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Vectorized Monte Carlo methods for reactor lattice analysis

Some of the new computational methods and equivalent mathematical representations of physics models used in the MCV code, a vectorized continuous-enery Monte Carlo code for use on the CYBER-205 computer are discussed. While the principal application of MCV is the neutronics analysis of repeating reactor lattices, the new methods used in MCV should be generally useful for vectorizing Monte Carlo for other applications. For background, a brief overview of the vector processing features of the CYBER-205 is included, followed by a discussion of the fundamentals of Monte Carlo vectorization. The physics models used in the MCV vectorized Monte Carlo code are then summarized. The new methods used in scattering analysis are presented along with details of several key, highly specialized computational routines. Finally, speedups relative to CDC-7600 scalar Monte Carlo are discussed.

Brown, F. B.↗

Molten Salt Lattice Confinement Fusion (LCF) Fast Fission Reactor for Lunar and Planetary Surface Power

Molten salt fission reactors (MSR) have been suggested for lunar and planetary surface power systems. They have the advantage of operating at high temperature, for efficient thermal-electrical conversion, low pressure, long-lived with high nuclear fuel burnup. MSR are often designed to breed fissile 233U from natural 232Th by neutron capture and  decay via: 232Th(n,)233Th(,)233Pa(,)233U Unfortunately, this process requires 233Pa isotope separation and segregation to decay to 233U. This requirement prevents additional neutron capture that interferes with 233U breeding. Instead, Wooley’s sub-critical, fast fission, molten salt reactor would use externally generated tokamak fusion neutrons1,2 to fission all actinides.We propose a simpler fusion-fast-fission sub-critical reactor that generates fast neutrons in situ from lattice confinement fusion (LCF) to fission fertile and fissile actinides. This hybrid reactor doesn't require enriched 235U fissile pins to initiate fis-sion reactions, nor 233Pa separation and segregation during operation. Like Wooley’s, this hybrid reactor “burns” natural uranium (238U) or thorium (232Th) which avoids uranium enrichment and additional fissile material launch safety and security costs. The LCF neutron source is initiated by bremsstrahlung photoneutrons (Fig. 1)3 or isotopic neutron sources in electron-screened lattices (Fig. 2)4,5. Alternatively, the electrolytic Pd-deuterium co-deposition6protocol fast fissions7 both 232Th and 238U. However, an aqueous electrolyte-based system, without pressurization similar to conventional pressurized water fission reactors, is incapable of high temperatures due to the boiling point of the electrolyte slightly over 100 C. Molten salts can be used instead as was demonstrated at the University of Hawaii8 using a variety of Ni and Pd cathodes in lithiated, hydrided and deuterated salts. These salts have melting points often exceeding 500C making them suitable to efficiently produce electrical power9 through either Advanced Stirling Genera-tors (< 100 kWe) or closed-Brayton Cycle (> 100 kWe). This hybrid reactor could power a wide range of lunar or Martian applications from unmanned in-struments, to charging vehicles and entire facilities such as human habitats or in situ resource utilization. The power conversion cycles are Carnot Cycle limited, but generally 30% efficient at best. However, waste heat on the moon or Mars is important to surviving either two-week lunar nights or Martian nights as well as providing process heat for mineral extraction and “living off the land”

Lawrence Forsley↗

Nuclear modules for space electric propulsion

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Difilippo, F. C.↗

Lattice Confinement Fusion & Fast Fission for Space

Lattice Confinement Fusion (LCF) 1 • Fusion of hydrogen isotopes into helium isotopes • No tokamak magnets or laser power supplies • Nuclear fuel is confined and triggered within lattice • Nuclear reactions enhanced by electron screening • Extended Electrodynamics (EED) role • NASA published results in Physical Review C 2,3 • Locally hot but globally cold • Commercialized in 2025 to produce medical radioisotopes 4 Lattice Confinement Fusion Fast-Fission Hybrid Reactor • Fusion neutrons fission uranium, spent fuel rods or thorium • No enriched uranium, cleaner fission! • Demonstrated with US Navy and GEC • Supported by NASA STMD (NIAC) and NSF Application • LEO, Deep Space Power and High I sp Propulsion • (Terrestrial)

