THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1984-07-01

The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures inmore » the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.« less

PRATHAM: Parallel Thermal Hydraulics Simulations using Advanced Mesoscopic Methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Joshi, Abhijit S; Jain, Prashant K; Mudrich, Jaime A

2012-01-01

At the Oak Ridge National Laboratory, efforts are under way to develop a 3D, parallel LBM code called PRATHAM (PaRAllel Thermal Hydraulic simulations using Advanced Mesoscopic Methods) to demonstrate the accuracy and scalability of LBM for turbulent flow simulations in nuclear applications. The code has been developed using FORTRAN-90, and parallelized using the message passing interface MPI library. Silo library is used to compact and write the data files, and VisIt visualization software is used to post-process the simulation data in parallel. Both the single relaxation time (SRT) and multi relaxation time (MRT) LBM schemes have been implemented in PRATHAM.more » To capture turbulence without prohibitively increasing the grid resolution requirements, an LES approach [5] is adopted allowing large scale eddies to be numerically resolved while modeling the smaller (subgrid) eddies. In this work, a Smagorinsky model has been used, which modifies the fluid viscosity by an additional eddy viscosity depending on the magnitude of the rate-of-strain tensor. In LBM, this is achieved by locally varying the relaxation time of the fluid.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Alfonsi; C. Rabiti; D. Mandelli

The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data miningmore » module« less

An analytical study on excitation of nuclear-coupled thermal-hydraulic instability due to seismically induced resonance in BWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hirano, Masashi

1997-07-01

This paper describes the results of a scoping study on seismically induced resonance of nuclear-coupled thermal-hydraulic instability in BWRs, which was conducted by using TRAC-BF1 within a framework of a point kinetics model. As a result of the analysis, it is shown that a reactivity insertion could occur accompanied by in-surge of coolant into the core resulted from the excitation of the nuclear-coupled instability by the external acceleration. In order to analyze this phenomenon more in detail, it is necessary to couple a thermal-hydraulic code with a three-dimensional nuclear kinetics code.

Dynamic Stability of the Rate, State, Temperature, and Pore Pressure Friction Model at a Rock Interface

NASA Astrophysics Data System (ADS)

Sinha, Nitish; Singh, Arun K.; Singh, Trilok N.

2018-05-01

In this article, we study numerically the dynamic stability of the rate, state, temperature, and pore pressure friction (RSTPF) model at a rock interface using standard spring-mass sliding system. This particular friction model is a basically modified form of the previously studied friction model namely the rate, state, and temperature friction (RSTF). The RSTPF takes into account the role of thermal pressurization including dilatancy and permeability of the pore fluid due to shear heating at the slip interface. The linear stability analysis shows that the critical stiffness, at which the sliding becomes stable to unstable or vice versa, increases with the coefficient of thermal pressurization. Critical stiffness, on the other hand, remains constant for small values of either dilatancy factor or hydraulic diffusivity, but the same decreases as their values are increased further from dilatancy factor (˜ 10^{ - 4} ) and hydraulic diffusivity (˜ 10^{ - 9} {m}2 {s}^{ - 1} ) . Moreover, steady-state friction is independent of the coefficient of thermal pressurization, hydraulic diffusivity, and dilatancy factor. The proposed model is also used for predicting time of failure of a creeping interface of a rock slope under the constant gravitational force. It is observed that time of failure decreases with increase in coefficient of thermal pressurization and hydraulic diffusivity, but the dilatancy factor delays the failure of the rock fault under the condition of heat accumulation at the creeping interface. Moreover, stiffness of the rock-mass also stabilizes the failure process of the interface as the strain energy due to the gravitational force accumulates in the rock-mass before it transfers to the sliding interface. Practical implications of the present study are also discussed.

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE PAGES

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

2017-09-01

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

COBRA-SFS thermal-hydraulic analysis code for spent fuel storage and transportation casks: Models and methods

DOE Office of Scientific and Technical Information (OSTI.GOV)

Michener, Thomas E.; Rector, David R.; Cuta, Judith M.

COBRA-SFS, a thermal-hydraulics code developed for steady-state and transient analysis of multi-assembly spent-fuel storage and transportation systems, has been incorporated into the Used Nuclear Fuel-Storage, Transportation and Disposal Analysis Resource and Data System tool as a module devoted to spent fuel package thermal analysis. This paper summarizes the basic formulation of the equations and models used in the COBRA-SFS code, showing that COBRA-SFS fully captures the important physical behavior governing the thermal performance of spent fuel storage systems, with internal and external natural convection flow patterns, and heat transfer by convection, conduction, and thermal radiation. Of particular significance is themore » capability for detailed thermal radiation modeling within the fuel rod array.« less

VERA and VERA-EDU 3.5 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sieger, Matt; Salko, Robert K.; Kochunas, Brendan M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms. Neutronics analysis can be performed for 2D lattices, 2D core and 3D core problems for pressurized water reactor geometries that can be used to calculate criticality and fission rate distributions by pin for input fuel compositions. MPACT uses the Method of Characteristics transport approach for 2D problems.more » For 3D problems, MPACT uses the 2D/1D method which uses 2D MOC in a radial plane and diffusion or SPn in the axial direction. MPACT includes integrated cross section capabilities that provide problem-specific cross sections generated using the subgroup methodology. The code can be executed both 2D and 3D problems in parallel to reduce overall run time. A thermal-hydraulics capability is provided with CTF (an updated version of COBRA-TF) that allows thermal-hydraulics analyses for single and multiple assemblies using the simplified VERA common input. This distribution also includes coupled neutronics/thermal-hydraulics capabilities to allow calculations using MPACT coupled with CTF. The VERA fuel rod performance component BISON calculates, on a 2D or 3D basis, fuel rod temperature, fuel rod internal pressure, free gas volume, clad integrity and fuel rod waterside diameter. These capabilities allow simulation of power cycling, fuel conditioning and deconditioning, high burnup performance, power uprate scoping studies, and accident performance. Input/Output capabilities include the VERA Common Input (VERAIn) script which converts the ASCII common input file to the intermediate XML used to drive all of the physics codes in the VERA Core Simulator (VERA-CS). VERA component codes either input the VERA XML format directly, or provide a preprocessor which can convert the XML into native input. VERAView is an interactive graphical interface for the visualization and engineering analyses of output data from VERA. The python-based software is easy to install and intuitive to use, and provides instantaneous 2D and 3D images, 1D plots, and alpha-numeric data from VERA multi-physics simulations. Testing within CASL has focused primarily on Westinghouse four-loop reactor geometries and conditions with example problems included in the distribution.« less

Assessment of uncertainties of the models used in thermal-hydraulic computer codes

NASA Astrophysics Data System (ADS)

Gricay, A. S.; Migrov, Yu. A.

