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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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At least 127 records · Page 7
All sourcesNASA NTRSDOE OSTI
DOE OSTI2015

Neutronics, Steady-State, and Transient Analyses for the Kazakhstan VVR-K Reactor with LEU Fuel

Calculations have been performed for steady state and postulated transients in the VVRK reactor at the Institute of Nuclear Physics (INP) in Alatau, Kazakhstan. These calculations have been performed at the request of staff of the INP who have performed similar calculations. Calculations were performed for the fresh low-enriched uranium (LEU) core and for four subsequent cores as beryllium is added as a radial reflector to maintain criticality during the first 15 cycles of operation. The calculations include neutronics parameters, steady-state power and temperature distributions, and response to transients. These calculations have been performed independently from those being performed by INP and serve as one step in the verification process.

Hanan, N. A.↗
DOE OSTI2017

U-10Mo Conversion Burnup Benchmarking Activities for Full-Size LEU Fuel Plates

Validation benchmark experiments for low-enriched uranium (LEU) U-Mo fuel systems are currently being developed from existing experiments irradiated in the Advanced Test Reactor. (ATR) and ATR Critical Facility (ATRC). The experiments chosen include the ATR Full size plate In center Flux trap Position (AFIP-7) experiment and the Ki-Jang Research Reactor Lead Test Assembly (KJRR-LTA) located in the Northeast Flux Trap. Each experiment was modeled using a high-fidelity three-dimensional Monte Carlo code with depletion capabilities. A description of the validation experiments will be provided as well as comparison to post irradiation examination results for the AFIP-7 experiment. The PIE results for the AFIP-7 experiment have demonstrated good agreement with analytical models. Due to the complex nature in of the ATR, modeling simplifications to support benchmark activities using lower fidelity reactor physics software. A discussion of the analytical bias associated with modeling simplifications will also be provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Modeling a Thermal-Spectrum LEU-fueled Molten Salt Reactor Co-fueled with Thorium in SCALE 6.3.1 and Serpent-2

Interest in the development of advanced nuclear fission reactors for commercial electricity production has risen in recent years, with Generation IV reactor designs offering numerous advantages in safety, efficiency, fuel cycle sustainability, and waste management. In particular, thorium-fueled molten salt reactors (MSRs) are considerably promising for enhancing fuel cycle sustainability in their ability to breed fissile 233U fuel from thorium, a presently untapped and widely abundant resource. A novel MSR fuel cycle concept, the ”Sourdough” refueling and waste management strategy, has previously been demonstrated with a traditional uranium-based fuel cycle in a thermal-spectrum MSR operating with low enriched uranium (LEU) fuel with favorable neutronic performance. However, the ability to use this unique fuel cycle approach with thorium-based molten salt fuels has not yet been studied. In this work, the Sourdough fuel cycle was implemented in a small, thermal-spectrum MSR fueled with high assay low enriched uranium (HALEU) and fertile 232Th for breeding 233U fuel. Relevant neutronic data, including fuel and isothermal temperature feedback behavior, was studied using the SCALE 6.3.1 and Serpent-2 code systems, and keff data was measured during simulated depletion at 400 MWth. The Sourdough fuel cycle concept is shown to perform favorably with a thorium-fueled MSR model, thus warranting further study into its use in other MSR designs.

11 - NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2017

Heat deposition analysis for the High Flux Isotope Reactor’s HEU and LEU core models

The High Flux Isotope Reactor at Oak Ridge National Laboratory is an 85 MW th pressurized light-water-cooled and -moderated flux-trap type research reactor. The reactor is used to conduct numerous experiments, advancing various scientific and engineering disciplines. As part of an ongoing program sponsored by the US Department of Energy National Nuclear Security Administration Office of Material Management and Minimization, studies are being performed to assess the feasibility of converting the reactor’s highly enriched uranium fuel to low-enriched uranium fuel. To support this conversion project, reference models with representative experiment target loading and explicit fuel plate representation were developed and benchmarked for both fuels to (1) allow for consistent comparison between designs for both fuel types and (2) assess the potential impact of low-enriched uranium conversion. These high-fidelity models were used to conduct heat deposition analyses at the beginning and end of the reactor cycle and are presented herein. This article (1) discusses the High Flux Isotope Reactor models developed to facilitate detailed heat deposition analyses of the reactor’s highly enriched and low-enriched uranium cores, (2) examines the computational approach for performing heat deposition analysis, which includes a discussion on the methodology for calculating the amount of energy released per fission, heating rates, power and volumetric heating rates, and (3) provides results calculated throughout various regions of the highly enriched and low-enriched uranium core at the beginning and end of the reactor cycle. These are the first detailed high-fidelity heat deposition analyses for the High Flux Isotope Reactor’s highly enriched and low-enriched core models with explicit fuel plate representation. Lastly, these analyses are used to compare heat distributions obtained for both fuel designs at the beginning and end of the reactor cycle, and they are essential for enabling comprehensive thermal hydraulics and safety analyses that require detailed estimates of the heat source within all of the reactor’s fuel element regions.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2018

Completion of the ATR LEU Conversion Conceptual Design

The review period was initiated with a kick-off meeting held October 26, 2017. The design review package (Table 1) was presented and reviewer responsibilities and expectations discussed. Presentations by the project team included a project overview and summaries of the neutronic, thermal, and structural analysis. The participants were instructed to review the design in accordance with their knowledge area and submit written comments, questions, or deficiencies (CQDs). The CQDs were received from reviewers on form 412.13. Their comments were resolved and acceptance signatures were received.

