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At least 91 records · Page 5

Superionic-like diffusion in yttrium dihydride

For the next-generation high temperature microreactors, yttrium dihydride (YH 2 ) is an attractive solid state neutron moderator. Despite a number of recent investigations, the mechanism of hydrogen transport remains poorly understood. Experimental evaluations of diffusivity are inconclusive with large variations in diffusivities and activation energies. In this work, we perform ab initio molecular dynamics (AIMD) simulations on YH 2 for temperatures spanning 300 K to 1200 K. Our main finding is that YH 2 shows a superionic-like behavior with hydrogen atoms hopping from one native site to another above a characteristic temperature of 800 K. This correlated motion results in quasi-one-dimensional string-like displacements that enable the hydrogen atoms to diffuse rapidly. We confirm that the octahedral sites are mostly unoccupied, although channeling through them is the most favored pathway between lattice hops above 800 K. At the highest temperature of 1200 K, the string relaxation time is merely of the order of a few picoseconds, which indicates a liquid-like diffusive behavior. Based on the formation of spontaneous thermal vacancies, an order-disorder crossover temperature T α ~ 800 K is established for YH 2 with an activation energy of 0.83 eV for hydrogen diffusion in the superionic-like state.

Superionic-like Diffusion↗

Investigation of High-Temperature Compatibility of Select Oxides with Yttrium Hydride for Micro and Space Reactor Applications

Yttrium hydride is a promising material for a high-temperature neutron moderator in advanced micro and space reactors due to its high hydrogen density and relative thermal stability compared to other metal hydrides. However, yttrium hydride desorbs hydrogen rapidly at temperatures above 800°C, which is below the operational temperature range of some reactor designs. A hydrogen barrier coating of oxide on the hydride surface may inhibit hydrogen loss at 800°C and beyond, but the high-temperature compatibility between yttrium hydride and many oxides is currently unknown. The high-temperature compatibility of Al 2 O 3 , MgO, and Y 2 O 3 with YH 1.92 was examined by subjecting mixed oxide–hydride pellets to a 1200°C heat treatment then using a combination of diffractometry, microscopy, and spectroscopy to determine changes in the pellet composition as a result. Yttrium scavenged oxygen from both Al 2 O 3 and MgO to form Y 2 O 3 , resulting in significant loss of YH 1.92 . Yttrium reacted with reduced aluminum to form YAl 2 and several other compounds. Reduced magnesium volatilized above 1091°C and vacated the pellet. Only Y 2 O 3 did not appreciably react with YH 1.92 . Of the three oxides tested, only Y 2 O 3 was compatible with YH 1.92 at 1200°C based on the experimental criteria.

compatibility↗

Advanced High-Temperature Sodium-Cooled Thermal Reactors Using Less Than 10% Enriched UO 2 Fuel

Here, this paper presents a 1200-MW(thermal) advanced sodium-cooled thermal reactor concept that uses online refueling of 3.5% to 9.95% enriched UO2 fuel pin bundles; uses either graphite or beryllium oxide (BeO) as a neutron moderator; reaches outlet temperatures of 650°C enabling a thermal efficiency of at least 45%; has a high specific power of 133 W/g U; has average power densities of 16.4 and 43.2 W/cm 3 with graphite and BeO, respectively; reaches an average discharge burnup of 100 MWd/kg U; and generates 52% less spent fuel volume, 28% less fission products, and 47% to 64% less transuranics than a typical large pressurized water reactor for the same amount of electricity produced.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

Coupling between TOUGH3 and FLAC3D

Sandia National Laboratories has hired Itasca Consulting Group, Inc., the authors of the FLAC3D geomechanics software, to couple FLAC3D with TOUGH3, the porous media flow solver. The work is being done to enable a coupled mechanical-thermal-hydraulic analysis of a potential criticality event in a dual purpose cannister (DPC). The U.S. Department of Energy Office of Spent Fuel and Waste Science & Technology is investigating the performance of DPCs for direct geological disposal of spent nuclear fuel. Post closure criticality control is an important aspect of this investigation. Over geological timescales, it is envisioned that the canister and canister overpack will develop fractures due to stress corrosion processes. A breach in the canister could allow groundwater to fill the canister. Fresh water is a neutron moderator; thus, if the canister internals and fuel assemblies have been sufficiently degraded, a criticality event could occur. Such an event would release enough energy to boil the water between the fuel rods and pressurize the cannister. This internal pressurization may cause the initial fractures in the canister and overpack to grow. It is important to understand the change in hydraulic transmissivity between the canister and surroundings for two reasons: first, because it may control the potential for and frequency of subsequent criticality events; second, because it will control the release of radionuclides from the canister. The motivation for this work is to better understand the potential for periodic criticality events, cannister damage, and release of radionuclides during a criticality event in a DPC.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗

A Brief Introduction to Gamma and Neutron Radiation [Slides]

Objectives: Describe gamma rays in plain English; Give 3 reasons we measure gamma rays; Describe shielding effect vs gamma energy; List 4 components of the natural background; Detail the main reason we care about neutrons; and List three good neutron moderators.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗

Beryllium Oxide (BeO) Handbook

Beryllium oxide (BeO) is an oxide ceramic material. BeO ceramics provide thermal conductivity second only to diamond among electrically insulating materials, dissipating nearly 300 W/mK at room temperature. This property leads to electronic applications for BeO. BeO has a thermal expansion coefficient intermediate between gallium arsenide and refractory metal composites. Its low dielectric constant (6.7) and low loss index (0.0012 at 1 MHz) permits electrical circuit performance at high frequencies. It is employed as a window material for high frequency microwave applications. BeO is very stable under oxidizing and reducing high temperature and high humidity environments, unlike nitrides which will decompose to their oxide equivalent. Similar to beryllium metal, BeO acts as a scatterer of neutrons and is used as a neutron moderator material in nuclear reactor applications. The largest use of BeO is as the raw material for the manufacture of beryllium-copper alloys, followed closely as the raw material for conversion to beryllium metal. Most of the research performed on BeO was conducted in the 1960’s and 1970’s. 75% of the references cited in this Handbook are from the time period 1960-1980. Up to 1995, there were three commercial sources of BeO: Materion Corporation, Eagle-Picher, and Ceradyne Corporation. However, since 1995 Eagle-Picher and Ceradyne no longer supply BeO commercial products. Thus, the only current commercial source of BeO materials is Materion Corporation.

36 MATERIALS SCIENCE↗

Xenon-Doped Liquid Argon TPCs as a Neutrinoless Double Beta Decay Platform

Next-generation large liquid argon time-projection chambers offer an unprecedented amount of active detector mass in a deep location. Modifications to the detector design could enable neutrinoless double beta decay searches, with the possibility of reaching the normal ordering region. These modifications include adding external neutron moderation, filling the detector with argon depleted in Ar42, doping with xenon, and a method to achieve percent-level energy resolution at the MeV scale. One way to achieve this desired level of energy resolution is to introduce a photosensitive dopant into the argon, converting the isotropic scintillation photons into a directional ionization signal. This would enhance the achievable energy resolution in large LArTPCs, and would lead to sizeable improvements to many facets of the broad physics program such a detector could offer.

72 PHYSICS OF ELEMENTARY PARTICLES AND FIELDS↗

Assessing the Impact of Large Removable Beryllium Reflector Experiments on HFIR Performance and Safety Metrics

This study investigated the effect of various removable beryllium (RB) experiment configurations on High Flux Isotope Reactor (HFIR) metrics to address increasing interest in these facilities for materials and fuels irradiation research. The RB reflector contains eight large and four small irradiation experiment facilities that offer excellent neutron flux conditions to perform fission and fusion reactor materials and fuels irradiation research. This work aims to outline acceptable RB configurations based on safety, performance, and programmatic metrics. This study assessed experiment effects on reactivity, cycle length, fission rate density distributions, and neutron flux distributions using the Shift, HFIRCON, and SCALE ORIGEN codes. Initial evaluations focused on generic experiment materials including aluminum plugs, stainless steel plugs, molybdenum plugs, and aluminum plugs with gadolinium shields. Subsequent analyses of MiniFuel experiments were performed to evaluate a heterogenous experiment, consisting of a more complex geometry and bearing several materials, and to test the correlations developed with the generic materials on a real experiment. Furthermore, the effects of neutron poison concentrations in the standard beryllium plugs were evaluated. When assuming a reference RB configuration with eight large fresh beryllium plugs, perturbed configurations with three aluminum plugs, one stainless steel plug, one molybdenum plug, one gadolinium-shielded aluminum plug, and one MiniFuel experiment resulted in a cycle length reduction of less than 1.3 days, the current threshold before requiring additional approvals. Configurations with five aluminum plugs, one stainless steel plug, one molybdenum plug, one gadolinium-shielded aluminum plug, and two MiniFuel experiments meet the 1.3 day limit if irradiated beryllium plugs are considered. Fuel element fission rate density distributions remained within safety limits, with maximum local increases under 9%. Neutron flux calculations revealed large thermal flux depressions inside and around the perturbed RB facilities, while epithermal and fast neutron fluxes increased because of reduced neutron moderation by the perturbed materials. These findings provide valuable guidance for optimizing RB configurations to balance safety and performance at HFIR.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗

Non-destructive structural characterization of graphite components using mechanical resonance and deep learning

As compared to conventional nuclear reactors, microreactors have the potential to significantly reduce construction timelines and capital costs, decreasing the barriers for advanced nuclear reactor technologies. However, the lower power output of these microreactors (typically < 20 MWe) creates challenging economics if operation and maintenance costs cannot be sufficiently reduced. The compact size of these designs presents an opportunity for comprehensive in-situ structural health monitoring to provide real-time feedback in order to reduce operational costs associated with maintenance and downtime. Many microreactor concepts use graphite for both in-core neutron moderation and as a structural material, which has typically required some form of periodic and laborious inspection. This report provides a description and assessment of recent work with graphite to couple acoustic-based experimental measurements and characterization with machine learning models to mature structural health monitoring capabilities and generate benefits for the nuclear microreactor industry. With resilient embedded sensors in development in other programs funded by the US Department of Energy’s Office of Nuclear Energy and elsewhere, the work described herein builds upon previously funded efforts to mature non-destructive testing technology that relates measured vibrational signatures to structural changes, using a combination of new experimental measurements and machine learning processing. Building on past successful demonstrations of predictive workflows to identify structural changes in a hexagonal stainless steel test article with excellent acoustic propagation, we first performed baseline characterization on graphite samples with canonical geometries to ensure compatibility and confidence in the applied techniques for a material with distinctly different mechanical properties. In contrast to efforts in previous years, we worked exclusively with unidirectional vibration data that is more comparable to those expected from the existing embedded sensor technologies which are suitable for deployment in a reactor setting. Established acoustic and modern machine-learning-based characterization approaches were applied to the resulting datasets from these simple geometries. Both approaches were found to be highly capable of detecting even small geometric irregularities amongst nominally identical samples. As such, we then moved to testing these approaches for detection of artificial local stress perturbations introduced into a more complex geometry: a hexagonal block with drilled holes. A main outcome of this work is that a generalizable ML workflow can be used to detect and predict the characteristics of small artificial anomalies in a graphite component with a relevant geometry. While this work was performed using surficial vibration data, we expect the approach to be flexible and viable for other monitoring scenarios, such as those with different arrangements or types of sensor arrays. As compared to previously funded efforts, an existing ML workflow based on neural networks was enhanced through the addition of recently developed Fourier neural operators. As applied to previously collected and new vibration datasets, prediction accuracies of anomaly characterizations were greatly improved with minimal added computational cost. As trained on small durations of vibration data (tens of seconds) collected over a realistic number of locations, the model was able to reliably determine the presence of a subtle stress anomaly and begin to provide location estimates. Such an approach is likely to be viable for more relevant reactor damage scenarios for graphite components, such as progressive crack growth or creep.

36 MATERIALS SCIENCE↗

Fabrication of neutron absorbing metal hydride entrained ceramic matrix shield composites

With significant improvement in High Temperature Superconductors (HTS), several projects are adopting HTS technology for fusion power systems. Compact HTS tokamaks offer potential advantages including lower plant costs, enhanced plasma control, and ultimately lower cost of electricity. However, as compact reactors have a reduced radial build to accommodate shielding, HTS degradation due to radiation damage or heating is a significant and potentially design limiting issue. Shielding must mitigate threats to the superconducting coils: neutron cascade damage, heat deposition and potentially organic insulator damage due x-rays. Unfortunately, there are currently no hi-performance shielding materials to enable the potential performance enhancement offered by HTS. In this work, we present a manufacturing method to fabricate a new class of composite shields that are high performance, high operating temperature, and simultaneously neutron absorbing and neutron moderating. The composite design consists of an entrained metal-hydride phase within a radiation stable MgO ceramic host matrix. We discuss the fabrication, characterization, and thermophysical performance data for a series of down-selected composite materials inspired by future fusion core designs and their operational performance metrics. To our knowledge these materials represent the first ceramic composite shield materials containing significant metal hydrides.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗

