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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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All sourcesNASA NTRSDOE OSTI
DOE OSTI2024

First operation and validation of simulations for the divertor cryo-vacuum pump in Wendelstein 7-X

Ten cryo-vacuum pumps (CVPs) were installed in the subdivertor region of each island divertor in the stellarator Wendelstein 7-X (W7-X) and operated for the first time during the recently completed plasma campaign OP2.1. A pumping speed of 70 ± 1 $\frac{m^3}{s}$ was measured during dedicated tests with known hydrogen gas injection. Based on a conductance model, the estimated pumping speed ranges from 86-93 $\frac{m^3}{s}$ for different sticking coefficients between 0.6 and 0.8. After completion of the initial tests the CVPs were operated successfully throughout the campaign, with regeneration performed once a week. Neutral gas pressures in the subdivertor in the range of 10 −4 mbar are well within the molecular flow regime and limit the particle exhaust capabilities of the CVPs. Simulations of the neutral gas pressure in the three-dimensional complex geometry of the subdivertor were performed using the DIVGAS code based on the direct simulation Monte Carlo method and a model implemented in the steady-state thermal package in ANSYS, which are in agreement with the measured values during plasma operation.

Cryo-vacuum pumping↗
DOE OSTI2024

Optimization of the Second Target Station cold source moderators using an automated workflow

The Second Target Station (STS) at the US Department of Energy’s Oak Ridge National Laboratory is designed to become the world’s highest peak-brightness spallation source of cold neutrons. Successful completion of the STS, which is currently in the preliminary design phase, will provide transformative new capabilities to examine novel materials for future technologies. At STS, neutrons will be generated by spallation reactions in a solid tungsten target. They will be moderated and thermalized in two cold (20 K) para-hydrogen moderators. Careful optimization of these moderators is essential to the project’s success. To find optimal moderator designs, an advanced optimization workflow integrates high-fidelity neutronics calculations using the Monte Carlo N-Particle (MCNP) transport code MCNP6.2 with state-of-the-art optimization algorithms in the Dakota optimization toolkit. For each design iteration, a parametrized solid CAD geometry is generated in Creo and automatically converted into an unstructured mesh geometry by Attila 4MC for the neutronics calculation with MCNP. Iterations repeat until optimal designs are found. Herein this paper presents the results of a sensitivity and optimization study for the cylindrical and tube moderators. Both moderators can be optimized for maximum peak brightness, maximum time-integrated brightness, or any combination between these extremes. Maximum peak brightness is achieved by using smaller optimal dimensions of the moderators, whereas maximum time-integrated brightness is achieved by using larger dimensions. A Pareto front details the designs that optimally balance both brightness metrics. The Pareto front can be found in only 40–110 iterations with 4–5 design parameters when using the efficient global and Pareto-set optimization algorithms in Dakota. Additionally, important engineering constraints can be taken into account, such as the coupling between the cylindrical moderator radius and aluminum vessel wall thicknesses required to ensure structural integrity of the vessels. This interaction has a significant impact on the resulting optimal designs. Our new, highly efficient, fully automated optimization workflow will be used to optimize additional STS components in the future and can be adopted for design and optimization studies at other experimental neutron and accelerator facilities.

46 INSTRUMENTATION RELATED TO NUCLEAR SCIENCE AND ↗
DOE OSTI2021

Synthesis of magnesiowüstite nanocrystallites embedded in an amorphous silicate matrix via low energy multiple ion implantations

The synthesis process is presented for experimentally simulating modifications in cosmic dust grains using sequential ion implantations or irradiations followed by thermal annealing. Cosmic silicate dust analogues were prepared via implantation of 20–80 keV Fe - , Mg - , and O - ions into commercially available p-type silicon (100)wafers. The as-implanted analogues are amorphous with a Mg/(Fe+Mg) ratio of 0.5 tailored to match theoretical abundances in circumstellar dusts. Before the ion implantations were performed, Monte-Carlo-based ion-solid interaction codes were used to model the dynamic redistribution of the implanted atoms in the silicon substrate. 600 keV helium ion irradiation was performed on one of the samples before thermal annealing. Two samples were thermally annealed at a temperature appropriate for an M-class stellar wind, 1000 K, for 8.3 h in a vacuum chamber with a pressure of 1 x 10 -7 torr. The elemental depth profiles were extracted utilizing Rutherford Backscattering Spectrometry (RBS) in the samples before and after thermal annealing. X-ray diffraction (XRD)analysis was employed for the identification of various phases in crystalline minerals in the annealed analogues. Transmission electron microscopy (TEM) analysis was utilized to identify specific crystal structures. RBS analysis shows redistribution of the implanted Fe, Mg, and O after thermal annealing due to incorporation into the crystal structures for each sample type. XRD patterns along with TEM analysis showed nanocrystalline Mg and Fe oxides with possible incorporation of additional silicate minerals.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2020

