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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 181 records · Page 10
All sourcesNASA NTRSDOE OSTI
DOE OSTI2019

Combining Observational and Computational Uncertainty in Calibration Experiments

Traditional computer experiments using emulators have treat the emulator output as containing no error; however, this is not always feasible, as there exist computational codes which produce stochastic output, and future computations may require a balance between the time or resources to run the code and the accuracy of the output. This paper proposes a framework for incorporating both computational uncertainty and observational uncertainty into an analysis using a Gaussian process regression and calibration, and provides an example using simulated output from the one-dimensional diffusion equation. The particular values of observational and computation uncertainty which minimize a cost function are obtained.

97 MATHEMATICS AND COMPUTING↗
DOE OSTI2016

Shaping of the axial power density distribution in the core to minimize the vapor volume fraction at the outlet of the VVER-1200 fuel assemblies

The possibility of decreasing the vapor fraction at the VVER-1200 fuel assembly outlet by shaping the axial power density field is considered. The power density field was shaped by axial redistribution of the concentration of the burnable gadolinium poison in the Gd-containing fuel rods. The mathematical modeling of the VVER-1200 core was performed using the NOSTRA computer code.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2016

Numerical simulation on pressure propagation in fuel-coolant interaction with MCBA-simple algorithm

A computer code named METRIC is developed to simulate multiphase interactions of molten fuel and coolant at pressure propagation stage, also to examine the model's ability to predict the physical process. The fluids modeled in current code are liquid coolant, vapor, non-condensable gas, molten drop and molten fragment. Conservation equations are solved by MCBA-SIMPLE method, which comes from the mass conservation-based algorithms (MCBA) group. The numerical methods and constitutive models in the code are described in this paper. Pressure wave propagation in two-phase medium is simulated to validate current two-phase model. Computational simulation of KROTOS 21 is carried out. Parameter sensitivity study is performed. Effects of void fraction, fragment size, and fragmentation model are discussed. Results of the simulations are physically reasonable and close to the experimental observations. It is suggested that the physical models still need improved to avoid the uncertainty of empirical parameters and user effect. (authors)

42 ENGINEERING↗
DOE OSTI2013

Linking RESRAD-OFFSITE and HYDROGEOCHEM Model for Performance Assessment of Low-Level Radioactive Waste Disposal Facility - 13429

Performance assessments are crucial steps for the long-term radiological safety requirements of low-level waste (LLW) disposal facility. How much concentration of radionuclides released from the near-field to biosphere and what radiation exposure levels of an individual can influence on the satisfactory performance of the LLW disposal facility and safety disposal environment. Performance assessment methodology for the radioactive waste disposal consists of the reactive transport modeling of safety-concerned radionuclides released from the near-field to the far-field, and the potential exposure pathways and the movements of radionuclides through the geosphere, biosphere and man of which the accompanying dose. Therefore, the integration of hydrogeochemical transport model and dose assessment code, HYDROGEOCHEM code and RESRAD family of codes is imperative. The RESRAD family of codes such as RESRAD-OFFSITE computer code can evaluate the radiological dose and excess cancer risk to an individual who is exposed while located within or outside the area of initial (primary) contamination. The HYDROGEOCHEM is a 3-D numerical model of fluid flow, thermal, hydrologic transport, and biogeochemical kinetic and equilibrium reactions in saturated and unsaturated media. The HYDROGEOCHEM model can also simulate the crucial geochemical mechanism, such as the effect of redox processes on the adsorption/desorption, hydrogeochemical influences on concrete degradation, adsorption/desorption of radionuclides (i.e., surface complexation model) between solid and liquid phase in geochemically dynamic environments. To investigate the safety assessment of LLW disposal facility, linking RESRAD-OFFSITE and HYDROGEOCHEM model can provide detailed tools of confidence in the protectiveness of the human health and environmental impact for safety assessment of LLW disposal facility. (authors)

