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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 217 records · Page 12
All sourcesNASA NTRSDOE OSTI
DOE OSTI2026

Sensitivity Analysis of Numerical Modeling Input Parameters on Wind Turbine Loads in Deterministic Transient Load Cases

Aero-hydro-elastic-servo numerical models used to design and analyze wind turbines are based on thousands of variable input parameters that dictate the inflow, aerodynamic, structural, and control characteristics of the system as well as sea state, hydrodynamic, and mooring characteristics for fixed-bottom and floating offshore wind turbines. Each of these parameters has some level of uncertainty, which can significantly impact the predicted loads. Understanding the uncertainty in the inputs is critical to understanding the uncertainty in the outputs. This work demonstrates a screening technique to identify which parameters ultimate loads are most sensitive to so that more focus can be given to quantifying the possible range of those parameters. This technique has been demonstrated previously for different turbine and load case types and is extended here for a floating offshore wind turbine in design load cases with transient events both in the inflow and operations. Each load case features a deterministic gust, including variations in wind speed, direction, and shear. Load cases are considered with an operating turbine as well as with prescribed fault, startup, and shutdown procedures. The study found that key input parameters with a large impact on loads include the length of the gust, the magnitude of direction change and speed in the gust, the initial wind speed, and the shape of the gust profile.

17 WIND ENERGY↗
DOE OSTI2025

ADEPT: A Pedagogical Framework for Integrating Agentic AI with Deterministic Scientific Workflows

The integration of Large Language Models (LLMs) into scientific research promises to accelerate discovery, yet a significant gap remains between the dynamic reasoning of Artificial Intelligence (AI) agents and the static, deterministic nature of canonical scientific workflows. This paper introduces ADEPT (Agentic Discovery and Exploration Platform for Tools), a reference architecture and pedagogical framework explicitly designed to bridge this gap. ADEPT's primary mission is to provide a transparent, "glass-box" environment where researchers and engineers can learn to effectively wrap established scientific software (e.g., BLAST, Nextflow pipelines) and compose it into reliable, agent-driven workflows. We describe its modular, multi-server architecture, which leverages the Model Context Protocol (MCP) for tool serving, LangGraph for robust agentic orchestration, and a secure nsjail-based sandbox for safe code execution. By prioritizing architectural clarity, safety, and modularity, ADEPT serves as an extensible blueprint for building trustworthy AI-augmented systems and fosters the collaborative development necessary to responsibly employ agentic AI for science. We provide practical examples of how to adapt and extend this framework, highlighting its utility in workforce development and AI-readiness capabilities across research and development projects.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2012

Nano transfer and nanoreplication using deterministically grown sacrificial nanotemplates

Methods, manufactures, machines and compositions are described for nanotransfer and nanoreplication using deterministically grown sacrificial nanotemplates. An apparatus, includes a substrate and a nanoconduit material coupled to a surface of the substrate. The substrate defines an aperture and the nanoconduit material defines a nanoconduit that is i) contiguous with the aperture and ii) aligned substantially non-parallel to a plane defined by the surface of the substrate.

77 NANOSCIENCE AND NANOTECHNOLOGY↗
DOE OSTI2014

A deterministic discrete ordinates transport proxy application

Kripke is a simple 3D deterministic discrete ordinates (Sn) particle transport code that maintains the computational load and communications pattern of a real transport code. It is intended to be a research tool to explore different data layouts, new programming paradigms and computer architectures.

Source record↗
DOE OSTI2019

Coupled Deterministic and Monte Carlo Neutronics for Vessel Fluence Calculations

Throughout core life, the reactor vessel is continually bombarded with neutron flux embrittling the vessel and limiting its useful life. As a result, it is imperative to adequately predict how much flux has passed through the vessel during all stages of its life, as a means of ensuring it has adequate ductility to preclude a brittle fracture even under design accident conditions. At present many of these predictions are made either by purely deterministic or purely stochastic means, each offering their own advantages and disadvantages. This paper introduces a coupled method for calculating fluence through the VERA framework, which enables modeling incremental accumulations of fluence while simultaneously accounting for several factors as they evolve through core life. The intention of this method is to simplify the calculation process and provide better resource utilization in obtaining the required accuracy.

Henderson, Shane↗
DOE OSTI2020

Deterministic Discrete Fracture Network (DFN) Model for the EGS Collab Project on the 4850 Level of the Sanford Underground Research Facility (SURF)