lattice confinement fusion↗

Group-III Nitride Field Emitters

Field-emission devices (cold cathodes) having low electron affinities can be fabricated through lattice-mismatched epitaxial growth of nitrides of elements from group III of the periodic table. Field emission of electrons from solid surfaces is typically utilized in vacuum microelectronic devices, including some display devices. The present field-emission devices and the method of fabricating them were developed to satisfy needs to reduce the cost of fabricating field emitters, make them compatible with established techniques for deposition of and on silicon, and enable monolithic integration of field emitters with silicon-based driving circuitry. In fabricating a device of this type, one deposits a nitride of one or more group-III elements on a substrate of (111) silicon or other suitable material. One example of a suitable deposition process is chemical vapor deposition in a reactor that contains plasma generated by use of electron cyclotron resonance. Under properly chosen growth conditions, the large mismatch between the crystal lattices of the substrate and the nitride causes strains to accumulate in the growing nitride film, such that the associated stresses cause the film to crack. The cracks lie in planes parallel to the direction of growth, so that the growing nitride film becomes divided into microscopic growing single-crystal columns. The outer ends of the fully-grown columns can serve as field-emission tips. By virtue of their chemical compositions and crystalline structures, the columns have low work functions and high electrical conductivities, both of which are desirable for field emission of electrons. From examination of transmission electron micrographs of a prototype device, the average column width was determined to be about 100 nm and the sharpness of the tips was determined to be characterized by a dimension somewhat less than 100 nm. The areal density of the columns was found to about 5 x 10(exp 9)/sq cm . about 4 to 5 orders of magnitude greater than the areal density of tips in prior field-emission devices. The electric field necessary to turn on the emission current and the current per tip in this device are both lower than in prior field-emission devices, such that it becomes possible to achieve longer operational lifetime. Moreover, notwithstanding the lower current per tip, because of the greater areal density of tips, it becomes possible to achieve greater current density averaged over the cathode area. The thickness of the grown nitride film (equivalently, the length of the columns) could lie between about 0.5 microns and a few microns; in any event, a thickness of about 1 micron is sufficient and costs less than do greater thicknesses. It may be possible to grow nitride emitter columns on glass or other substrate materials that cost less than silicon does. What is important in the choice of substrate material is the difference between the substrate and nitride crystalline structures. Inasmuch as the deposition process is nondestructive, an ability to grow emitter columns on a variety of materials would be advantageous in that it would facilitate the integration of field-emitter structures onto previously processed integrated circuits.

Bensaoula, Abdelhak↗

Accessing Icy Worlds Using Lattice Confinement Fusion (LCF) Fast Fission

Exploring the oceans of icy moons and planets requires a robust robotic probe powered by an energy source that can operate under icy world ocean conditions and be self-contained. We propose a non-fissile, compact, scalable nuclear energy source to electrically power untethered, autonomous probes to melt or bore through icy world crusts. The probe can be used for planetary (i.e. Pluto, lunar (i.e., Europa), or asteroid exploration (i.e. Ceres) where ice caps are encountered. This new approach may yield a variable output fission power source with a higher performance than 238Pu and a non-fissile alternative to a highly enriched uranium (HEU) core. This approach saves uranium enrichment expense, and both HEU and 238Pu security and launch safety costs. The reactor will produce electrical power with thermal waste heat to melt through the ice crust with possible sonic assistance. This NIAC funded proposal and presentation will introduce the hybrid fusion-fission reactor concept applied to the Saturnian moon, Enceladus, explore the potential for and benefits of this new type of nuclear reactor, and discuss applying this new technology.

Icy Worlds↗

Heat Pipe Heat Exchanger for Nuclear Electric Propulsion Power Conversion System

Heat pipe reactors have been considered by the Space Nuclear Propulsion program for Nuclear Electric Propulsion (NEP) power conversion systems and will require the use of heat exchangers to transfer heat via heat pipes to the Brayton working fluid from the reactor. Sodium (Na) and lithium (Li) were considered as viable working fluids inside the heat pipes which were assumed to have the same geometry based on studies and information from the Los Alamos National Laboratory. The heat exchanger was assumed to be a rectangular duct with heat pipes serving as tubes from previous NEP work and recommendations. Based on this geometry, Zukauskas correlations were used to model the convective heat transfer and pressure losses. Parametric sizing of the reactor component involved operational limits-based heat pipe thermal hydraulic modeling in cohesion with required user input geometry for the in-core lattice and various subcomponents. This work considered various power conversion inlet temperatures (PCIT) of 1100 K, 1150 K, and 1200 K for Na heat pipes and 1100 K, 1150 K, 1200 K, and 1400 K for Li heat pipes based on recommendations from prior work. Using these different PCITs, the subsystem masses and pressure losses were determined and analyzed. Na showed a lower overall operating temperature and about a fifth of the maximum heat throughput capability than that of Li for the same geometry. Due to this, the entire Na-based subsystem ended up being three times more massive than the Li-based subsystem given five times the required number of heat pipes. At the low PCIT of 1100 K, the Na-based subsystem exhibited the lowest pressure losses given the large overall cross sectional flow area and relatively low frictional pressure losses. However, as the PCIT increased, the frictional pressure losses increased resulting in higher pressure losses at the 1200 K PCIT than Li-based subsystem. However, the Li-based subsystem exhibited the largest pressure losses of all analyzed cases at the 1400 K PCIT due to the low density of the Brayton working fluid at this temperature.

