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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 973 records · Page 54
All sourcesNASA NTRSDOE OSTI
DOE OSTI2016

Preliminary structural assessment of VHTR graphite components according to ASME code - 18532

In this study, a preliminary structural assessment on a graphite component of a VHTR was done in accordance with the semi-probabilistic assessment procedures of the latest 2015 edition of the ASME Boiler and Pressure Vessel Code, Section III, Div. 5, Subpart HHA-Graphite Materials. Establishment of the Weibull distribution for the tensile strength is essential to set up the stress limits based on the probability of failure of the graphite materials. All parameters required in the ASME code were identified and evaluated for a nuclear graphite material, IG-110. The assessment of design level was done on one of fuel elements in a Korean VHTR design by two methodologies in the ASME code: (1) Simplified assessment and (2) Full assessment. The structural integrity of the fuel element was assured by both assessment methods. Also, the effect of the number of integration volume groups on the probability of failure in the full assessment was investigated and a recommendation for the conservative results was suggested. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

The Optimized Parallel MOC module in Code cosLATC

A new neutron transport lattice physics code, cosLATC, which is an essential part in COSINE (Core and System Integrated Engine) code package, has been developed by SPICRI. It is mainly used for 2-D transport calculations on Pressurized Water Reactor (PWR) assemblies or pin cells. cosLATC generates few group constants for the core physics code cosCORE in COSINE package. With regards to cross-section data, WIMS-D 69 and 281 energy groups library is employed currently. cosLATC also has capability of resonance treatment, depletion and fuel temperature calculation. cosLATC uses modular tracking and parallelization technique in MOC transport calculation, which can save lots of memory and running time. In this paper, MOC module in cosLATC are introduced, the accuracy and efficiency of the module are verified by C5G7 benchmark and a PWR typical assembly. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Dakota Uncertainty Quantification Methods Applied to the CFD code Nek5000

Predictive modeling and simulation of the performance of a nuclear reactor and its fuel is a challenging task because of the large number of coupled physical phenomena that must be addressed. In addition to this intrinsic complexity, model uncertainty must be accounted for in any analysis if the model will be used to facilitate design or operational decisions which may impact safety and performance. Rigorous, structured uncertainty analyses are performed by first characterizing the model's input uncertainties and, by propagating the uncertainty through the model in order to estimate the output's uncertainty. This paper presents the state of advancement of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) project to characterize the uncertainty of the CFD code Nek5000 using the Dakota package. Nek5000 is a high order spectral element computational fluid dynamics (CFD) code developed at Argonne National Laboratory for high resolution spectral-filtered large eddy simulations (LESs) and unsteady Reynolds averaged Navier-Stokes (RANS) simulations. The Dakota package developed at Sandia National Laboratory can be integrated with many scientific and engineering codes to facilitate efficient, effective uncertainty quantification (UQ) and sensitivity analyses. Dakota is designed to interface with external packages/codes such as Nek5000 through a script-driven interface called fork. Such interface provides maximum flexibility to the users for pre- and post-processing tasks preceding and following each function evaluation. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2017

Evaluation of the New Boron Model of the TRACE Code

System codes such as TRACE and RELAP5 have the option of adding boron to be dissolved in the water. In principle, the density and the viscosity of the water are not affected when the amounts of dissolved boron are very small. However, if the boron concentration is significant, water density and viscosity will be affected. Version 5.890 of the TRACE code has a new boron model with the option of modifying the water density and viscosity to account for the presence of dissolved boron. The input variable ISOLUT is used for this purpose. If ISOLUT=0, no boron is dissolved in the water. If ISOLUT=1, boron is dissolved in the water, and the liquid water density and viscosity are not modified by the dissolved boron. If ISOLUT=2, boron is dissolved in the water, and the liquid water density and viscosity are modified by the dissolved boron. This paper evaluates some features of this new boron model in the TRACE code.

42 ENGINEERING↗
DOE OSTI2017

Verification of the Coupled Neutronics/Thermal-Hydraulics Code TANSY against the MSFR Benchmark