Modeling DC electrical breakdown using a truncated emission spectrum for trapped radiation

Spontaneously emitted radiation from excited atoms can be of principle importance in certain modes of electrical breakdown, especially positive streamers and some regimes of Townsend breakdown. The electrostatic particle-in-cell code Aleph utilizes the direct simulation Monte Carlo method to compute radiation transport. When there is strong radiation trapping, this approach is limited in that it must resolve the timescale associated with self-absorption. This renders many cases computationally intractable as sub-femtosecond time steps can be required to compute solutions for phenomena that occur over nanoseconds or microseconds. For two specific cases which exhibit strong radiation trapping, we find that spontaneous emissions having a frequency near the line center are inactive in the breakdown process and can be neglected. This enables larger time steps and a computational speedup of up to two orders of magnitude is observed. Some considerations for determining the validity of making such an approximation for Townsend breakdown problems and positive ionization wave problems are presented.

Roberds, N. A.↗
DOE OSTI2022

Computational investigation of incident ion angles and material erosion at rough graphite and silicon carbide divertor surfaces

We present a computational investigation of the dependence of material erosion on the incident ion angle at rough graphite and silicon carbide divertor surfaces. Ion angle distributions (IADs) for D plasmas at NSTX-U and DIII-D divertors were calculated by an equation-of-motion model that traces the ion trajectories in the sheath. Then, the effective sputtering yields and ion shadowed area fractions were calculated by a Monte Carlo micro-patterning and roughness code that applied the calculated IADs to surface topographic data that were obtained from optical confocal microscopy of rough graphite and SiC divertor surfaces from NSTX-U and DIII-D experiments. The calculations found that the effective sputtering yields, the sputtering pattern, and the shadowed area are determined by the detailed surface topology rather than the root mean square roughness R RMS , which represents deviations from a flat surface. The suppression of the effective sputtering yields for rough surfaces compared to the yield for a smooth surface was accounted for by the change of the mean local incident ion angle (LIIA) (θ'). The mean surface inclination angle distribution (SIAD) (δ) was found to be a useful parameter to estimate the LIIA from the calculated IADs. We report global empirical formulas for the mean LIIA and fraction of the area shadowed from the main ions for D plasmas for rough surfaces with B-field incident angles α = 85°–89° as a function of the mean SIAD (δ). We propose the use of the mean LIIA (θ') to estimate the sputtering yield for rough surfaces from the angular dependence of the sputtering yield.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2020

An Integral Experiment on Polyethylene Using Radiative Capture in Indium Foils in a High Flux D-D Neutron Generator

In this work, we report here the results of a measurement of the scattered versus unscattered neutron fluence on polyethylene determined via neutron activation of multiple natural indium foils from a deuterium-deuterium (D-D) neutron generator. The neutrons were produced by the High Flux Neutron Generator (HFNG) at the University of California, Berkeley, a specially designed source to maximize neutron flux on a sample while minimizing the total neutron yield. During the experiment, approximately 10 8 n/s were produced with the energies at the indium foils ranging from 2.2 to 2.8 MeV. Both the angle-integrated and the partial angle differential results are consistent with the predictions of the Monte Carlo N-Particle Transport (MCNP) code, using ENDF/B-VII.1. This supports shielding calculations in the fast energy region with high-density polyethylene.