Real-Time Observation of Nanoscale Kink Band Mediated Plasticity in Ion-Irradiated Graphite: An In Situ TEM Study

Graphite IG-110 is a synthetic polycrystalline material used as a neutron moderator in reactors. Graphite is inherently brittle and is known to exhibit a further increase in brittleness due to radiation damage at room temperature. To understand the irradiation effects on pre-existing defects and their overall influence on external load, micropillar compression tests were performed using in situ nanoindentation in the Transmission Electron Microscopy (TEM) for both pristine and ion-irradiated samples. While pristine specimens showed brittle and subsequent catastrophic failure, the 2.8 MeV Au2+ ion (fluence of 4.378 × 1014 cm−2) irradiated specimens sustained extensive plasticity at room temperature without failure. In situ TEM characterization showed nucleation of nanoscale kink band structures at numerous sites, where the localized plasticity appeared to close the defects and cracks while allowing large average strain. We propose that compressive mechanical stress due to dimensional change during ion irradiation transforms buckled basal layers in graphite into kink bands. The externally applied load during the micropillar tests proliferates the nucleation and motion of kink bands to accommodate the large plastic strain. The inherent non-uniformity of graphite microstructure promotes such strain localization, making kink bands the predominant mechanism behind unprecedented toughness in an otherwise brittle material.

36 MATERIALS SCIENCE↗

Nuclear pumped laser modeling

A model for an indirectly nuclear pumped laser experiment was developed. The apparatus consists of the usual cylindrical neutron moderator enclosing a gas filled quartz cell. The quartz cell in this case has two concentric compartments; the larger output compartment is filled with He-3 and fluorescent gases which are coupled radiatively to the inner compartment which is filled with a lasing gas for efficient optical pumping. The fluorescent gas in this study was a mixture of Ar, Kr and F2. Perfluoroalkyliodide was chosen as the lasant material of the inner compartment. The degree of photoabsorption in the lasant gas was calculated. The kinetic efficiencies of various gas combinations in the outer fluorescence tube were then evaluated in order to maximize the fluorescence output in the iodine pump band.

Wilson, J. W.↗

Functionally Graded Nanophase Beryllium/Carbon Composites

Beryllium, beryllium alloys, beryllium carbide, and carbon are the ingredients of a class of nanophase Be/Be2C/C composite materials that can be formulated and functionally graded to suit a variety of applications. In a typical case, such a composite consists of a first layer of either pure beryllium or a beryllium alloy, a second layer of B2C, and a third layer of nanophase sintered carbon derived from fullerenes and nanotubes. The three layers are interconnected through interpenetrating spongelike structures. These Be/Be2C/C composite materials are similar to Co/WC/diamond functionally graded composite materials, except that (1) W and Co are replaced by Be and alloys thereof and (2) diamond is replaced by sintered carbon derived from fullerenes and nanotubes. (Optionally, one could form a Be/Be2C/diamond composite.) Because Be is lighter than W and Co, the present Be/Be2C/C composites weigh less than do the corresponding Co/WC/diamond composites. The nanophase carbon is almost as hard as diamond. WC/Co is the toughest material. It is widely used for drilling, digging, and machining. However, the fact that W is a heavy element (that is, has high atomic mass and mass density) makes W unattractive for applications in which weight is a severe disadvantage. Be is the lightest tough element, but its toughness is less than that of WC/Co alloy. Be strengthened by nanophase carbon is much tougher than pure or alloy Be. The nanophase carbon has an unsurpassed strength-to-weight ratio. The Be/Be2C/C composite materials are especially attractive for terrestrial and aerospace applications in which there are requirements for light weight along with the high strength and toughness of the denser Co/WC/diamond materials. These materials could be incorporated into diverse components, including cutting tools, bearings, rocket nozzles, and shields. Moreover, because Be and C are effective as neutron moderators, Be/Be2C/C composites could be attractive for some nuclear applications.

