts and amorphous as well as graphitic carbon are oxidized by reflux in acids including HCl, HNO3 and HF and other oxidizers such as H2O2. High-temperature annealing in vacuum and in inert atmosphere helps to improve the quality of SWNTs by increasing crystallinity and reducing intercalation.
Gorelik, Olga P.↗
Design and Analysis of a 250 MW Plate-fuel Reactor for Nuclear Thermal Propulsion
Nuclear thermal propulsion (NTP) system is a viable option for deep space missions considering its high thrust and lightweight. Promising fuel options being considered for the high operating temperature of NTP include the graphite composite fuel and the tungsten cermet fuel. As there remains uncertainties in the fabrication and performance of these two fuel elements, alternative designs using fuel plates and more standard fuels that are based on proven and tested nuclear technologies are being considered. This paper provides a summary of thermal hydraulic and neutronic analyses of the straight fuel plate concept used for a 250 MW NTP core. The summary of technical findings from a series of scoping studies might be useful for the future NTP engine designs.
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Chicago Pile-1 paved the way for nuclear science and a lab in Los Alamos First self-sustaining nuclear chain reaction was nearly 80 years ago
On a bitter-cold winter day, 43 scientists gathered at an abandoned squash court at the University of Chicago where they would ultimately enable a secret lab in Los Alamos to change the world just years later. It was December 2, 1942. The group, led by Italian physicist and Nobel laureate Enrico Fermi, stacked graphite bricks, piling 57 layers that totaled more than 770,000 pounds. Later named Chicago Pile-1, their goal was to create the world’s first self-sustaining, controlled nuclear chain reaction. Inside the approximately 20-feet-tall pile were smaller blocks of uranium and control rods that, when removed, would cause the reaction to go critical – meaning create a nuclear chain reaction. It was roughly $1 million worth of materials, equivalent to nearly $16 million today, and a concept that a nuclear chain reaction would allow the weaponization of the atom. “Its success would be the crucial proof needed to know it would be possible to create an atomic bomb,” said LANL Historian Roger Meade (C-NR). “This was the precursor to the Lab we have today, nearly 80 years later.”
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Crosslinked Polyethyleneimine Gel Polymer Interface to Improve Cycling Stability of RFBs
Redox flow batteries are considered a promising technology for grid energy storage. However, capacity decay caused by crossover of active materials is a universal challenge for many flow battery systems, which are based on various chemistries. In this paper, using the vanadium redox flow battery as an example, we demonstrate a new gel polymer interface (GPI) consisting of crosslinked polyethyleneimine with a large amount of amino and carboxylic acid groups introduced between the positive electrode and the membrane. The GPI functions as a key component to prevent vanadium ions from crossing the membrane, thus supporting stable long-term cycling. Cyclic voltammetry (CV) and electrochemical impedance spectroscopy (EIS) measurements were conducted to investigate the effect of GPI on the electrochemical properties of graphitic carbon electrodes (GCFs) and redox reaction of catholyte. X-ray photoelectron spectroscopy (XPS) and 1 H nuclear magnetic resonance (NMR) spectra demonstrated that the crosslinked GPI is chemically stable for 100 cycles without dissolution of polymers and swelling in the strong acidic electrolytes. Results from inductively coupled plasma mass spectrometry (ICP-MS), Fourier-transform infrared (FTIR) spectroscopy, and energy-dispersive X-ray (EDX) spectroscopy proved that the GPI is effective in maintaining the concentration of vanadium species in their respective half-cells, resulting in improved cycling stability because of it prevents active species from crossing the membrane and stabilizes the oxidation states of active species.
Lim, Hyung-Seok↗
AGR-5/6/7 Test Train Disassembly & Tools Phases I & II
This document provides the process for qualification of the AGR-5/6/7 Test Train Disassembly equipment for remote operation in the HFEF Main Cell. The test train is composed of five welded stainless steel capsules that are expected to be transported in two sections. The first section will be Capsules 1 and 2. The second section will include Capsules 3, 4 and 5. The capsules range in length from about 9 inches to about 12 inches. Inside the capsules is a graphite holder that contains multiple fueled AGR compacts.
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Tritium Transport Phenomena in Molten-Salt Reactors
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. The models reasonably reproduced the MSRE data, but did so best when the permeability of structures was reduced by about a factor of 1,000. This issue does not appear to have been conclusively resolved, and all of the more recent attempts to model tritium transport in molten salts appear to make use of the same methodology. MSRE remains, however, essentially our only source of integral tritium transport data relevant to MSRs. Here, we generalize the MSRE approach to permeation in order to include potential rate-limiting effects at interfaces, as well the effects of added hydrogen. Appropriately non-dimensionalized, this system of equations identifies two dimensionless numbers whose relative values clearly delineate the conditions under which mass transport, surface effects, permeation, and hydrogen swamping are expected to become rate-limiting. We also describe the preliminary conceptual design of a forced convection FLiBe loop, into which tritium would be introduced for the purpose of providing validation data for such a model. The primary purpose of this is to investigate the coupled transport phenomena described above and identify those that are rate-limiting in MSRs. Additional test section configurations are described that would address other transport phenomena relevant to MSRs, including bubbly flows and graphite interactions.