42 ENGINEERING↗
DOE OSTI2022

Effects of neoclassical tearing modes and toroidal field ripple on lost alpha power in the SPARC tokamak

Here, using the SPIRAL Monte Carlo, full particle-orbit simulation code, we investigate the effects of neoclassical tearing modes (NTMs) and toroidal field (TF) ripple on alpha power losses during steady-state operation of the SPARC primary reference discharge. Model perturbations for TF ripple and the $m/n = 2/1$ and $3/2$ NTMs with exaggerated widths selected based on an H-mode plasma approaching thermal quench are added to a simulated SPARC magnetic equilibrium through which marker particles are tracked. The 3/2 and 2/1 NTMs are located at $\rho_{\mathrm{pol}} \sim 0.76$ and $\rho_\mathrm{pol} \sim 0.86$ respectively, well positioned to increase alpha particle transport into and within an outer lossy region of the plasma beyond $\rho_\mathrm{pol} \sim 0.8$ where over 95% of lost alpha particles are born. Total alpha power losses are shown to increase modestly from 1.73% lost at a minimum to 2.34% lost at a maximum, and alpha particle surface power densities form localized hotspots on the first-wall near the low-field side midplane due to NTMs and TF ripple. We establish a conservative upper limit for first-wall alpha surface power densities on a toroidally symmetric wall for typical, flattop operation and motivate the consideration of NTMs in the design of three dimensional limiter surfaces for SPARC.

alpha power losses↗
DOE OSTI2026

Numerical-heating effects in atmospheric pressure streamer discharges simulated with a PIC code

Artificial heating in plasma simulations is a well-known phenomenon which occurs when, among other things, the Debye length is poorly resolved by the simulation mesh. Here, in this work, the degree to which numerical-heating occurs during a simulation of a nanosecond atmospheric pressure streamer discharge is examined. The streamer is simulated using a two-dimensional finite-element, particle-in-cell code Empire, which uses direct simulation Monte Carlo for binary particle interactions. Initially, an estimate of the numerical-heating rate applied to Empire is performed using a simple plasma model. Second, a positive atmospheric pressure streamer discharge simulation is performed to study the effects of numerical heating on plasma density, electron temperature, and streamer velocity. The nominal Debye length is approximately 1 μm and the amount of numerical heating introduced in the simulation is varied by using mesh sizes ranging from 2 μm to 20 μm. A measurable numerical heating quantity is proposed that can be used to estimate the appropriate element size and quantify the numerical-heating that can be expected over the simulation time for an atmospheric pressure streamer. In conclusion while Δx/λ D violations can be an issue it is not likely to be an issue with streamer discharges that are temporally short and occur in environments where collision frequencies are high. This result validates the rationale of grid size choices for a large amount of previously published works where Δx/λ D violation was not clearly addressed. Primary finding of this work is that numerical heating is of minor concern for plasma simulations where electron–neutral collisions are numerous such that multiple collisions can occur within a single plasma period.

Nikic, Dejan [University of New Mexico, Albuquerqu↗
DOE OSTI2017

DaMaSCUS: the impact of underground scatterings on direct detection of light dark matter

Conventional dark matter direct detection experiments set stringent constraints on dark matter by looking for elastic scattering events between dark matter particles and nuclei in underground detectors. However these constraints weaken significantly in the sub-GeV mass region, simply because light dark matter does not have enough energy to trigger detectors regardless of the dark matter-nucleon scattering cross section. Even if future experiments lower their energy thresholds, they will still be blind to parameter space where dark matter particles interact with nuclei strongly enough that they lose enough energy and become unable to cause a signal above the experimental threshold by the time they reach the underground detector. Therefore in case dark matter is in the sub-GeV region and strongly interacting, possible underground scatterings of dark matter with terrestrial nuclei must be taken into account because they affect significantly the recoil spectra and event rates, regardless of whether the experiment probes DM via DM-nucleus or DM-electron interaction. To quantify this effect we present the publicly available Dark Matter Simulation Code for Underground Scatterings (DaMaSCUS), a Monte Carlo simulator of DM trajectories through the Earth taking underground scatterings into account. Our simulation allows the precise calculation of the density and velocity distribution of dark matter at any detector of given depth and location on Earth. The simulation can also provide the accurate recoil spectrum in underground detectors as well as the phase and amplitude of the diurnal modulation caused by this shadowing effect of the Earth, ultimately relating the modulations expected in different detectors, which is important to decisively conclude if a diurnal modulation is due to dark matter or an irrelevant background.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2024