The MSRs is a liquid-fuel reactor in which uranium, fission products and actinides are dissolved in the molten salt. The molten salt reactors are featured by circulating fuel, low pressure and high temperature. The MSRs meets many of the future goals of nuclear technology, in particular for what concerns as improved sustainability and unique characteristics in terms of actinide burning. The fluid nature of fuel in MSRs gives extra flexibility in reactor design, fuel fabrication and recycling. But because of the application of fluid fuel, MSRs have several special characteristics compared with the traditional solid-fuel reactors. Firstly, the transport of the delayed neutron precursors by the fluid flow reduces the contribution of delayed neutrons to the chain reaction.. Secondly, the fact that the fuel is dissolved in the coolant rather than separated from the coolant by the claddings results in a much stronger coupling phenomenon between the neutronics and thermal-hydraulics. Thirdly, the shape of the fuel is determined by the container because of the characteristics of fluid, the unstructured meshes are required to model the complex geometry accurately. However most of the conventional reactor codes were developed for solid-fuel reactor and therefore not capable of taking liquid fuel flow into account. To address all these particular modeling issues of MSRs, a multi-physics code named TANSY has been developed in our laboratory. In this paper, the benchmark problem based on the MSFR design was calculated to verify the code.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2014

Use of the portable decay gamma source feature in the R2Smesh code system for the accurate estimation of photon radiation dose fields around activated ITER components - Paper 111

The R2Smesh ('Rigorous 2-step') code system, developed by KIT for the determination of shut-down dose rate distribution on mesh grid, can be utilized to provide photon radiation dose field around activated components which are moved from the irradiation site in fusion reactor to another external location. R2Smesh couples MCNP transport calculations and FISPACT inventory calculations through suitable interfaces utilizing MCNP mesh tally capability. Neutron flux and the decay gamma source are provided with high resolution on the mesh grid superimposed to the real geometry to take into account the spatial variations of the flux and the decay gamma source distributions. This feature enables the decay gamma source distribution to be transferred from the irradiation site in the reactor to any other external location for the determination of photon flux and dose rate distributions. In this work, we applied the feature of the portable decay gamma source in R2Smesh code system to a dose rate analysis for a tunnel connecting the ITER tokamak complex building (TCB) and the hot cell facility building (HCFB). This paper presents the use of the portability feature of the decay gamma source in R2Smesh interface code system developed by KIT for a dose rate analysis of the ITER tunnel connecting TCB and HCFB. This application shows that the decay gamma source of the activated components could be transferred successfully from one coordinate system to another coordinate system and superimposed to the geometry of the components. This is very useful when the activated components need to be transported from the reactor to somewhere else for maintenance, testing or storage and the dose rate analysis is requested. Another portable feature, which could convert the 3D decay gamma source distribution generated by R2Smesh system to the MCNP 'SDEF' card source description and avoid the use of source sampling subroutine, is under consideration. With this feature, one decay gamma source result for the specific reactor components could be shared by different institutions who need to use the same decay gamma source for different analysis. (authors)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2015

Inter-code comparison of TRIPOLI{sup R} And MVP on the MCNP criticality validation suite - 14421

This paper presents a joint work conducted at CEA Saclay and JAEA Tokai aimed at comparing the Monte Carlo codes TRIPOLI-4{sup R} and MVP on a selection of ICSBEP benchmarks. Our goal is to establish a common set of Monte Carlo input decks, as a basis for rigorous inter-code comparison in criticality-safety. As a reference, we will use the MCNP Criticality Validation Suite: other Monte Carlo developers might easily join this effort in the future. For the purpose of inter-code comparison, the TRIPOLI-4{sup R} and MVP input decks have been exactly translated from those of MCNP, without any further assumptions. Both TRIPOLI-4{sup R} and MVP have been run with the same ENDF/B-VII.0 evaluated nuclear data and, as far as possible, the same simulation options as in the original LANL work (same initial source, same number of active and discarded cycles and neutrons per cycle). (authors)

42 ENGINEERING↗
DOE OSTI2015

Quantification of calculation accuracy for code systems in burn-up credit applications by recalculations of experimental data - 14480