2015-09-01

The article deals with matters concerned with the problem of determining the statistical characteristics of variable parameters (the variation range and distribution law) in analyzing the uncertainty and sensitivity of calculation results to uncertainty in input data. A comparative analysis of modern approaches to uncertainty in input data is presented. The need to develop an alternative method for estimating the uncertainty of model parameters used in thermal-hydraulic computer codes, in particular, in the closing correlations of the loop thermal hydraulics block, is shown. Such a method shall feature the minimal degree of subjectivism and must be based on objective quantitative assessment criteria. The method includes three sequential stages: selecting experimental data satisfying the specified criteria, identifying the key closing correlation using a sensitivity analysis, and carrying out case calculations followed by statistical processing of the results. By using the method, one can estimate the uncertainty range of a variable parameter and establish its distribution law in the above-mentioned range provided that the experimental information is sufficiently representative. Practical application of the method is demonstrated taking as an example the problem of estimating the uncertainty of a parameter appearing in the model describing transition to post-burnout heat transfer that is used in the thermal-hydraulic computer code KORSAR. The performed study revealed the need to narrow the previously established uncertainty range of this parameter and to replace the uniform distribution law in the above-mentioned range by the Gaussian distribution law. The proposed method can be applied to different thermal-hydraulic computer codes. In some cases, application of the method can make it possible to achieve a smaller degree of conservatism in the expert estimates of uncertainties pertinent to the model parameters used in computer codes.

Interface design of VSOP'94 computer code for safety analysis

NASA Astrophysics Data System (ADS)

Natsir, Khairina; Yazid, Putranto Ilham; Andiwijayakusuma, D.; Wahanani, Nursinta Adi

2014-09-01

Today, most software applications, also in the nuclear field, come with a graphical user interface. VSOP'94 (Very Superior Old Program), was designed to simplify the process of performing reactor simulation. VSOP is a integrated code system to simulate the life history of a nuclear reactor that is devoted in education and research. One advantage of VSOP program is its ability to calculate the neutron spectrum estimation, fuel cycle, 2-D diffusion, resonance integral, estimation of reactors fuel costs, and integrated thermal hydraulics. VSOP also can be used to comparative studies and simulation of reactor safety. However, existing VSOP is a conventional program, which was developed using Fortran 65 and have several problems in using it, for example, it is only operated on Dec Alpha mainframe platforms and provide text-based output, difficult to use, especially in data preparation and interpretation of results. We develop a GUI-VSOP, which is an interface program to facilitate the preparation of data, run the VSOP code and read the results in a more user friendly way and useable on the Personal 'Computer (PC). Modifications include the development of interfaces on preprocessing, processing and postprocessing. GUI-based interface for preprocessing aims to provide a convenience way in preparing data. Processing interface is intended to provide convenience in configuring input files and libraries and do compiling VSOP code. Postprocessing interface designed to visualized the VSOP output in table and graphic forms. GUI-VSOP expected to be useful to simplify and speed up the process and analysis of safety aspects.

Coupled Monte Carlo neutronics and thermal hydraulics for power reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bernnat, W.; Buck, M.; Mattes, M.

The availability of high performance computing resources enables more and more the use of detailed Monte Carlo models even for full core power reactors. The detailed structure of the core can be described by lattices, modeled by so-called repeated structures e.g. in Monte Carlo codes such as MCNP5 or MCNPX. For cores with mainly uniform material compositions, fuel and moderator temperatures, there is no problem in constructing core models. However, when the material composition and the temperatures vary strongly a huge number of different material cells must be described which complicate the input and in many cases exceed code ormore » memory limits. The second problem arises with the preparation of corresponding temperature dependent cross sections and thermal scattering laws. Only if these problems can be solved, a realistic coupling of Monte Carlo neutronics with an appropriate thermal-hydraulics model is possible. In this paper a method for the treatment of detailed material and temperature distributions in MCNP5 is described based on user-specified internal functions which assign distinct elements of the core cells to material specifications (e.g. water density) and temperatures from a thermal-hydraulics code. The core grid itself can be described with a uniform material specification. The temperature dependency of cross sections and thermal neutron scattering laws is taken into account by interpolation, requiring only a limited number of data sets generated for different temperatures. Applications will be shown for the stationary part of the Purdue PWR benchmark using ATHLET for thermal- hydraulics and for a generic Modular High Temperature reactor using THERMIX for thermal- hydraulics. (authors)« less

Verification of combined thermal-hydraulic and heat conduction analysis code FLOWNET/TRUMP

NASA Astrophysics Data System (ADS)

Maruyama, Soh; Fujimoto, Nozomu; Kiso, Yoshihiro; Murakami, Tomoyuki; Sudo, Yukio

1988-09-01

This report presents the verification results of the combined thermal-hydraulic and heat conduction analysis code, FLOWNET/TRUMP which has been utilized for the core thermal hydraulic design, especially for the analysis of flow distribution among fuel block coolant channels, the determination of thermal boundary conditions for fuel block stress analysis and the estimation of fuel temperature in the case of fuel block coolant channel blockage accident in the design of the High Temperature Engineering Test Reactor(HTTR), which the Japan Atomic Energy Research Institute has been planning to construct in order to establish basic technologies for future advanced very high temperature gas-cooled reactors and to be served as an irradiation test reactor for promotion of innovative high temperature new frontier technologies. The verification of the code was done through the comparison between the analytical results and experimental results of the Helium Engineering Demonstration Loop Multi-channel Test Section(HENDEL T(sub 1-M)) with simulated fuel rods and fuel blocks.

IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yamada, K.; Aksan, S. N.

The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present,more » 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)« less

Leap Frog and Time Step Sub-Cycle Scheme for Coupled Neutronics and Thermal-Hydraulic Codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, S.

2002-07-01

As the result of the advancing TCP/IP based inter-process communication technology, more and more legacy thermal-hydraulic codes have been coupled with neutronics codes to provide best-estimate capabilities for reactivity related reactor transient analysis. Most of the coupling schemes are based on closely coupled serial or parallel approaches. Therefore, the execution of the coupled codes usually requires significant CPU time, when a complicated system is analyzed. Leap Frog scheme has been used to reduce the run time. The extent of the decoupling is usually determined based on a trial and error process for a specific analysis. It is the intent ofmore » this paper to develop a set of general criteria, which can be used to invoke the automatic Leap Frog algorithm. The algorithm will not only provide the run time reduction but also preserve the accuracy. The criteria will also serve as the base of an automatic time step sub-cycle scheme when a sudden reactivity change is introduced and the thermal-hydraulic code is marching with a relatively large time step. (authors)« less

Numerical analysis of one-dimensional temperature data for groundwater/surface-water exchange with 1DTempPro

NASA Astrophysics Data System (ADS)

Voytek, E. B.; Drenkelfuss, A.; Day-Lewis, F. D.; Healy, R. W.; Lane, J. W.; Werkema, D. D.