12 MANAGEMENT OF RADIOACTIVE AND NON-RADIOACTIVE W↗
DOE OSTI2016

Comparison of HFIR Core Hydraulic Models with Operational Data

The core flow channels of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL) have been modeled with three different computer codes and calculated results have been compared with available plant operational data. The three codes employed are RELAP5-3D, TRACE, and COMSOL. This work was performed to develop consistent boundary conditions for three-dimensional (3-D) computational fluid dynamics (CFD) models of the HFIR core. The variables of interest for the CFD models include: (1) the inlet velocity and (2) outlet pressure for each coolant channel, and (3) the axial pressure difference between narrow and wide adjacent channels. (authors)

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2023

Modeling Codes Used for LAMP

LAMP team is planning on using multiple computer codes as tools for modeling beam dynamics, electromagnetics, and plasma environment in ion sources. Below is a description of the computer tools we have been using or planning to use in the future. Most of the codes chosen for the LAMP project are capable to model most of the beam dynamics in LAMP. A judicious choice of the appropriate code to use will be based on speed and fidelity of the modeling with faster calculations produced by simplifying some features.

43 PARTICLE ACCELERATORS↗
DOE OSTI2013

Application of adaptive hierarchical sparse grid collocation to the uncertainty quantification of nuclear reactor simulators

Recent efforts in the application of uncertainty quantification to nuclear systems have utilized methods based on generalized perturbation theory and stochastic sampling. While these methods have proven to be effective they both have major drawbacks that may impede further progress. A relatively new approach based on spectral elements for uncertainty quantification is applied in this paper to several problems in reactor simulation. Spectral methods based on collocation attempt to couple the approximation free nature of stochastic sampling methods with the determinism of generalized perturbation theory. The specific spectral method used in this paper employs both the Smolyak algorithm and adaptivity by using Newton-Cotes collocation points along with linear hat basis functions. Using this approach, a surrogate model for the outputs of a computer code is constructed hierarchically by adaptively refining the collocation grid until the interpolant is converged to a user-defined threshold. The method inherently fits into the framework of parallel computing and allows for the extraction of meaningful statistics and data that are not within reach of stochastic sampling and generalized perturbation theory. This paper aims to demonstrate the advantages of spectral methods-especially when compared to current methods used in reactor physics for uncertainty quantification-and to illustrate their full potential. (authors)

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Verification of the Time-dependent Transport Code TDKENO with the Monte Carlo Code KENO-VI

In this paper we focus on the computer code TDKENO that employs a hybrid methodology for solving the time dependent, three-dimensional Boltzmann transport equation. Originally, TDKENO used the Monte Carlo Transport code KENO V.a from SCALE to calculate flux shapes and fixed sources. KENO V.a is quick to compute these quantities but is limited to simple geometries. For this work TDKENO has been modified to support the Monte Carlo Transport code KENO-VI in addition to KENO V.a. KENO-VI allows generalized geometry input not possible in KENO V.a inputs. In order to verify the changes we ran TDKENO using KENO V.a and KENO-VI inputs then compared the results. Three different inputs were analyzed, the 16a1 benchmark,a simple TREAT model, and a transient model (2855) from the M8 calibration tests performed at INL.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗
DOE OSTI2017

An efficient hybrid technique in RCS predictions of complex targets at high frequencies

Most computer codes in Radar Cross Section (RCS) prediction use Physical Optics (PO) and Physical theory of Diffraction (PTD) combined with Geometrical Optics (GO) and Geometrical Theory of Diffraction (GTD). The latter approaches are computationally cheaper and much more accurate for curved surfaces, but not applicable for the computation of the RCS of all surfaces of a complex object due to the presence of caustic problems in the analysis of concave surfaces or flat surfaces in the far field. The main contribution of this paper is the development of a hybrid method based on a new combination of two asymptotic techniques: GTD and PO, considering the advantages and avoiding the disadvantages of each of them. A very efficient and accurate method to analyze the RCS of complex structures at high frequencies is obtained with the new combination. The proposed new method has been validated comparing RCS results obtained for some simple cases using the proposed approach and RCS using the rigorous technique of Method of Moments (MoM). Some complex cases have been examined at high frequencies contrasting the results with PO. This study shows the accuracy and the efficiency of the hybrid method and its suitability for the computation of the RCS at really large and complex targets at high frequencies.