Determining the 9 $\mathrm{Be}$($n, γ$) 10 $\mathrm{Be}$ integral cross section at fission neutron energies

The 9 B neutron capture cross section has significant implications for Be materials in the nuclear industry as well as the α process in stellar nucleosynthesis. While the cross section is well constrained at thermal neutron energies, there is a lack of experimental data at higher neutron energies, and the evaluated nuclear data libraries can differ by up to two orders of magnitude. We calculate the 9 Be(n, γ) 10 Be integral cross section at fission neutron energies in an effort to resolve disagreements amongst the nuclear data libraries. Foil irradiation experiments were performed using the Flattop critical assembly at the National Criticality Experiments Research Center with either the highly enriched U or Pu cores, with target foil stacks placed at multiple locations to exploit different neutron energy profiles. Accelerator mass spectrometry was used to measure the 10 Be/ 9 Be ratio in irradiated Be foils, while all other activation products were quantified through gamma spectrometry. The experiments were simulated using the Monte Carlo N-Particle radiation transport code and combined with experimental results to determine the total neutron fluence, while the staysl-pnnl suite and fispact-ii code were used to validate the model and assess the systematic uncertainty. The new 9 Be(n, γ) 10 Be integral cross sections calculated in this work are 26.5 ± 2.2µb at 0.59 ± 0.07 MeV, 24 ± 3 µb at 0.98 ± 0.14 MeV, 21.7 ± 1.3 µb at 1.26 ± 0.11 MeV, 21.8 ± 1.4 µb at 1.32 ± 0.11 MeV, and 18.6 ± 1.1 µb at 1.46 ± 0.13 MeV. These results do not agree with integral cross sections from any of the nuclear data library evaluations. Discrepancies between the new integral cross sections reported here and the nuclear data libraries suggest a more complex cross-section structure in the MeV range which allows for more resonance contributions, and more work is needed to further constrain the evaluated cross sections.

38 RADIATION CHEMISTRY, RADIOCHEMISTRY, AND NUCLEA↗
DOE OSTI2024

Testing the Activation Analysis for Fusion in OpenMC

OpenMC is a community-developed Monte Carlo neutron and photon transport simulation code. It can perform fission simulations such as fixed-source, k-eigenvalue, and subcritical multiplication calculations on models built using either a constructive solid geometry or CAD representation. To explore the use of OpenMC for fusion activation analysis, a detailed model of the Fusion Neutronics Science Facility (FNSF) was first developed for comparisons against an existing SERPENT model. A 90-degree model of FNSF in Standard-Triangle-Language (STL) CAD format was converted to Constructive Solid Geometry (CSG) using each code's built-in functions, and the geometries were validated by ensuring no cells overlapped and no particles were lost during simulations. The neutron fluxes were calculated and compared for multiple components close to the plasma. The results show differences mostly below 1% in fluxes and averaged 8% for activity and decay heat. Here, the work described in this study tests the CAD-based geometry using the DagMC toolkit in OpenMC and compares the activation analysis of OpenMC to SERPENT code.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2021

Yet Another Hauser-Feshbach Code

Code framework to perform simulations of nuclear reactions with incident particles including gammas, protons, neutrons, deuterons, tritons, helium-3, and alphas to develop nuclear data libraries and study nuclear reaction phenomona. The code can model nuclear reactions or initial populations specified by the user. The decays principally follow the compound-nucleus hypothesis and the statistical decay of Hauser and Feshbach. The code models nuclear decays with a Monte Carlo process that tracks the decay of each state. Data libraries are produced that can be translated into GNDS (generalized Nuclear Data Structure). YAHFC can also be used as an event generator, with individual decays written to disk for offline analysis. Both serial and MPI versions are provided. YAHFC does NOT perform transport of nuclear reactions through materials, only the primary reaction with a target material.

Ormand, WilliamE.↗
DOE OSTI2019

FLUKA-MARS15 simulations to optimize the Fermilab PIP-II movable beam absorber

PIP-II is the Fermilab's flagship project to provide powerful, high-intensity proton beams to the laboratory's experiments. The heart of the PIP-II project is an H- 800 MeV superconducting linear accelerator. In order to commission the beam and operate safely the linac, several constraints were evaluated. The design of a movable 5 kW beam absorber was finalized to allow staged beam commissioning in different linac locations. Prompt and residual radiation levels were calculated, and radiation shields were optimized to keep those values within the acceptable levels in the areas surrounding beam absorber. Monte Carlo calculations with FLUKA and MARS15 codes are presented in the paper to support these studies.