2012-12-01

Temperature is a naturally occurring tracer, which can be exploited to infer the movement of water through the vadose and saturated zones, as well as the exchange of water between aquifers and surface-water bodies, such as estuaries, lakes, and streams. One-dimensional (1D) vertical temperature profiles commonly show thermal amplitude attenuation and increasing phase lag of diurnal or seasonal temperature variations with propagation into the subsurface. This behavior is described by the heat-transport equation (i.e., the convection-conduction-dispersion equation), which can be solved analytically in 1D under certain simplifying assumptions (e.g., sinusoidal or steady-state boundary conditions and homogeneous hydraulic and thermal properties). Analysis of 1D temperature profiles using analytical models provides estimates of vertical groundwater/surface-water exchange. The utility of these estimates can be diminished when the model assumptions are violated, as is common in field applications. Alternatively, analysis of 1D temperature profiles using numerical models allows for consideration of more complex and realistic boundary conditions. However, such analyses commonly require model calibration and the development of input files for finite-difference or finite-element codes. To address the calibration and input file requirements, a new computer program, 1DTempPro, is presented that facilitates numerical analysis of vertical 1D temperature profiles. 1DTempPro is a graphical user interface (GUI) to the USGS code VS2DH, which numerically solves the flow- and heat-transport equations. Pre- and post-processor features within 1DTempPro allow the user to calibrate VS2DH models to estimate groundwater/surface-water exchange and hydraulic conductivity in cases where hydraulic head is known. This approach improves groundwater/ surface-water exchange-rate estimates for real-world data with complexities ill-suited for examination with analytical methods. Additionally, the code allows for time-varying temperature and hydraulic boundary conditions. Here, we present the approach and include examples for several datasets from stream/aquifer systems.

An assessment of the CORCON-MOD3 code. Part 1: Thermal-hydraulic calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strizhov, V.; Kanukova, V.; Vinogradova, T.

1996-09-01

This report deals with the subject of CORCON-Mod3 code validation (thermal-hydraulic modeling capability only) based on MCCI (molten core concrete interaction) experiments conducted under different programs in the past decade. Thermal-hydraulic calculations (i.e., concrete ablation, melt temperature, melt energy, concrete temperature, and condensible and non-condensible gas generation) were performed with the code, and compared with the data from 15 experiments, conducted at different scales using both simulant (metallic and oxidic) and prototypic melt materials, using different concrete types, and with and without an overlying water pool. Sensitivity studies were performed in a few cases involving, for example, heat transfer frommore » melt to concrete, condensed phase chemistry, etc. Further, special analysis was performed using the ACE L8 experimental data to illustrate the differences between the experimental and the reactor conditions, and to demonstrate that with proper corrections made to the code, the calculated results were in better agreement with the experimental data. Generally, in the case of dry cavity and metallic melts, CORCON-Mod3 thermal-hydraulic calculations were in good agreement with the test data. For oxidic melts in a dry cavity, uncertainties in heat transfer models played an important role for two melt configurations--a stratified geometry with segregated metal and oxide layers, and a heterogeneous mixture. Some discrepancies in the gas release data were noted in a few cases.« less

SMITHERS: An object-oriented modular mapping methodology for MCNP-based neutronic–thermal hydraulic multiphysics

DOE PAGES

Richard, Joshua; Galloway, Jack; Fensin, Michael; ...

2015-04-04

A novel object-oriented modular mapping methodology for externally coupled neutronics–thermal hydraulics multiphysics simulations was developed. The Simulator using MCNP with Integrated Thermal-Hydraulics for Exploratory Reactor Studies (SMITHERS) code performs on-the-fly mapping of material-wise power distribution tallies implemented by MCNP-based neutron transport/depletion solvers for use in estimating coolant temperature and density distributions with a separate thermal-hydraulic solver. The key development of SMITHERS is that it reconstructs the hierarchical geometry structure of the material-wise power generation tallies from the depletion solver automatically, with only a modicum of additional information required from the user. In addition, it performs the basis mapping from themore » combinatorial geometry of the depletion solver to the required geometry of the thermal-hydraulic solver in a generalizable manner, such that it can transparently accommodate varying levels of thermal-hydraulic solver geometric fidelity, from the nodal geometry of multi-channel analysis solvers to the pin-cell level of discretization for sub-channel analysis solvers.« less

Sensitivity analysis of hydraulic and thermal parameters inducing anomalous heat flow in the Lower Yarmouk Gorge

NASA Astrophysics Data System (ADS)

Goretzki, Nora; Inbar, Nimrod; Kühn, Michael; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Siebert, Christian; Magri, Fabien

2016-04-01

The Lower Yarmouk Gorge, at the border between Israel and Jordan, is characterized by an anomalous temperature gradient of 46 °C/km. Numerical simulations of thermally-driven flow show that ascending thermal waters are the result of mixed convection, i.e. the interaction between the regional flow from the surrounding heights and buoyant flow within permeable faults [1]. Those models were calibrated against available temperature logs by running several forward problems (FP), with a classic "trial and error" method. In the present study, inverse problems (IP) are applied to find alternative parameter distributions that also lead to the observed thermal anomalies. The investigated physical parameters are hydraulic conductivity and thermal conductivity. To solve the IP, the PEST® code [2] is applied via the graphical interface FEPEST® in FEFLOW® [3]. The results show that both hydraulic and thermal conductivity are consistent with the values determined with the trial and error calibrations, which precede this study. However, the IP indicates that the hydraulic conductivity of the Senonian Paleocene aquitard can be 8.54*10-3 m/d, which is three times lower than the originally estimated value in [1]. Moreover, the IP suggests that the hydraulic conductivity in the faults can increase locally up to 0.17 m/d. These highly permeable areas can be interpreted as local damage zones at the faults/units intersections. They can act as lateral pathways in the deep aquifers that allow deep outflow of thermal water. This presentation provides an example about the application of FP and IP to infer a wide range of parameter values that reproduce observed environmental issues. [1] Magri F, Inbar N, Siebert C, Rosenthal E, Guttman J, Möller P (2015) Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520, 342-355 [2] Doherty J (2010) PEST: Model-Independent Parameter Estimation. user manual 5th Edition. Watermark, Brisbane, Australia [3] Diersch H.-J.G. (2014) FEFLOW Finite Element Modeling of Flow, Mass and Heat Transport in Porous and Fractured Media. Springer- Verlag Berlin Heidelberg, 996p

Performance of a parallel thermal-hydraulics code TEMPEST

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fann, G.I.; Trent, D.S.