electric↗

Atmospheric pressure organometallic vapor-phase epitaxial growth of (Al/x/Ga/1-x/)0.51In0.49P (x from 0 to 1) using trimethylalkyls

This paper describes growth of (Al/x/Ga/1-x)0.51In0.49P layers (with x from 0 to 1) lattice-matched to (001)-oriented GaAs substrates by atmospheric-pressure OMVPE, using trimethylindium, trimethylaluminum, and trimethylgallium and PH3 as source materials in a horizontal reactor. Excellent surface morphologies were obtained over the entire range of Al compositions at a growth temperature of 680 C. Photoluminescence (PL) was observed for all samples with x values not below 0.52, with PL peak energies as high as 2.212 eV. The PL FWHM for Ga(0.51)In(0.49)P was 7.2 meV at 10 K and 35 meV at 300 K. At 10 K, the PL intensity was nearly a constant over the composition range from x = 0 to 0.52.

Cao, D. S.↗

Lattice Confinement Fusion and Fusion-Fast-Fission Energy Source Development

NASA and the US Navy have demonstrated Lattice Confinement Fusion (LCF) and the Fusion-Fast-Fission of natural uranium and thorium. Both methods build upon decades of research, and the latter benefits from conventional nuclear fission and fusion results. However, this nascent technology doesn’t require fissile isotopes and avoids nuclear weapons proliferation concerns from uranium 235U isotopic enrichment or plutonium 239Pu separation. Of particular note, LCF doesn’t need power hungry magnets, lasers, or particle beams. Potentially, it could provide watts to hundreds of kilowatts of electrical power and process heat suitable for space power and high Isp nuclear electric propulsion (NEP) as well as distributed terrestrial power. Finally, there are indications that LCF fast-fission products are more benign than those of conventional fission reactors. Our goal is to scale the reactions and increase the power output through higher temperature operation and increased material masses.

lattice confinement fusion↗

Modeling of Macroscopic/Microscopic Transport and Growth Phenomena in Zeolite Crystal Solutions Under Microgravity Conditions

Crystals grown from liquid solutions have important industrial applications. Zeolites, for instance, a class of crystalline aluminosilicate materials, form the backbone of the chemical process industry worldwide, as they are used as adsorbents and catalysts. Many of the phenomena associated with crystal growth processes are not well understood due to complex microscopic and macroscopic interactions. Microgravity could help elucidate these phenomena and allow the control of defect locations, concentration, as well as size of crystals. Microgravity in an orbiting spacecraft could help isolate the possible effects of natural convection (which affects defect formation) and minimize sedimentation. In addition, crystals will stay essentially suspended in the nutrient pool under a diffusion-limited growth condition. This is expected to promote larger crystals by allowing a longer residence time in a high-concentration nutrient field. Among other factors, the crystal size distribution depends on the nucleation rate and crystallization. These two are also related to the "gel" polymerization/depolymerization rate. Macroscopic bulk mass and flow transport and especially gravity, force the crystals down to the bottom of the reactor, thus forming a sedimentation layer. In this layer, the growth rate of the crystals slows down as crystals compete for a limited amount of nutrients. The macroscopic transport phenomena under certain conditions can, however, enhance the nutrient supply and therefore, accelerate crystal growth. Several zeolite experiments have been performed in space with mixed results. The results from our laboratory have indicated an enhancement in size of 30 to 70 percent compared to the best ground based controls, and a reduction of lattice defects in many of the space grown crystals. Such experiments are difficult to interpret, and cannot be easily used to derive empirical or other laws since many physical parameters are simultaneously involved in the process. At the same time, however, there is increased urgency to develop such an understanding in order to more accurately quantify the process. In order to better understand the results obtained from our prior space experiments, and design future experiments, a detailed fluid dynamic model simulating the crystal growth mechanism is required. This will not only add to the fundamental knowledge on the crystallization of zeolites, but also be useful in predicting the limits of size and growth of these important industrial materials. Our objective is to develop macro/microscopic theoretical and computational models to study the effect of transport phenomena in the growth of crystals grown in solutions. Our effort has concentrated so far in the development of separate macroscopic and microscopic models. The major highlights of our accomplishments are described.