For application in criticality safety, numerical code systems require profound validation against experimental data. In terms of burn-up credit, this refers to the inventory determination as well as the subsequent criticality calculation. Within this framework, a dedicated, conservative estimation of uncertainties, bias and bias uncertainties for the resulting calculated multiplication factor k{sub eff} is indispensable. However, measures to reduce excess conservatism may become desirable, depending on the application case. For this reason GRS implemented a methodology to validate a calculation code system for criticality safety analyses for systems with irradiated nuclear fuel. The work at hand demonstrates this methodology by means of the GRS home-built inventory determination system KENOREST and ORNL's well-known SCALE 6 CSASS criticality safety calculation sequence. A number of 40 radiochemical analysis samples taken from SFCOMPO and elsewhere have been calculated to derive isotopic correction factors in different ways, depending on the required level of conservatism and the calculational efforts to be carried out. This includes bounding methods as well as different Monte Carlo sampling approaches. Furthermore, from the ICSBEP a number of 515 critical benchmark experiment configurations comprising water moderated low enriched UO{sub 2} fuel rods and water moderated mixed oxide (MOX) fuel rods have been selected and evaluated. Important absorber and structure materials are also covered. Various trending and sensitivity and uncertainty analysis methods have been applied, again to derive conservative estimates on the various contributions to k{sub eff} uncertainties. The overall methodology is applicable to low enriched, water moderated PWR uranium dioxide fuels in a burn-up range of about 10 to 60 GWd/tHM, and validates the reactivity contributions of the major uranium and plutonium isotopes. The methodology itself has no limitation which requires the exclusion of further isotopes as e.g. fission products. However, its application is constrained by the experimental database being publicly available. After the potential provision of an extended database in the future, the range of applicability can easily be extended to more nuclides. In addition, it is not limited to the code systems being used here but can also be applied using different numerical tools. For a generic PWR spent fuel pool rack and different enrichment and burn-up combinations, the impact of the current level of validation on k{sub eff} is demonstrated and quantified. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2024

The Impact of the Halden Reactor Project on the Development of the Bison Fuel Performance Code

The Bison fuel performance code has been under development at Idaho National Laboratory by various programs within the U.S. Department of Energy since 2009. As part of any fuel performance code development, validation is necessary. To this end, several Halden experiments have been used. Early on, the focus was on validating the thermal mechanical behaviour, including pellet cladding mechanical interaction and fission gas release models implemented in Bison. Support for design-basis accident capabilities began in approximately 2014. Also in 2014, the development of models for accident tolerant fuel (ATF) concepts also began. This includes dopedUO2 fuel for which Halden has some experiments with temperature and fission gas release measurements. The Bison team has been fortunate to have two senior developers be secondees to Halden for a year: the late Dr. Giovanni Pastore from May 2015 to April 2016 and the now retired Dr. Richard Williamson in 2018. The paper will provide examples of Bison validation using Halden data for standard UO2 fuel under normal operation and loss-of-coolant accident condition, and ATF concepts. A tribute to the legacy that both the late Giovanni and retired Richard left on the Bison code is provided.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2025

Status of the MARS code

This report describes major features of the most recent version of the MARS code as well as ongoing developments. The list of features includes various options for geometry models, a beam line builder based on MADX code, import of geometry models in GDML format, use of structured and unstructured meshes for scoring purposes, an update to the recent TENDL library for a number of projectiles at low energies (up to 250 MeV), and a recently implemented method to calculate spatial distribution of residual dose in a single computer run without an intermediate source. Examples of the code application to various projects are presented as well.

Rakhno, Igor [Fermilab] (ORCID:0000000265828058)↗
DOE OSTI2025

Using Calibrated Sodium Data for Preliminary Validation of the SRT Code for Advanced Reactors

Various types of non-light water reactors are currently engaged in the U.S. licensing process. Because of inherent differences compared with well-established large light water reactors, appropriate assessment tools are needed. Specifically, source term analysis, which determines environmental dose impacts from potential accident scenarios, is a crucial part of design and licensing. The U.S. Nuclear Regulatory Commission has emphasized the importance of mechanistic source term analysis for advanced reactor deployments. To align with these needs, Argonne National Laboratory has developed the Simplified Radionuclide Transport (SRT) source term analysis code for metal fuel Sodium-cooled Fast Reactors (SFRs) and microreactors. SRT conducts time-dependent radionuclide transport and retention in SFRs for core and ex-core radionuclide source accident sequences. The main objective of SRT is to provide rapid sensitivity and uncertainty analyses, incorporating parametric uncertainties and summarizing probabilistic results. As part of the code validation process, a study focused on the bubble scrubbing module was performed using an experiment recently carried out by the University of Wisconsin-Madison. Based on the analysis, the modeling approach in SRT provides accurate results for small and large aerosols, while slight underprediction of radionuclide aerosol removal are observed for medium sized aerosols. However, the deviation is minor, considering the highly uncertain phenomenon and range of results, and is in the conservative direction. In addition, uncertainty information derived from the experiments is further implemented, reflecting the actual span of parameters, which leads to enhanced agreement with code predictions. The results demonstrate that SRT provides reasonable predictions for the bubble scrubbing process in sodium pool.