The authors describe the parallelization of the Tempest thermal-hydraulics code. The serial version of this code is used for production quality 3-D thermal-hydraulics simulations. Good speedup was obtained with a parallel diagonally preconditioned BiCGStab non-symmetric linear solver, using a spatial domain decomposition approach for the semi-iterative pressure-based and mass-conserved algorithm. The test case used here to illustrate the performance of the BiCGStab solver is a 3-D natural convection problem modeled using finite volume discretization in cylindrical coordinates. The BiCGStab solver replaced the LSOR-ADI method for solving the pressure equation in TEMPEST. BiCGStab also solves the coupled thermal energy equation. Scalingmore » performance of 3 problem sizes (221220 nodes, 358120 nodes, and 701220 nodes) are presented. These problems were run on 2 different parallel machines: IBM-SP and SGI PowerChallenge. The largest problem attains a speedup of 68 on an 128 processor IBM-SP. In real terms, this is over 34 times faster than the fastest serial production time using the LSOR-ADI solver.« less

Validation of CESAR Thermal-hydraulic Module of ASTEC V1.2 Code on BETHSY Experiments

NASA Astrophysics Data System (ADS)

Tregoures, Nicolas; Bandini, Giacomino; Foucher, Laurent; Fleurot, Joëlle; Meloni, Paride

The ASTEC V1 system code is being jointly developed by the French Institut de Radioprotection et Sûreté Nucléaire (IRSN) and the German Gesellschaft für Anlagen und ReaktorSicherheit (GRS) to address severe accident sequences in a nuclear power plant. Thermal-hydraulics in primary and secondary system is addressed by the CESAR module. The aim of this paper is to present the validation of the CESAR module, from the ASTEC V1.2 version, on the basis of well instrumented and qualified integral experiments carried out in the BETHSY facility (CEA, France), which simulates a French 900 MWe PWR reactor. Three tests have been thoroughly investigated with CESAR: the loss of coolant 9.1b test (OECD ISP N° 27), the loss of feedwater 5.2e test, and the multiple steam generator tube rupture 4.3b test. In the present paper, the results of the code for the three analyzed tests are presented in comparison with the experimental data. The thermal-hydraulic behavior of the BETHSY facility during the transient phase is well reproduced by CESAR: the occurrence of major events and the time evolution of main thermal-hydraulic parameters of both primary and secondary circuits are well predicted.

Burner liner thermal/structural load modeling: TRANCITS program user's manual

NASA Technical Reports Server (NTRS)

Maffeo, R.

1985-01-01

Transfer Analysis Code to Interface Thermal/Structural Problems (TRANCITS) is discussed. The TRANCITS code satisfies all the objectives for transferring thermal data between heat transfer and structural models of combustor liners and it can be used as a generic thermal translator between heat transfer and stress models of any component, regardless of the geometry. The TRANCITS can accurately and efficiently convert the temperature distributions predicted by the heat transfer programs to those required by the stress codes. It can be used for both linear and nonlinear structural codes and can produce nodal temperatures, elemental centroid temperatures, or elemental Gauss point temperatures. The thermal output of both the MARC and SINDA heat transfer codes can be interfaced directly with TRANCITS, and it will automatically produce stress model codes formatted for NASTRAN and MARC. Any thermal program and structural program can be interfaced by using the neutral input and output forms supported by TRANCITS.

Transient thermal, hydraulic, and mechanical analysis of a counter flow offset strip fin intermediate heat exchanger using an effective porous media approach

NASA Astrophysics Data System (ADS)

Urquiza, Eugenio

This work presents a comprehensive thermal hydraulic analysis of a compact heat exchanger using offset strip fins. The thermal hydraulics analysis in this work is followed by a finite element analysis (FEA) to predict the mechanical stresses experienced by an intermediate heat exchanger (IHX) during steady-state operation and selected flow transients. In particular, the scenario analyzed involves a gas-to-liquid IHX operating between high pressure helium and liquid or molten salt. In order to estimate the stresses in compact heat exchangers a comprehensive thermal and hydraulic analysis is needed. Compact heat exchangers require very small flow channels and fins to achieve high heat transfer rates and thermal effectiveness. However, studying such small features computationally contributes little to the understanding of component level phenomena and requires prohibitive computational effort using computational fluid dynamics (CFD). To address this issue, the analysis developed here uses an effective porous media (EPM) approach; this greatly reduces the computation time and produces results with the appropriate resolution [1]. This EPM fluid dynamics and heat transfer computational code has been named the Compact Heat Exchanger Explicit Thermal and Hydraulics (CHEETAH) code. CHEETAH solves for the two-dimensional steady-state and transient temperature and flow distributions in the IHX including the complicating effects of temperature-dependent fluid thermo-physical properties. Temperature- and pressure-dependent fluid properties are evaluated by CHEETAH and the thermal effectiveness of the IHX is also calculated. Furthermore, the temperature distribution can then be imported into a finite element analysis (FEA) code for mechanical stress analysis using the EPM methods developed earlier by the University of California, Berkeley, for global and local stress analysis [2]. These simulation tools will also allow the heat exchanger design to be improved through an iterative design process which will lead to a design with a reduced pressure drop, increased thermal effectiveness, and improved mechanical performance as it relates to creep deformation and transient thermal stresses.

Thermal-hydraulics Analysis of a Radioisotope-powered Mars Hopper Propulsion System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Robert C. O'Brien; Andrew C. Klein; William T. Taitano

Thermal-hydraulics analyses results produced using a combined suite of computational design and analysis codes are presented for the preliminary design of a concept Radioisotope Thermal Rocket (RTR) propulsion system. Modeling of the transient heating and steady state temperatures of the system is presented. Simulation results for propellant blow down during impulsive operation are also presented. The results from this study validate the feasibility of a practical thermally capacitive RTR propulsion system.

Burner liner thermal-structural load modeling

NASA Technical Reports Server (NTRS)

Maffeo, R.

1986-01-01

The software package Transfer Analysis Code to Interface Thermal/Structural Problems (TRANCITS) was developed. The TRANCITS code is used to interface temperature data between thermal and structural analytical models. The use of this transfer module allows the heat transfer analyst to select the thermal mesh density and thermal analysis code best suited to solve the thermal problem and gives the same freedoms to the stress analyst, without the efficiency penalties associated with common meshes and the accuracy penalties associated with the manual transfer of thermal data.

The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 1: Basic Models

NASA Astrophysics Data System (ADS)

Mosunova, N. A.

2018-05-01

The article describes the basic models included in the EUCLID/V1 integrated code intended for safety analysis of liquid metal (sodium, lead, and lead-bismuth) cooled fast reactors using fuel rods with a gas gap and pellet dioxide, mixed oxide or nitride uranium-plutonium fuel under normal operation, under anticipated operational occurrences and accident conditions by carrying out interconnected thermal-hydraulic, neutronics, and thermal-mechanical calculations. Information about the Russian and foreign analogs of the EUCLID/V1 integrated code is given. Modeled objects, equation systems in differential form solved in each module of the EUCLID/V1 integrated code (the thermal-hydraulic, neutronics, fuel rod analysis module, and the burnup and decay heat calculation modules), the main calculated quantities, and also the limitations on application of the code are presented. The article also gives data on the scope of functions performed by the integrated code's thermal-hydraulic module, using which it is possible to describe both one- and twophase processes occurring in the coolant. It is shown that, owing to the availability of the fuel rod analysis module in the integrated code, it becomes possible to estimate the performance of fuel rods in different regimes of the reactor operation. It is also shown that the models implemented in the code for calculating neutron-physical processes make it possible to take into account the neutron field distribution over the fuel assembly cross section as well as other features important for the safety assessment of fast reactors.

Advances in modelling of condensation phenomena

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liu, W.S.; Zaltsgendler, E.; Hanna, B.