Voronov, Oleg A.↗

Functionally Graded Nanophase Beryllium/Carbon Composites

Beryllium, beryllium alloys, beryllium carbide, and carbon are the ingredients of a class of nanophase Be/Be2C/C composite materials that can be formulated and functionally graded to suit a variety of applications. In a typical case, such a composite consists of a first layer of either pure beryllium or a beryllium alloy, a second layer of B2C, and a third layer of nanophase sintered carbon derived from fullerenes and nanotubes. The three layers are interconnected through interpenetrating spongelike structures. These Be/Be2C/C composite materials are similar to Co/WC/diamond functionally graded composite materials, except that (1) W and Co are replaced by Be and alloys thereof and (2) diamond is replaced by sintered carbon derived from fullerenes and nanotubes. (Optionally, one could form a Be/Be2C/diamond composite.) Because Be is lighter than W and Co, the present Be/Be2C/C composites weigh less than do the corresponding Co/WC/diamond composites. The nanophase carbon is almost as hard as diamond. WC/Co is the toughest material. It is widely used for drilling, digging, and machining. However, the fact that W is a heavy element (that is, has high atomic mass and mass density) makes W unattractive for applications in which weight is a severe disadvantage. Be is the lightest tough element, but its toughness is less than that of WC/Co alloy. Be strengthened by nanophase carbon is much tougher than pure or alloy Be. The nanophase carbon has an unsurpassed strength-to-weight ratio. The Be/Be2C/C composite materials are especially attractive for terrestrial and aerospace applications in which there are requirements for light weight along with the high strength and toughness of the denser Co/WC/diamond materials. These materials could be incorporated into diverse components, including cutting tools, bearings, rocket nozzles, and shields. Moreover, because Be and C are effective as neutron moderators, Be/Be2C/C composites could be attractive for some nuclear applications.

Choi, Michael K.↗

Overview of High Temperature Material Needs for Space Nuclear Propulsion Reactors

Space nuclear propulsion systems are capable of enabling future crewed missions to Mars. Two primary options of these systems exist: nuclear thermal propulsion (NTP) and nuclear electric propulsion (NEP). NTP systems use a reactor as a heat exchanger to heat a hydrogen propellant to produce high specific impulse (≥ 900 s) and thrust (10 - 100 klbf). A NTP materials challenge is development of nuclear fuel capable of operating for short durations (hours) under high power densities (≥ 5 MW/L) and temperatures (> 2700 K) in a corrosive hydrogen environment. In NEP systems, the reactor heats a working fluid which transfers heat to a power conversion system to power electric thrusters. NEP is capable of higher specific impulse (≥ 2000 s) but limited to low thrust (10s lbf). Light weight vessel and in-core structural materials that are radiation-resistant, creep-resistant, and dimensionally stable while operating at ~1500 K over mission times up to several years are needed for NEP. Both systems can benefit from high temperature neutron moderators. Through the Space Nuclear Propulsion project, NASA, supported by the DOE, is developing NTP and NEP systems. Technology maturation planning and risk reduction activities related to fabrication and testing of reactor material candidates is ongoing. This presentation overviews NTP and NEP materials needs and current ceramic and composite development activities.

nuclear thermal propulsion↗

Apparatus and method for generating medical isotopes

An apparatus for generating medical isotopes provides an annular fissile solution vessel surrounding a neutron generator. The annular fissile solution vessel provides for good capture of the emitted neutrons and a geometry that provides enhanced stability in an aqueous reactor. A neutron multiplier and/or a neutron moderator may be used to improve the efficiency and control the criticality of the reaction in the annular fissile solution vessel.

Piefer, Gregory Richard↗

Need for research and training reactors for advanced reactor designs

Full text of publication follows. Research and training reactors have served a valuable role in helping train workforce for currently operating fleet of light water reactors. These research and training reactors have been used in reactor laboratory classes to familiarize the students with such vital concepts as approach to criticality, reactor period, neutron moderation, reactivity, flux distribution and leakage, etc. As the industry moves toward advanced non-light-water reactor designs, it is critical that research and training reactors be developed and deployed at university campuses to help train the new generation of nuclear and non-nuclear engineers who are likely to design, build, and operate these advanced reactors. Among the designs currently being pursued for nuclear power generation include molten salt, sodium cooled, and gas cooled designs, with options for various fuel forms. Thus, industry and DOE in collaboration with academic institutions should devise plans on how to familiarize the next generation of nuclear workforce with hands-on experience necessary for such designs. These research and training reactors will play a vital role in familiarizing the future workforce with hands-on experience with concepts associated with fast spectrum reactors, gas cooled reactors, and other features not associated with light water reactors. In addition to classical nuclear engineering concepts, these advanced research and training reactors can also be used for hands-on training as well as for research on features being considered in the design of GEN-IV reactors: cyber security for digital control room operations, hybrid energy system, hydrogen generation, district heating, autonomous control... (author)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