71 CLASSICAL AND QUANTUM MECHANICS, GENERAL PHYSIC↗
DOE OSTI2016

Using Python For 3D Visualizations of MCNP output

MCNP is ubiquitously used within industries, universities and national laboratories as a tool for accurately modeling radiation. With so much focus on the details of transport, attention is seldom given to the visualizations of these meticulously constructed outputs. Python is a programing language designed to promote code readability and has steadily grown in popularity since the 1990's. Python is also a heavily crowd-sourced programing language which generally means that the problems that you are trying to solve with computer codes are already solved; The user is often left with the now-simplified task of cobbling together a good and useful compilation of these tools and adapting them to solve the problem at hand. I have taken advantage of this situation and constructed some visualization tools in python that can be used to duplicate some of the 2D radiation fields that MCNP produces and also produce a meaningful 3D representation of a radiation field. (author)

61 RADIATION PROTECTION AND DOSIMETRY↗
DOE OSTI2016

FY2017 Pilot Project Plan for the Nuclear Energy Knowledge and Validation Center Initiative

To prepare for technical development of computational code validation under the Nuclear Energy Knowledge and Validation Center (NEKVAC) initiative, several meetings were held by a group of experts of the Idaho National Laboratory (INL) and the Oak Ridge National Laboratory (ORNL) to develop requirements of, and formulate a structure for, a transient fuel database through leveraging existing resources. It was concluded in discussions of these meetings that a pilot project is needed to address the most fundamental issues that can generate immediate stimulus to near-future validation developments as well as long-lasting benefits to NEKVAC operation. The present project is proposed based on the consensus of these discussions. Analysis of common scenarios in code validation indicates that the incapability of acquiring satisfactory validation data is often a showstopper that must first be tackled before any confident validation developments can be carried out. Validation data are usually found scattered in different places most likely with interrelationships among the data not well documented, incomplete with information for some parameters missing, nonexistent, or unrealistic to experimentally generate. Furthermore, with very different technical backgrounds, the modeler, the experimentalist, and the knowledgebase developer that must be involved in validation data development often cannot communicate effectively without a data package template that is representative of the data structure for the information domain of interest to the desired code validation. This pilot project is proposed to use the legendary TREAT Experiments Database to provide core elements for creating an ideal validation data package. Data gaps and missing data interrelationships will be identified from these core elements. All the identified missing elements will then be filled in with experimental data if available from other existing sources or with dummy data if nonexistent. The resulting hybrid validation data package (composed of experimental and dummy data) will provide a clear and complete instance delineating the structure of the desired validation data and enabling effective communication among the modeler, the experimentalist, and the knowledgebase developer. With a good common understanding of the desired data structure by the three parties of subject matter experts, further existing data hunting will be effectively conducted, new experimental data generation will be realistically pursued, knowledgebase schema will be practically designed; and code validation will be confidently planned.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2016

Criticality Calculations with MCNP6 - Practical Lectures

These slides are used to teach MCNP (Monte Carlo N-Particle) usage to nuclear criticality safety analysts. The following are the lecture topics: course information, introduction, MCNP basics, criticality calculations, advanced geometry, tallies, adjoint-weighted tallies and sensitivities, physics and nuclear data, parameter studies, NCS validation I, NCS validation II, NCS validation III, case study 1 - solution tanks, case study 2 - fuel vault, case study 3 - B&W core, case study 4 - simple TRIGA, case study 5 - fissile mat. vault, criticality accident alarm systems. After completion of this course, you should be able to: Develop an input model for MCNP; Describe how cross section data impact Monte Carlo and deterministic codes; Describe the importance of validation of computer codes and how it is accomplished; Describe the methodology supporting Monte Carlo codes and deterministic codes; Describe pitfalls of Monte Carlo calculations; Discuss the strengths and weaknesses of Monte Carlo and Discrete Ordinants codes; The diffusion theory model is not strictly valid for treating fissile systems in which neutron absorption, voids, and/or material boundaries are present. In the context of these limitations, identify a fissile system for which a diffusion theory solution would be adequate.