1997-07-01

The physical parameters in the modelling of condensation phenomena in the CANDU reactor system codes are discussed. The experimental programs used for thermal-hydraulic code validation in the Canadian nuclear industry are briefly described. The modelling of vapour generation and in particular condensation plays a key role in modelling of postulated reactor transients. The condensation models adopted in the current state-of-the-art two-fluid CANDU reactor thermal-hydraulic system codes (CATHENA and TUF) are described. As examples of the modelling challenges faced, the simulation of a cold water injection experiment by CATHENA and the simulation of a condensation induced water hammer experiment by TUFmore » are described.« less

ARCADIA{sup R} - A New Generation of Coupled Neutronics / Core Thermal- Hydraulics Code System at AREVA NP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Curca-Tivig, Florin; Merk, Stephan; Pautz, Andreas

2007-07-01

Anticipating future needs of our customers and willing to concentrate synergies and competences existing in the company for the benefit of our customers, AREVA NP decided in 2002 to develop the next generation of coupled neutronics/ core thermal-hydraulic (TH) code systems for fuel assembly and core design calculations for both, PWR and BWR applications. The global CONVERGENCE project was born: after a feasibility study of one year (2002) and a conceptual phase of another year (2003), development was started at the beginning of 2004. The present paper introduces the CONVERGENCE project, presents the main feature of the new code systemmore » ARCADIA{sup R} and concludes on customer benefits. ARCADIA{sup R} is designed to meet AREVA NP market and customers' requirements worldwide. Besides state-of-the-art physical modeling, numerical performance and industrial functionality, the ARCADIA{sup R} system is featuring state-of-the-art software engineering. The new code system will bring a series of benefits for our customers: e.g. improved accuracy for heterogeneous cores (MOX/ UOX, Gd...), better description of nuclide chains, and access to local neutronics/ thermal-hydraulics and possibly thermal-mechanical information (3D pin by pin full core modeling). ARCADIA is a registered trademark of AREVA NP. (authors)« less

Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems

NASA Astrophysics Data System (ADS)

Alipchenkov, V. M.; Anfimov, A. M.; Afremov, D. A.; Gorbunov, V. S.; Zeigarnik, Yu. A.; Kudryavtsev, A. V.; Osipov, S. L.; Mosunova, N. A.; Strizhov, V. F.; Usov, E. V.

2016-02-01

The conceptual fundamentals of the development of the new-generation system thermal-hydraulic computational HYDRA-IBRAE/LM code are presented. The code is intended to simulate the thermalhydraulic processes that take place in the loops and the heat-exchange equipment of liquid-metal cooled fast reactor systems under normal operation and anticipated operational occurrences and during accidents. The paper provides a brief overview of Russian and foreign system thermal-hydraulic codes for modeling liquid-metal coolants and gives grounds for the necessity of development of a new-generation HYDRA-IBRAE/LM code. Considering the specific engineering features of the nuclear power plants (NPPs) equipped with the BN-1200 and the BREST-OD-300 reactors, the processes and the phenomena are singled out that require a detailed analysis and development of the models to be correctly described by the system thermal-hydraulic code in question. Information on the functionality of the computational code is provided, viz., the thermalhydraulic two-phase model, the properties of the sodium and the lead coolants, the closing equations for simulation of the heat-mass exchange processes, the models to describe the processes that take place during the steam-generator tube rupture, etc. The article gives a brief overview of the usability of the computational code, including a description of the support documentation and the supply package, as well as possibilities of taking advantages of the modern computer technologies, such as parallel computations. The paper shows the current state of verification and validation of the computational code; it also presents information on the principles of constructing of and populating the verification matrices for the BREST-OD-300 and the BN-1200 reactor systems. The prospects are outlined for further development of the HYDRA-IBRAE/LM code, introduction of new models into it, and enhancement of its usability. It is shown that the program of development and practical application of the code will allow carrying out in the nearest future the computations to analyze the safety of potential NPP projects at a qualitatively higher level.

The kinetics of aerosol particle formation and removal in NPP severe accidents

NASA Astrophysics Data System (ADS)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.; Dolganov, Rostislav A.

2016-06-01

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal-hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into the KUPOL-M thermal-hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.

PEBBLE: a two-dimensional steady-state pebble bed reactor thermal hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vondy, D.R.

1981-09-01

This report documents the local implementation of the PEBBLE code to treat the two-dimensional steady-state pebble bed reactor thermal hydraulics problem. This code is implemented as a module of a computation system used for reactor core history calculations. Given power density data, the geometric description in (RZ), and basic heat removal conditions and thermal properties, the coolant properties, flow conditions, and temperature distributions in the pebble fuel elements are predicted. The calculation is oriented to the continuous fueling, steady state condition with consideration of the effect of the high energy neutron flux exposure and temperature history on the thermal conductivity.more » The coolant flow conditions are calculated for the same geometry as used in the neutronics calculation, power density and fluence data being used directly, and temperature results are made available for subsequent use.« less

Summary of comparison and analysis of results from exercises 1 and 2 of the OECD PBMR coupled neutronics/thermal hydraulics transient benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mkhabela, P.; Han, J.; Tyobeka, B.

2006-07-01

The Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) has accepted, through the Nuclear Science Committee (NSC), the inclusion of the Pebble-Bed Modular Reactor 400 MW design (PBMR-400) coupled neutronics/thermal hydraulics transient benchmark problem as part of their official activities. The scope of the benchmark is to establish a well-defined problem, based on a common given library of cross sections, to compare methods and tools in core simulation and thermal hydraulics analysis with a specific focus on transient events through a set of multi-dimensional computational test problems. The benchmark includes three steady state exercises andmore » six transient exercises. This paper describes the first two steady state exercises, their objectives and the international participation in terms of organization, country and computer code utilized. This description is followed by a comparison and analysis of the participants' results submitted for these two exercises. The comparison of results from different codes allows for an assessment of the sensitivity of a result to the method employed and can thus help to focus the development efforts on the most critical areas. The two first exercises also allow for removing of user-related modeling errors and prepare core neutronics and thermal-hydraulics models of the different codes for the rest of the exercises in the benchmark. (authors)« less

Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

NASA Astrophysics Data System (ADS)

Hashim, Zaredah; Lanyau, Tonny Anak; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi; Azhar, Noraishah Syahirah

2016-01-01

The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel's center and surface, cladding, coolant temperatures as well as DNBR's values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

Inverse problem to constrain the controlling parameters of large-scale heat transport processes: The Tiberias Basin example

NASA Astrophysics Data System (ADS)

Goretzki, Nora; Inbar, Nimrod; Siebert, Christian; Möller, Peter; Rosenthal, Eliyahu; Schneider, Michael; Magri, Fabien