99 GENERAL AND MISCELLANEOUS↗
DOE OSTI2022

Nondestructive and destructive assay for forensics characterization of weapons-grade plutonium produced in LEU irradiated in a thermal neutron spectrum

A post-irradiation examination (PIE) of an enriched uranium sample was performed to advance a nuclear forensics methodology, based on intra-elemental nuclide ratios. The PIE was carried out on a few milligrams of low enriched uranium dioxide (LEUO 2 ) sample, irradiated to a burnup of 1 GWd/MTU. The PIE of the LEUO 2 sample provided the concentration of selected plutonium and fission product nuclides ( 239 Pu, 240 Pu, 241 Pu, 91 Y, 95 Zr, 95 Nb, 103 Ru, 133 Cs, 134 Cs, 135 Cs, 137 Cs, 136 Ba, 138 Ba, 140 Ba, 140 La, 141 Ce, 144 Ce, 149 Sm, 150 Sm, 152 Sm, 153 Eu, and 154 Eu). The PIE supported validation of the MCNP (Monte Carlo N-Particle neutronics simulation code) predicted concentration of these nuclides in the irradiated LEUO 2 sample. This study showed that most of the MCNP predicted nuclide concentrations were within 15% and several within 10% compared to the PIE results. Furthermore, gamma and mass spectrometry of nuclides in irradiated LEUO 2 enabled the determination of uranium burnup and time since irradiation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

Modeling and experimental assessment of a buried Leu–Ile mutation in dengue envelope domain III

Envelope protein domain III (ED3) of the dengue virus is important for both antibody binding and host cell interaction. Here, we focused on how a L387I mutation in the protein core could take place in DEN4 ED3, but cannot be accommodated in DEN3 ED3 without destabilizing its structure. To this end, we modeled a DEN4-L387I structure using the Penultimate Rotamer Library and taking the DEN4 ED3 main-chain as a fixed template. We found that three out of seven Ile{sup 387} conformers fit in DEN4 ED3 without introducing the severe atomic clashes that are observed when DEN3 serotype’s ED3 is used as a template. A more extensive search using 273 side-chain rotamers of the residues surrounding Ile{sup 387} confirmed this prediction. In order to assess the prediction, we determined the crystal structure of DEN4-L387I at 2 Å resolution. Ile{sup 387} indeed adopted one of the three predicted rotamers. Altogether, this study demonstrates that the effects of single mutations are to a large extent successfully predicted by systematically modeling the side-chain structures of the mutated as well as those of its surrounding residues using fixed main-chain structures and assessing inter-atomic steric clashes. More accurate and reliable predictions require considering sub-angstrom main-chain deformation, which remains a challenging task. - Highlights: • We mutated L387I of DEN4 ED3 and examined its effects on structure and stability. • We modeled the side-chain of Ile{sup 387} using DEN4 ED3's structure as a template. • We determined the crystal structure of DEN4-L387I and confirmed the modeling. • Side-chain repacking occurring around Ile{sup 387} involved >3 inter-connected residues. • These results explained why L387I mutation in DEN4 ED3 conserves thermostability.

60 APPLIED LIFE SCIENCES↗
DOE OSTI2024

Upsampling Monte Carlo reactor simulation tallies in depleted LWR assemblies fueled with LEU and HALEU using a convolutional neural network

Simulating nuclear reactor cores at the highest achievable spatial and energy resolution is critical in modeling these systems accurately. Increasing the resolution, however, can dramatically increase the memory and central processing unit time required to run simulations. A convolutional neural network was shown previously to accurately upsample tally results of simulated light water reactor assemblies fueled with fresh, low enriched uranium. Here, we show that a convolutional neural network can be used to upsample tally results in assemblies containing fresh and depleted fuel enriched from 1.6 to 19.9 atom percent. The network was trained using neutron flux tallies from simulations of light water reactor assemblies with a range of fuel and coolant temperatures and a diverse selection of geometries. Accurate predictions of flux tallies are possible even on test assemblies with geometries and burnup levels well outside the range of those present in the training and validation data. The network improves the data density by a factor of 8 over a broad range of light water reactor assemblies while incurring insignificant additional computational cost to a Monte Carlo simulation.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