Gatsonis, Nikos A.↗

Statistical Models of Fracture Relevant to Nuclear-Grade Graphite: Review and Recommendations

The nuclear-grade (low-impurity) graphite needed for the fuel element and moderator material for next-generation (Gen IV) reactors displays large scatter in strength and a nonlinear stress-strain response from damage accumulation. This response can be characterized as quasi-brittle. In this expanded review, relevant statistical failure models for various brittle and quasi-brittle material systems are discussed with regard to strength distribution, size effect, multiaxial strength, and damage accumulation. This includes descriptions of the Weibull, Batdorf, and Burchell models as well as models that describe the strength response of composite materials, which involves distributed damage. Results from lattice simulations are included for a physics-based description of material breakdown. Consideration is given to the predicted transition between brittle and quasi-brittle damage behavior versus the density of damage (level of disorder) within the material system. The literature indicates that weakest-link-based failure modeling approaches appear to be reasonably robust in that they can be applied to materials that display distributed damage, provided that the level of disorder in the material is not too large. The Weibull distribution is argued to be the most appropriate statistical distribution to model the stochastic-strength response of graphite.

Nemeth, Noel N.↗

Accessing Icy World Oceans using Lattice Confinement Fusion Fast Fission

Exploring the oceans of icy moons and planets requires a robust robotic probe powered by an energy source that can operate under icy world ocean conditions and be self-contained. We propose a non-fissile, compact, scalable nuclear energy source to electrically power untethered, autonomous probes to melt or bore through icy world crusts. The probe can be used for planetary (i.e. Pluto, lunar (i.e., Europa), or asteroid exploration (i.e. Ceres) where ice caps are encountered. This new approach may yield a variable output fission power source with a higher performance than 238Pu and a non-fissile alternative to a highly enriched uranium (HEU) core. This approach saves uranium enrichment expense, and both HEU and 238Pu security and launch safety costs. The reactor will produce electrical power with thermal waste heat to melt through the ice crust with possible sonic assistance. This paper will introduce the hybrid fusion-fission reactor concept, explore the potential for and benefits of this new type of nuclear reactor, and discuss the technical approach to developing this new technology.

Theresa L Benyo↗

Development And Testing Of The Inertial Electrostatic Confinement Diffusion Thruster

The Inertial Electrostatic Confinement (IEC) diffusion thruster is an experiment in active development that takes advantage of physical phenomenon that occurs during operation of an IEC device. The IEC device has been proposed as a fusion reactor design that relies on traditional electrostatic ion acceleration and is typically arranged in a spherical geometry. The design incorporates two radially-symmetric spherical electrodes. Often the inner electrode utilizes a grid of wire shaped in a sphere with a radius 15 to 50 percent of the radius of the outer electrode. The inner electrode traditionally has 90 percent or more transparency to allow particles (ions) to pass to the center of the spheres and collide/recombine in the dense plasma core at r=0. When operating the IEC, an unsteady plasma leak is typically observed passing out one of the gaps in the lattice grid of the inner electrode. The IED diffusion thruster is based upon the idea that this plasma leak can be used for propulsive purposes. The IEC diffusion thruster utilizes the radial symmetry found in the IEC device. A cylindrical configuration is employed here as it will produce a dense core of plasma the length of the cylindrical grid while promoting the plasma leak to exhaust through an electromagnetic nozzle at one end of the apparatus. A proof-of-concept IEC diffusion thruster is operational and under testing using argon as propellant (Figure 1).

Becnel, Mark D.↗

Fabrication of cermet bearings for the control system of a high temperature lithium cooled nuclear reactor

The techniques used to fabricate cermet bearings for the fueled control drums of a liquid metal cooled reference-design reactor concept are presented. The bearings were designed for operation in lithium for as long as 5 years at temperatures to 1205 C. Two sets of bearings were fabricated from a hafnium carbide - 8-wt. % molybdenum - 2-wt. % niobium carbide cermet, and two sets were fabricated from a hafnium nitride - 10-wt. % tungsten cermet. Procedures were developed for synthesizing the material in high purity inert-atmosphere glove boxes to minimize oxygen content in order to enhance corrosion resistance. Techniques were developed for pressing cylindrical billets to conserve materials and to reduce machining requirements. Finishing was accomplished by a combination of diamond grinding, electrodischarge machining, and diamond lapping. Samples were characterized in respect to composition, impurity level, lattice parameter, microstructure and density.

Yacobucci, H. G.↗