2015-04-01

Salty and thermal springs exist along the lakeshore of the Sea of Galilee, which covers most of the Tiberias Basin (TB) in the northern Jordan- Dead Sea Transform, Israel/Jordan. As it is the only freshwater reservoir of the entire area, it is important to study the salinisation processes that pollute the lake. Simulations of thermohaline flow along a 35 km NW-SE profile show that meteoric and relic brines are flushed by the regional flow from the surrounding heights and thermally induced groundwater flow within the faults (Magri et al., 2015). Several model runs with trial and error were necessary to calibrate the hydraulic conductivity of both faults and major aquifers in order to fit temperature logs and spring salinity. It turned out that the hydraulic conductivity of the faults ranges between 30 and 140 m/yr whereas the hydraulic conductivity of the Upper Cenomanian aquifer is as high as 200 m/yr. However, large-scale transport processes are also dependent on other physical parameters such as thermal conductivity, porosity and fluid thermal expansion coefficient, which are hardly known. Here, inverse problems (IP) are solved along the NW-SE profile to better constrain the physical parameters (a) hydraulic conductivity, (b) thermal conductivity and (c) thermal expansion coefficient. The PEST code (Doherty, 2010) is applied via the graphical interface FePEST in FEFLOW (Diersch, 2014). The results show that both thermal and hydraulic conductivity are consistent with the values determined with the trial and error calibrations. Besides being an automatic approach that speeds up the calibration process, the IP allows to cover a wide range of parameter values, providing additional solutions not found with the trial and error method. Our study shows that geothermal systems like TB are more comprehensively understood when inverse models are applied to constrain coupled fluid flow processes over large spatial scales. References Diersch, H.-J.G., 2014. FEFLOW Finite Element Modeling of Flow, Mass and Heat Transport in Porous and Fractured Media. Springer- Verlag Berlin Heidelberg ,996p. Doherty J., 2010, PEST: Model-Independent Parameter Estimation. user manual 5th Edition. Watermark, Brisbane, Australia Magri, F., Inbar, N., Siebert C., Rosenthal, E., Guttman, J., Möller, P., 2015. Transient simulations of large-scale hydrogeological processes causing temperature and salinity anomalies in the Tiberias Basin. Journal of Hydrology, 520(0), 342-355.

Development of NSSS Thermal-Hydraulic Model for KNPEC-2 Simulator Using the Best-Estimate Code RETRAN-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kyung-Doo; Jeong, Jae-Jun; Lee, Seung-Wook

The Nuclear Steam Supply System (NSSS) thermal-hydraulic model adopted in the Korea Nuclear Plant Education Center (KNPEC)-2 simulator was provided in the early 1980s. The reference plant for KNPEC-2 is the Yong Gwang Nuclear Unit 1, which is a Westinghouse-type 3-loop, 950 MW(electric) pressurized water reactor. Because of the limited computational capability at that time, it uses overly simplified physical models and assumptions for a real-time simulation of NSSS thermal-hydraulic transients. This may entail inaccurate results and thus, the possibility of so-called ''negative training,'' especially for complicated two-phase flows in the reactor coolant system. To resolve the problem, we developedmore » a realistic NSSS thermal-hydraulic program (named ARTS code) based on the best-estimate code RETRAN-3D. The systematic assessment of ARTS has been conducted by both a stand-alone test and an integrated test in the simulator environment. The non-integrated stand-alone test (NIST) results were reasonable in terms of accuracy, real-time simulation capability, and robustness. After successful completion of the NIST, ARTS was integrated with a 3-D reactor kinetics model and other system models. The site acceptance test (SAT) has been completed successively and confirmed to comply with the ANSI/ANS-3.5-1998 simulator software performance criteria. This paper presents our efforts for the ARTS development and some test results of the NIST and SAT.« less

1DTempPro V2: new features for inferring groundwater/surface-water exchange

USGS Publications Warehouse

Koch, Franklin W.; Voytek, Emily B.; Day-Lewis, Frederick D.; Healy, Richard W.; Briggs, Martin A.; Lane, John W.; Werkema, Dale D.

2016-01-01

A new version of the computer program 1DTempPro extends the original code to include new capabilities for (1) automated parameter estimation, (2) layer heterogeneity, and (3) time-varying specific discharge. The code serves as an interface to the U.S. Geological Survey model VS2DH and supports analysis of vertical one-dimensional temperature profiles under saturated flow conditions to assess groundwater/surface-water exchange and estimate hydraulic conductivity for cases where hydraulic head is known.

Thermal hydraulic simulations, error estimation and parameter sensitivity studies in Drekar::CFD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, Thomas Michael; Shadid, John N.; Pawlowski, Roger P.

2014-01-01

This report describes work directed towards completion of the Thermal Hydraulics Methods (THM) CFD Level 3 Milestone THM.CFD.P7.05 for the Consortium for Advanced Simulation of Light Water Reactors (CASL) Nuclear Hub effort. The focus of this milestone was to demonstrate the thermal hydraulics and adjoint based error estimation and parameter sensitivity capabilities in the CFD code called Drekar::CFD. This milestone builds upon the capabilities demonstrated in three earlier milestones; THM.CFD.P4.02 [12], completed March, 31, 2012, THM.CFD.P5.01 [15] completed June 30, 2012 and THM.CFD.P5.01 [11] completed on October 31, 2012.

An approach to model reactor core nodalization for deterministic safety analysis

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

2016-01-01

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

An approach to model reactor core nodalization for deterministic safety analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my; Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my

Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to bemore » employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.« less

Methodology, status and plans for development and assessment of the code ATHLET

DOE Office of Scientific and Technical Information (OSTI.GOV)

Teschendorff, V.; Austregesilo, H.; Lerchl, G.

1997-07-01

The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is being developed by the Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors. The aim of the code development is to cover the whole spectrum of design basis and beyond design basis accidents (without core degradation) for PWRs and BWRs with only one code. The main code features are: advanced thermal-hydraulics; modular code architecture; separation between physical models and numerical methods; pre- and post-processing tools; portability. The codemore » has features that are of special interest for applications to small leaks and transients with accident management, e.g. initialization by a steady-state calculation, full-range drift-flux model, dynamic mixture level tracking. The General Control Simulation Module of ATHLET is a flexible tool for the simulation of the balance-of-plant and control systems including the various operator actions in the course of accident sequences with AM measures. The code development is accompained by a systematic and comprehensive validation program. A large number of integral experiments and separate effect tests, including the major International Standard Problems, have been calculated by GRS and by independent organizations. The ATHLET validation matrix is a well balanced set of integral and separate effects tests derived from the CSNI proposal emphasizing, however, the German combined ECC injection system which was investigated in the UPTF, PKL and LOBI test facilities.« less

Development of a three-dimensional transient code for reactivity-initiated events of BWRs (boiling water reactors) - Models and code verifications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uematsu, Hitoshi; Yamamoto, Toru; Izutsu, Sadayuki

1990-06-01

A reactivity-initiated event is a design-basis accident for the safety analysis of boiling water reactors. It is defined as a rapid transient of reactor power caused by a reactivity insertion of over $1.0 due to a postulated drop or abnormal withdrawal of the control rod from the core. Strong space-dependent feedback effects are associated with the local power increase due to control rod movement. A realistic treatment of the core status in a transient by a code with a detailed core model is recommended in evaluating this event. A three-dimensional transient code, ARIES, has been developed to meet this need.more » The code simulates the event with three-dimensional neutronics, coupled with multichannel thermal hydraulics, based on a nonequilibrium separated flow model. The experimental data obtained in reactivity accident tests performed with the SPERT III-E core are used to verify the entire code, including thermal-hydraulic models.« less

RAMONA-3B application to Browns Ferry ATWS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Slovik, G.C.; Neymotin, L.; Cazzoli, E.