The EGS Collab is conducting hydraulic fracture stimulation and fluid circulation experiments in the Sanford Underground Research Facility (SURF) located in Lead, South Dakota. A total of eight ~60m-long subhorizontal boreholes were drilled from the 4850 Level (~1.5 km below the ground surface) into the crystalline rock of this former mine. Six of these holes are used for geophysical monitoring, one is used for hydraulic fracture stimulation, and the remaining hole was designed as a production borehole that receives water from the injection well via the induced and natural fracture system. The primary goal of creating the discrete fracture network model is to show that these modeling methods are critical for the development of enhanced geothermal systems (EGS). This includes the prediction of rock behavior during fracturing and during an extended period of water flow between the parallel injection and production boreholes. Understanding the results from the induced fracturing and flow is complicated by the presence of significant natural fractures that interact with the stimulation and/or flow pathways. The delineation and characterization of natural fractures is thus an important part of the project, and therefore a model of the Discrete Fracture Network (DFN) was developed on a deterministic basis. The DFN was populated using observations and interpretations integrated from drift (horizontal passageways that allow access in the underground) fracture mapping, analysis of core recovered from the eight boreholes, borehole televiewer logs and videos, and observations of flow between and within boreholes and in the drift. The natural fracture system is dominated by a pervasive northwest-trending, steeply dipping shear system that is identifiable in the drifts and the core. Hydraulic fracture stimulation, flow/tracer circulation tests, and geophysical monitoring revealed that the behavior of the injected water, and perhaps the growth of induced fractures, has been significantly influenced by the existing fractures identified in the DFN.

40 EE - Geothermal Technologies Office (EE-4G)↗
DOE OSTI2021

Hierarchical porous metals with deterministic 3D morphology and shape via de-alloying of 3D printed alloys

The present disclosure relates to a method for forming a three dimensional, hierarchical, porous metal structure with deterministically controlled 3D multiscale pore architectures. The method may involve providing a feedstock able to be applied in an additive manufacturing process, and using an additive manufacturing process to produce a three dimensional (3D) structure using the feedstock. The method may involve further processing the 3D structure through at least a de-alloying operation to form a metallic 3D structure having an engineered, digitally controlled macropore morphology with integrated nanoporosity.

Qi, Zhen↗
DOE OSTI2022

Multi-channel high-speed converter clock synchronization with autonomous coherent deterministic latency

A system and corresponding method that achieves coherency and deterministic latency (CDL) autonomously upon power on is disclosed. The system, for example, a multi-channel RF system, may require CDL with respect to the digital-to-analog converters (DACs) and analog-to-digital converters (ADCs) assigned to the channels in the system. CDL is achieved through a timed combination of external reference and synchronization signals, resetting and disabling of various clock dividers, and enabling clock generation. In addition to synchronizing all of the clocks, the data acquisition sequence must be synchronized across all of the channels, whether they are on chips, cards, or chassis. Data acquisition synchronization may be implemented using an initiator/target or a wired OR mode configuration.

Brady, Charles E.↗
DOE OSTI2012

Post irradiation experiment analysis using the APOLLO2 deterministic tool. Validation of JEFF-3.1.1 thermal and epithermal actinides neutron induced cross sections through MELUSINE experiments

Two different experiments performed in the 8 MWth MELUSINE experimental power pool reactor aimed at analyzing 1 GWd/t spent fuel pellets doped with several actinides. The goal was to measure the averaged neutron induced capture cross section in two very different neutron spectra (a PWR-like and an under-moderated one). This paper summarizes the combined deterministic APOLLO2-stochastic TRIPOLI4 analysis using the JEFF-3.1.1 European nuclear data library. A very good agreement is observed for most of neutron induced capture cross section of actinides and a clear underestimation for the {sup 241}Am(n,{gamma}) as an accurate validation of its associated isomeric ratio are emphasized. Finally, a possible huge resonant fluctuation (factor of 2.7 regarding to the 1=0 resonance total orbital momenta) is suggested for isomeric ratio. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2012

Simplified treatment of exact resonance elastic scattering model in deterministic slowing down equation

Simplified treatment of resonance elastic scattering model considering thermal motion of heavy nuclides and the energy dependence of the resonance cross section was implemented into NJOY [1]. In order to solve deterministic slowing down equation considering the effect of up-scattering without iterative calculations, scattering kernel for heavy nuclides is pre-calculated by the formula derived by Ouisloumen and Sanchez [2], and neutron spectrum in up-scattering term is expressed by NR approximation. To check the verification of the simplified treatment, the treatment is applied to U-238 for the energy range from 4 eV to 200 eV. Calculated multi-group capture cross section of U-238 is greater than that of conventional method and the increase of the capture cross sections is remarkable as the temperature becomes high. Therefore Doppler coefficient calculated in UO{sub 2} fuel pin is calculated more negative value than that on conventional method. The impact on Doppler coefficient is equivalent to the results of exact treatment of resonance elastic scattering reported in previous studies [2-7]. The agreement supports the validation of the simplified treatment and therefore this treatment is applied for other heavy nuclide to evaluate the Doppler coefficient in MOX fuel. The result shows that the impact of considering thermal agitation in resonance scattering in Doppler coefficient comes mainly from U-238 and that of other heavy nuclides such as Pu-239, 240 etc. is not comparable in MOX fuel. (authors)

73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗
DOE OSTI2012

Mesh generation and energy group condensation studies for the jaguar deterministic transport code

The deterministic transport code Jaguar is introduced, and the modeling process for Jaguar is demonstrated using a two-dimensional assembly model of the Hoogenboom-Martin Performance Benchmark Problem. This single assembly model is being used to test and analyze optimal modeling methodologies and techniques for Jaguar. This paper focuses on spatial mesh generation and energy condensation techniques. In this summary, the models and processes are defined as well as thermal flux solution comparisons with the Monte Carlo code MC21. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2013