22 GENERAL STUDIES OF NUCLEAR REACTORS↗
DOE OSTI2014

RDS - A systematic approach towards system thermal hydraulics input code development for a comprehensive deterministic safety analysis

Deterministic Safety Analysis (DSA) is one of the mandatory requirements conducted for Nuclear Power Plant licensing process, with the aim of ensuring safety compliance with relevant regulatory acceptance criteria. DSA is a technique whereby a set of conservative deterministic rules and requirements are applied for the design and operation of facilities or activities. Computer codes are normally used to assist in performing all required analysis under DSA. To ensure a comprehensive analysis, the conduct of DSA should follow a systematic approach. One of the methodologies proposed is the Standardized and Consolidated Reference Experimental (and Calculated) Database (SCRED) developed by University of Pisa. Based on this methodology, the use of Reference Data Set (RDS) as a pre-requisite reference document for developing input nodalization was proposed. This paper shall describe the application of RDS with the purpose of assessing its effectiveness. Two RDS documents were developed for an Integral Test Facility of LOBI-MOD2 and associated Test A1-83. Data and information from various reports and drawings were referred in preparing the RDS. The results showed that by developing RDS, it has made possible to consolidate all relevant information in one single document. This is beneficial as it enables preservation of information, promotes quality assurance, allows traceability, facilitates continuous improvement, promotes solving of contradictions and finally assisting in developing thermal hydraulic input regardless of whichever code selected. However, some disadvantages were also recognized such as the need for experience in making engineering judgments, language barrier in accessing foreign information and limitation of resources. Some possible improvements are suggested to overcome these challenges.

42 ENGINEERING↗
DOE OSTI2016

Impact of Concrete Composition on Criticality at Nuclear Power Plants

There is no standard concrete composition for criticality analysis. Concrete is a composite material composed of aggregate bonded together with cement which hardens over time. Aggregate consists of large chunks of material in a concrete mix, generally coarse gravel or crushed rocks such as limestone or granite, along with finer materials such as sand. Water is mixed with the dry powder cement/aggregate blend. The concrete solidifies and hardens. After the concrete is hardened there can be water loss with time. The main components in concrete generally have small neutron cross sections so the absorption cross section is often dominated by the impurities. The content of the impurities are rarely documented in a manner that is generally needed for crediting in a criticality analysis. This paper assesses the impact of conservative concrete on a sample dry storage rack and a spent fuel pool. The SCALE computer code has four types of concrete in the Standard Composition Library.

36 MATERIALS SCIENCE↗
DOE OSTI2016

Passive Safety Feature of KIPT Neutron Source Facility

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have cooperated on the development, design, and construction of a neutron source facility. The facility was constructed at Kharkov, Ukraine and its commissioning process is underway. The facility has an electron accelerator driven subcritical assembly. The electron beam power is 100 kW using 100 MeV electrons. Tungsten or natural uranium is the target material for generating neutrons driving the subcritical assembly. The subcritical assembly is composed of WWR-M2 Russian fuel assemblies with U-235 enrichment of 19.7 wt%, surrounded by beryllium reflector assembles and graphite blocks. The water coolant operates at room temperature during normal operation, when the facility is operating at full power using 100-KW electron beam power. The passive safety features of the facility are discussed in in this study. Monte Carlo computer code MCNPX was utilized in the analyses with ENDF/B VII.0 nuclear data library.

11 NUCLEAR FUEL CYCLE AND FUEL MATERIALS↗
DOE OSTI2016

University of Arizona High Energy Physics Program at the Cosmic Frontier 2014-2016

This is the final technical report from the University of Arizona High Energy Physics program at the Cosmic Frontier covering the period 2014-2016. The work aims to advance the understanding of dark energy using the Large Synoptic Survey Telescope (LSST). Progress on the engineering design of the power supplies for the LSST camera is discussed. A variety of contributions to photometric redshift measurement uncertainties were studied. The effect of the intergalactic medium on the photometric redshift of very distant galaxies was evaluated. Computer code was developed realizing the full chain of calculations needed to accurately and efficiently run large-scale simulations.

79 ASTRONOMY AND ASTROPHYSICS↗
DOE OSTI2017

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of the of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and deposition, we are able to reasonably capture the deposition of radionuclides to the northwest of the reactor site.

21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS↗