1984-01-01

This paper discusses two preliminary MSIV clsoure ATWS calculations done using the RAMONA-3B code and the work being done to create the necessary cross section sets for the Browns Ferry Unit 1 reactor. The RAMONA-3B code employs a three-dimensional neutron kinetics model coupled with one-dimensional, four equation, nonhomogeneous, nonequilibrium thermal hydraulics. To be compatible with 3-D neutron kinetics, the code uses parallel coolant channels in the core. It also includes a boron transport model and all necessary BWR components such as jet pump, recirculation pump, steam separator, steamline with safety and relief valves, main steam isolation valve, turbine stop valve,more » and turbine bypass valve. A summary of RAMONA-3B neutron kinetics and thermal hydraulics models is presented in the Appendix.« less

Post-Test Analysis of 11% Break at PSB-VVER Experimental Facility using Cathare 2 Code

NASA Astrophysics Data System (ADS)

Sabotinov, Luben; Chevrier, Patrick

The best estimate French thermal-hydraulic computer code CATHARE 2 Version 2.5_1 was used for post-test analysis of the experiment “11% upper plenum break”, conducted at the large-scale test facility PSB-VVER in Russia. The PSB rig is 1:300 scaled model of VVER-1000 NPP. A computer model has been developed for CATHARE 2 V2.5_1, taking into account all important components of the PSB facility: reactor model (lower plenum, core, bypass, upper plenum, downcomer), 4 separated loops, pressurizer, horizontal multitube steam generators, break section. The secondary side is represented by recirculation model. A large number of sensitivity calculations has been performed regarding break modeling, reactor pressure vessel modeling, counter current flow modeling, hydraulic losses, heat losses. The comparison between calculated and experimental results shows good prediction of the basic thermal-hydraulic phenomena and parameters such as pressures, temperatures, void fractions, loop seal clearance, etc. The experimental and calculation results are very sensitive regarding the fuel cladding temperature, which show a periodical nature. With the applied CATHARE 1D modeling, the global thermal-hydraulic parameters and the core heat up have been reasonably predicted.

Predicting Formation Damage in Aquifer Thermal Energy Storage Systems Utilizing a Coupled Hydraulic-Thermal-Chemical Reservoir Model

NASA Astrophysics Data System (ADS)

Müller, Daniel; Regenspurg, Simona; Milsch, Harald; Blöcher, Guido; Kranz, Stefan; Saadat, Ali

2014-05-01

In aquifer thermal energy storage (ATES) systems, large amounts of energy can be stored by injecting hot water into deep or intermediate aquifers. In a seasonal production-injection cycle, water is circulated through a system comprising the porous aquifer, a production well, a heat exchanger and an injection well. This process involves large temperature and pressure differences, which shift chemical equilibria and introduce or amplify mechanical processes. Rock-fluid interaction such as dissolution and precipitation or migration and deposition of fine particles will affect the hydraulic properties of the porous medium and may lead to irreversible formation damage. In consequence, these processes determine the long-term performance of the ATES system and need to be predicted to ensure the reliability of the system. However, high temperature and pressure gradients and dynamic feedback cycles pose challenges on predicting the influence of the relevant processes. Within this study, a reservoir model comprising a coupled hydraulic-thermal-chemical simulation was developed based on an ATES demonstration project located in the city of Berlin, Germany. The structural model was created with Petrel, based on data available from seismic cross-sections and wellbores. The reservoir simulation was realized by combining the capabilities of multiple simulation tools. For the reactive transport model, COMSOL Multiphysics (hydraulic-thermal) and PHREEQC (chemical) were combined using the novel interface COMSOL_PHREEQC, developed by Wissmeier & Barry (2011). It provides a MATLAB-based coupling interface between both programs. Compared to using COMSOL's built-in reactive transport simulator, PHREEQC additionally calculates adsorption and reaction kinetics and allows the selection of different activity coefficient models in the database. The presented simulation tool will be able to predict the most important aspects of hydraulic, thermal and chemical transport processes relevant to formation damage in ATES systems. We would like to present preliminary results of the structural reservoir model and the hydraulic-thermal-chemical coupling for the demonstration site. Literature: Wissmeier, L. and Barry, D.A., 2011. Simulation tool for variably saturated flow with comprehensive geochemical reactions in two- and three-dimensional domains. Environmental Modelling & Software 26, 210-218.

Combining Thermal And Structural Analyses

NASA Technical Reports Server (NTRS)

Winegar, Steven R.

1990-01-01

Computer code makes programs compatible so stresses and deformations calculated. Paper describes computer code combining thermal analysis with structural analysis. Called SNIP (for SINDA-NASTRAN Interfacing Program), code provides interface between finite-difference thermal model of system and finite-element structural model when no node-to-element correlation between models. Eliminates much manual work in converting temperature results of SINDA (Systems Improved Numerical Differencing Analyzer) program into thermal loads for NASTRAN (NASA Structural Analysis) program. Used to analyze concentrating reflectors for solar generation of electric power. Large thermal and structural models needed to predict distortion of surface shapes, and SNIP saves considerable time and effort in combining models.

Extensions of the MCNP5 and TRIPOLI4 Monte Carlo Codes for Transient Reactor Analysis

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Sjenitzer, Bart L.

2014-06-01

To simulate reactor transients for safety analysis with the Monte Carlo method the generation and decay of delayed neutron precursors is implemented in the MCNP5 and TRIPOLI4 general purpose Monte Carlo codes. Important new variance reduction techniques like forced decay of precursors in each time interval and the branchless collision method are included to obtain reasonable statistics for the power production per time interval. For simulation of practical reactor transients also the feedback effect from the thermal-hydraulics must be included. This requires coupling of the Monte Carlo code with a thermal-hydraulics (TH) code, providing the temperature distribution in the reactor, which affects the neutron transport via the cross section data. The TH code also provides the coolant density distribution in the reactor, directly influencing the neutron transport. Different techniques for this coupling are discussed. As a demonstration a 3x3 mini fuel assembly with a moving control rod is considered for MCNP5 and a mini core existing of 3x3 PWR fuel assemblies with control rods and burnable poisons for TRIPOLI4. Results are shown for reactor transients due to control rod movement or withdrawal. The TRIPOLI4 transient calculation is started at low power and includes thermal-hydraulic feedback. The power rises about 10 decades and finally stabilises the reactor power at a much higher level than initial. The examples demonstrate that the modified Monte Carlo codes are capable of performing correct transient calculations, taking into account all geometrical and cross section detail.

Cookbook Recipe to Simulate Seawater Intrusion with Standard MODFLOW

NASA Astrophysics Data System (ADS)

Schaars, F.; Bakker, M.