Feasibility of a Monte Carlo-deterministic hybrid method for fast reactor analysis

A Monte Carlo and deterministic hybrid method is investigated for the analysis of fast reactors in this paper. Effective multi-group cross sections data are generated using a collision estimator in the MCNP5. A high order Legendre scattering cross section data generation module was added into the MCNP5 code. Both cross section data generated from MCNP5 and TRANSX/TWODANT using the homogeneous core model were compared, and were applied to DIF3D code for fast reactor core analysis of a 300 MWe SFR TRU burner core. For this analysis, 9 groups macroscopic-wise data was used. In this paper, a hybrid calculation MCNP5/DIF3D was used to analyze the core model. The cross section data was generated using MCNP5. The k{sub eff} and core power distribution were calculated using the 54 triangle FDM code DIF3D. A whole core calculation of the heterogeneous core model using the MCNP5 was selected as a reference. In terms of the k{sub eff}, 9-group MCNP5/DIF3D has a discrepancy of -154 pcm from the reference solution, 9-group TRANSX/TWODANT/DIF3D analysis gives -1070 pcm discrepancy. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2013

Computing the moments of the neutron population using deterministic neutron transport

It is important to treat the inherent stochasticity of the fission process in systems where the behavior of the system is stochastic. This occurs when there are few neutrons in the system, or when the neutron source is weak. In order to characterize such systems, the capability to compute the first four moments of the neutron population distribution has been added to the deterministic neutral particle transport code, PARTISN. The moments are then fitted to probability density functions from the Pearson family. PARTISN is compared against MCNP6, with which it agrees well. (authors)

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Investigation of the Attila Deterministic Solver for Out-of-Field Organ Dose Calculations for Radiotherapy Patients

An undeniable paradox in the current practice of radiation oncology is that the therapeutic radiation often used to treat cancer is also a known carcinogen. Indeed, the potential for inducing second cancers in patients receiving radiotherapy has been well established. Retrospective epidemiological studies are an important way to assess the risks of late effects in radiotherapy patients. Such investigations attempt to correlate organ dose with the incidence of secondary cancers in a cohort of patients. Individualized patient dosimetry is a critical component of these studies and is challenging for two reasons. First, medical images describing the patient's anatomy are either non-existent or, if they exist, often only cover a portion of the body. This means that whole body computational human phantoms must be used as a surrogate for the patient's true anatomy in order to perform dose calculations far from the treatment site. Second, the gold standard dosimetry approach calls for the use of Monte Carlo radiation transport simulations which are far too slow for studies involving large-scale patient cohorts. This paper describes a preliminary effort at the National Cancer Institute aimed at addressing these dosimetry challenges. The Attila software (Varian Medical Systems, Palo Alto, CA) is investigated as an alternative to Monte Carlo for performing out-of-field organ dose calculations for radiotherapy patients. Whereas as Monte Carlo uses stochastic methods, Attila can deterministically solve the linear Boltzmann equation throughout the entire solution space. It is hypothesized that dose calculations with Attila will offer comparable accuracy to Monte Carlo, but with significant faster computation times. At least one study has demonstrated this for in-field dose calculations [3]. To our knowledge Attila has not yet been used to perform out-of-field organ dosimetry. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

Resonance Self-Shielding Using Deterministic Transport Methods with Pointwise Cross Sections (Invited) - Eugene P. Wigner Reactor Physics Award

The difficulty of predicting resonance absorption has been recognized since the birth of reactor physics. A paper published by Wigner, et al. in 1955 describes pioneering theoretical work done in 1941 to provide an estimate for the resonance absorption in heterogeneous uranium lattices. This work introduced many fundamental concepts, including the narrow resonance (NR) approximation and the Wigner rational expression for the escape probability from a fuel lump, and it established an equivalence relation for resonance absorption in homogeneous and heterogeneous systems. During the 1960's, Bondarenko proposed an approach in which resonance shielding factors are precomputed and parametrized as a function of the background cross section by using the NR approximation and equivalence theory. This provided an efficient technique for treating resonance absorption in reactor lattices. The Bondarenko method and its modern variations (e.g., certain subgroup techniques) have become the predominate method for resonance shielding calculations in reactor physics. However, this approach has several limitations, most of which are well known, such as the treatment of resonance interference effects and its reliance on homogeneous-heterogeneous equivalence theory relationships. Other limitations are more subtle and less generally recognized, such as neglecting the impact of self-shielding the multigroup (MG) elastic removal cross section and the effects of directionally dependent shielding on the total cross section. This paper and the accompanying presentation apply deterministic solutions of the continuous energy (CE) neutron transport equation to examine common approximations made in resonance methods for a uniform lattice. The calculated fine-structure spectra are used to calibrate the impact of the approximations and describe how results from CE flux solution can be incorporated in the Bondarenko self-shielding method to improve the accuracy. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