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DOE ART-GCR Program Overview
To provide Advanced Reactor Technologies Gas-Cooled Reactor Program: Overview of ART-GCR and Graphite program objectives, status, and activities. Also Identify research areas and outcomes that will benefit stakeholders and clients (HTGR designers, suppliers, regulators, DOE-NE, etc.), and to Identify remaining R&D gaps and future needs.
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Advanced Reactor Development: Gas-Cooled Reactor First Quarter 2023 Report
ART Program First Quarter report for 2023. Entries for Fuels Development and Qualification, High Temperature Materials Development, Graphite Development and Qualification, and Design Methods Modeling and Validation.
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Preliminary Evaluation of Loading DOE Standard Canisters in the INL CPP-603 Irradiated Fuel Storage Facility - 20543
This paper looks at the equipment and operations necessary to load United States Department of Energy (DOE)-owned Spent Nuclear Fuel (SNF) into DOE Standard Canisters in the CPP-603 Irradiated Fuel Storage Facility (IFSF) in the Idaho Nuclear Technology and Engineering Center (INTEC) area at Idaho National Laboratory (INL). Two types of fuels are looked at in this evaluation: Advanced Test Reactor (ATR) fuel (uranium-aluminide fuel with aluminum cladding) and Peach Bottom fuel (thorium-uranium carbide fuel in a graphite matrix). The fuel ready for loading would come from fuel storage canisters in the CPP-603 facility. The paper describes the facility, the fuel types, the DOE Standard Canisters, and existing equipment; lists the needed loading operations; reviews facility features and equipment to perform the operations; and then lists the decisions, analyses, designs, demonstrations, and modifications that will be needed to perform the loading of DOE-owned SNF into DOE Standard Canisters in the CPP-603 IFSF. (authors)
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Porosity evolution in proton irradiated microfine-grained POCO graphite
This work investigated the porosity evolution of POCO ZXF-5Q graphite that has been irradiated by 340 kW, 120 GeV protons inside NT02 target system in Fermilab's NuMI beamline. This POCO graphite has undergone direct bulk dimensional swelling at low dose irradiation and its local microstructural change is still not well-understood during this process. In this work, the (sub-) micrometre scale porosity from six locations across proton beam fluence and temperature gradients have been studied using focused ion beam-scanning electron microscopy (FIB-SEM) tomography. Here, a deep learning-based tomographic image segmentation technique has been established and implemented for porosity segmentation and quantification. It has been found that there is a decrease in the total volumetric percentage of the porosity at proton beam centre (~ 8 – 8.4 vol.%), by comparing to un-irradiated POCO (~ 12 – 13vol.%) and to beam 2σ and 5σ radii (~ 12vol.%). This decrease in porosity volume percentage was found to be caused by the reduction in pores with volumes > 0.1 μm 3 induced by material bulk dimensional swelling at proton beam centre area. The porosity reduction in relation to dimensional change and irradiation creep was discussed among with other contributing factors, and further investigations through well-controlled irradiation experiment are still needed.
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Preliminary Neutronics Design and Analysis of the Fast Modular Reactor
General Atomics is developing a new 100-MW(thermal) fast modular reactor (FMR) that provides safe, carbon-free electricity and is capable of incremental capacity additions. The modular design allows it to be factory built and assembled onsite to keep the capital cost low, while the use of dry cooling facilitates siting to complement renewables in nearly any location. The FMR uses high-assay low-enriched uranium-dioxide fuel encapsulated by recognized irradiationresistant silicon carbide composite (SiGA®) cladding that is derisked in the current accident-tolerant fuel program. The FMR fuel assembly is a hexagonal fuel bundle of 120 fuel rods. The total length of the fuel assembly is less than 4 m, with an active fuel length of 1.8 m. The fuel assemblies are configured in an annular core that is located and supported by the reactor internals. The coolant material is helium at a normal operating pressure of 7 MPa. The core is surrounded by zirconium silicide (Zr 3 Si 2 ) and graphite reflector blocks. The fuel, coolant, internals, and reflectors are contained within a reactor pressure vessel. Here, the preliminary nuclear design and analysis established the arrangement of the active core and reflector blocks. The nuclear design analyses of the FMR defined the design parameters, such as fuel enrichments, excess reactivity, fueling scheme, fuel cycle, power distribution, and control rod worth. The preliminary conceptual design determined the three-batch fueling scheme with the allowable total power peaking factor of 1.5. The average discharge burnup is 100 GW days per ton of uranium.
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