Kam, Dong Hoon↗
DOE OSTI2023

Code for multi-shape Gaussian process (GP) fitting with uncertainty quantification (UQ)

This is code associated with the publication “Nonparametric Multi-shape Modeling with Uncertainty Quantification,” authored by Hengrui Luo (Lawrence Berkeley National Laboratory) and Justin Strait (Los Alamos National Laboratory). The code is used to fit multiple-output Gaussian process (GP) models of planar closed curves to collections of ordered point sets, allowing for flexible nonlinear prediction of the underlying curve under dense or sparse point set samplings and with or without noise, as well as tractable uncertainty quantification. To do this, we employ use of a periodic kernel to account for the nonlinear input space of closed curves, and combine with coregionalization models to account for dependence both (i) between curve coordinates, and (ii) between pairs of curves. Functions in the code are capable of fitting these models, as well as performing additional tasks with the fitted curves such as (a) shape registration and alignment, (b) shape averaging, and (c) fitting for curve sub-populations / clusters.

Strait, Justin↗
DOE OSTI2023

Python codes for the paper - "Trade-offs in the latent representation of microstructure evolution"

SAND2024-00946O Python codes used in "Trade-offs in the latent representation of microstructure evolution," a manuscript accepted for publication in Acta Materialia, are part of a repository. The code was developed to perform analysis of microstructure evolution. The repository consists of two main directories: models, which train and test models such as autoencoders and diffusion maps, and analysis, which analyzes microstructures. Source code is used to perform dimensionality reduction of microstructure data for analysis of its evolution in time. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology & Engineering Solutions of Sandia, LLC, a wholly owned subsidiary of Honeywell International Inc., for the U.S. Department of Energy’s National Nuclear Security Administration under contract DE-NA0003525.

Dingreville, Remi↗
DOE OSTI2017

Special issue: Advanced stencil-code engineering

Here, stencil codes are compute-intensive algorithms, in which data points arranged in a large grid are being recomputed repeatedly from the values of data points in a predefined neighborhood. This fixed neighborhood pattern is called a stencil. Stencils codes see wide-spread use in computing the discrete solutions of partial differential equations and systems composed of such equations.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2013

Spacecraft charging analysis with the implicit particle-in-cell code iPic3D

We present the first results on the analysis of spacecraft charging with the implicit particle-in-cell code iPic3D, designed for running on massively parallel supercomputers. The numerical algorithm is presented, highlighting the implementation of the electrostatic solver and the immersed boundary algorithm; the latter which creates the possibility to handle complex spacecraft geometries. As a first step in the verification process, a comparison is made between the floating potential obtained with iPic3D and with Orbital Motion Limited theory for a spherical particle in a uniform stationary plasma. Second, the numerical model is verified for a CubeSat benchmark by comparing simulation results with those of PTetra for space environment conditions with increasing levels of complexity. In particular, we consider spacecraft charging from plasma particle collection, photoelectron and secondary electron emission. The influence of a background magnetic field on the floating potential profile near the spacecraft is also considered. Although the numerical approaches in iPic3D and PTetra are rather different, good agreement is found between the two models, raising the level of confidence in both codes to predict and evaluate the complex plasma environment around spacecraft.

42 ENGINEERING↗
DOE OSTI2025

Development and validation of non-axisymmetric heat flux simulations with 3D fields using the HEAT code

A new comprehensive module to simulate heat fluxes from three-dimensional (3D) magnetic fields has been implemented in the HEAT code. Especially compact tokamaks like SPARC require tools to predict and manage large heat fluxes. Existing release versions of HEAT can only simulate axisymmetric heat flux on 3D plasma facing components. The new module uses an M3D-C1 perturbed equilibrium and the MAFOT code to trace field lines of the perturbed 3D magnetic field. Heat flux is then assigned to the resulting footprints via a 3D layer model. The model distinguishes between the scrape-off layer, the magnetic lobes and the private flux region, and employs only 0D parameters like the layer width, diffusive spread and the last closed flux surface position in the perturbed edge to generate a heat flux profile. The magnitude is normalized to the total input power. Resulting heat flux simulations are compared and validated against infrared measurements in the DIII-D tokamak with applied 3D fields; good agreement is found for several cases. The new module can now be applied to the SPARC tokamak; a preliminary result for applied rotating 3D fields is shown.

70 PLASMA PHYSICS AND FUSION TECHNOLOGY↗
DOE OSTI2017

Energy Savings Analysis of the Proposed Revision of the Washington D.C. Non-Residential Energy Code

This report presents the results of an assessment of savings for the proposed Washington D.C. energy code relative to ASHRAE Standard 90.1-2010. It includes annual and life cycle savings for site energy, source energy, energy cost, and carbon dioxide emissions that would result from adoption and enforcement of the proposed code for newly constructed buildings in Washington D.C. over a five year period.

Commercial Energy Code↗