43 PARTICLE ACCELERATORS↗
DOE OSTI2020

Processing MCNP Elemental Edit Outputs

The Los Alamos National Laboratory's (LANL) Monte Carlo N-Particle® (MCNP®) transport code version 6 has the capability for tracking particles on unstructured mesh (UM) geometry models. The MCNP UM feature was originally designed for the UM models created by the Abaqus/CAE software product. The MCNP version 6.2.0 and later recognizes the UM models read from Abaqus INP files or MCNPUM files converted from Abaqus INP files. Other finite element analysis software packages may generate UM models and then convert these models into Abaqus INP formats.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2020

Generating MCNP Input Files for Unstructured Mesh Geometries

Los Alamos National Laboratory's (LANL) Monte Carlo N-Particle (MCNP) transport code version 6 has the capability for tracking particles on unstructured mesh (UM) geometry models. The MCNP UM feature has been developed for performing calculations of complex geometry models. This capability tracks particles on hybrid geometries where finite element meshes are embedded into constructive solid geometry (CSG) cells. The MCNP UM feature was originally designed to read UM models created by Abaqus/CAE software suite. MCNP versions 6.2.0 and later can process UM models read from Abaqus input les or MCNPUM les converted from Abaqus input les. Sandia National Laboratory's Cubit Toolkit and other finite element analysis software packages may generate UM models and then convert these models into Abaqus input formats.

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2019

Neutron Activation Self-Shielding Factors for Common Dosimetry Foils in an ACRR Equivalent Environment

The results of a computational analysis of self-shielding factors are presented. The analysis highlights the total self-shielding, which is a combination of energy and spatial self-shielding, associated with different neutron detection materials. The Monte Carlo N-Particle (MCNP) transport code was used in conjunction with the Evaluated Nuclear Data File (ENDF) and the International Reactor Dosimetry and Fusion Files (IRDFF). This analysis was done with neutron activation analysis in mind, and therefore is modeled and presented in a similar fashion.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2022

Preconceptual Design of Irradiated Fuel Salt Management System

The National Reactor Innovation Center was established by the U.S. Department of Energy to accelerate the demonstration and deployment of advanced nuclear reactors. To meet this mission, the National Reactor Innovation Center has developed the Laboratory for Operations and Testing in the U.S. (LOTUS) to support the first fast-spectrum molten salt reactor demonstration. The goal of this research is to understand the system requirements that may be applied to a molten salt reactor experiment in the LOTUS test bed and to perform systems analyses and preconceptual development of a storage container for the management of the irradiated molten salt fuel. This study proposes a set of irradiated fuel salt management processes for the molten salt reactor, from defueling to storage or disposal. For that, the historic Molten Salt Reactor Experiment (MSRE) at Oak Ridge National Laboratory has served as a crucial source of data for a mock application. This paper proposes a potential geometry for storing irradiated fuel salt based on MSRE data and the Monte Carlo N-Particle radiation transport code. The geometry was analyzed for criticality safety under abnormal situations, such as water ingress and interaction between containers. Based on the determined configuration of the storage system, the decay heat produced from the irradiated fuel salts was calculated to provide a reasonable time scale of the dormancy period. Using the same geometry assumed above, the neutron and gamma dose rates from the irradiated fuel salt at the end of the dormancy period were estimated, which will be useful for understanding the handling and radiation protection requirements for handling the irradiated fuel salt. Lastly, a reactor in the LOTUS test bed could use chloride-based fuel as opposed to the fluoride-based fuel used in MSRE. Thus, a literature study provided data to understand the similarities and differences between the characteristics of fluoride- and chloride-based salts in terms of salt chemistry, material characteristics, postirradiation behavior, neutronics, thermal hydraulics, and heat removal. This study, together with the proposed irradiated fuel salt management processes, preliminary criticality calculations, decay heat, and dose rate estimates, provides a strong basis for future investigation into the management of irradiated fuel salts generated during molten salt reactor technology demonstrations.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