2012-12-01

We developed a cookbook recipe to simulate steady interface flow in multi-layer coastal aquifers with regular groundwater codes such as standard MODFLOW. The main step in the recipe is a simple transformation of the hydraulic conductivities and thicknesses of the aquifers. Standard groundwater codes may be applied to compute the head distribution in the aquifer using the transformed parameters. For example, for flow in a single unconfined aquifer, the hydraulic conductivity needs to be multiplied with 41 and the base of the aquifer needs to be set to mean sea level (for a relative seawater density of 1.025). Once the head distribution is obtained, the Ghijben-Herzberg relationship is applied to compute the depth of the interface. The recipe may be applied to quite general settings, including spatially variable aquifer properties. Any standard groundwater code may be used, as long as it can simulate unconfined flow where the transmissivity is a linear function of the head. The proposed recipe is benchmarked successfully against a number of analytic and numerical solutions.

The development of a thermal hydraulic feedback mechanism with a quasi-fixed point iteration scheme for control rod position modeling for the TRIGSIMS-TH application

NASA Astrophysics Data System (ADS)

Karriem, Veronica V.

Nuclear reactor design incorporates the study and application of nuclear physics, nuclear thermal hydraulic and nuclear safety. Theoretical models and numerical methods implemented in computer programs are utilized to analyze and design nuclear reactors. The focus of this PhD study's is the development of an advanced high-fidelity multi-physics code system to perform reactor core analysis for design and safety evaluations of research TRIGA-type reactors. The fuel management and design code system TRIGSIMS was further developed to fulfill the function of a reactor design and analysis code system for the Pennsylvania State Breazeale Reactor (PSBR). TRIGSIMS, which is currently in use at the PSBR, is a fuel management tool, which incorporates the depletion code ORIGEN-S (part of SCALE system) and the Monte Carlo neutronics solver MCNP. The diffusion theory code ADMARC-H is used within TRIGSIMS to accelerate the MCNP calculations. It manages the data and fuel isotopic content and stores it for future burnup calculations. The contribution of this work is the development of an improved version of TRIGSIMS, named TRIGSIMS-TH. TRIGSIMS-TH incorporates a thermal hydraulic module based on the advanced sub-channel code COBRA-TF (CTF). CTF provides the temperature feedback needed in the multi-physics calculations as well as the thermal hydraulics modeling capability of the reactor core. The temperature feedback model is using the CTF-provided local moderator and fuel temperatures for the cross-section modeling for ADMARC-H and MCNP calculations. To perform efficient critical control rod calculations, a methodology for applying a control rod position was implemented in TRIGSIMS-TH, making this code system a modeling and design tool for future core loadings. The new TRIGSIMS-TH is a computer program that interlinks various other functional reactor analysis tools. It consists of the MCNP5, ADMARC-H, ORIGEN-S, and CTF. CTF was coupled with both MCNP and ADMARC-H to provide the heterogeneous temperature distribution throughout the core. Each of these codes is written in its own computer language performing its function and outputs a set of data. TRIGSIMS-TH provides an effective use and data manipulation and transfer between different codes. With the implementation of feedback and control- rod-position modeling methodologies, the TRIGSIMS-TH calculations are more accurate and in a better agreement with measured data. The PSBR is unique in many ways and there are no "off-the-shelf" codes, which can model this design in its entirety. In particular, PSBR has an open core design, which is cooled by natural convection. Combining several codes into a unique system brings many challenges. It also requires substantial knowledge of both operation and core design of the PSBR. This reactor is in operation decades and there is a fair amount of studies and developments in both PSBR thermal hydraulics and neutronics. Measured data is also available for various core loadings and can be used for validation activities. The previous studies and developments in PSBR modeling also aids as a guide to assess the findings of the work herein. In order to incorporate new methods and codes into exiting TRIGSIMS, a re-evaluation of various components of the code was performed to assure the accuracy and efficiency of the existing CTF/MCNP5/ADMARC-H multi-physics coupling. A new set of ADMARC-H diffusion coefficients and cross sections was generated using the SERPENT code. This was needed as the previous data was not generated with thermal hydraulic feedback and the ARO position was used as the critical rod position. The B4C was re-evaluated for this update. The data exchange between ADMARC-H and MCNP5 was modified. The basic core model is given a flexibility to allow for various changes within the core model, and this feature was implemented in TRIGSIMS-TH. The PSBR core in the new code model can be expanded and changed. This allows the new code to be used as a modeling tool for design and analyses of future code loadings.

Particle bed reactor modeling

NASA Technical Reports Server (NTRS)

Sapyta, Joe; Reid, Hank; Walton, Lew

1993-01-01

The topics are presented in viewgraph form and include the following: particle bed reactor (PBR) core cross section; PBR bleed cycle; fuel and moderator flow paths; PBR modeling requirements; characteristics of PBR and nuclear thermal propulsion (NTP) modeling; challenges for PBR and NTP modeling; thermal hydraulic computer codes; capabilities for PBR/reactor application; thermal/hydralic codes; limitations; physical correlations; comparison of predicted friction factor and experimental data; frit pressure drop testing; cold frit mask factor; decay heat flow rate; startup transient simulation; and philosophy of systems modeling.

Influence of geologic layering on heat transport and storage in an aquifer thermal energy storage system

NASA Astrophysics Data System (ADS)

Bridger, D. W.; Allen, D. M.

2014-01-01

A modeling study was carried out to evaluate the influence of aquifer heterogeneity, as represented by geologic layering, on heat transport and storage in an aquifer thermal energy storage (ATES) system in Agassiz, British Columbia, Canada. Two 3D heat transport models were developed and calibrated using the flow and heat transport code FEFLOW including: a "non-layered" model domain with homogeneous hydraulic and thermal properties; and, a "layered" model domain with variable hydraulic and thermal properties assigned to discrete geological units to represent aquifer heterogeneity. The base model (non-layered) shows limited sensitivity for the ranges of all thermal and hydraulic properties expected at the site; the model is most sensitive to vertical anisotropy and hydraulic gradient. Simulated and observed temperatures within the wells reflect a combination of screen placement and layering, with inconsistencies largely explained by the lateral continuity of high permeability layers represented in the model. Simulation of heat injection, storage and recovery show preferential transport along high permeability layers, resulting in longitudinal plume distortion, and overall higher short-term storage efficiencies.

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Ivanov, Aleksandar; Sanchez, Victor; Ivanov, Kostadin

2014-06-01

Monte Carlo methods have been used as reference reactor physics calculation tools worldwide. The advance in computer technology allows the calculation of detailed flux distributions in both space and energy. In most of the cases however, those calculations are done under the assumption of homogeneous material density and temperature distributions. The aim of this work is to develop a consistent methodology for providing realistic three-dimensional thermal-hydraulic distributions by coupling the in-house developed sub-channel code SUBCHANFLOW with the standard Monte-Carlo transport code MCNP. In addition to the innovative technique of on-the fly material definition, a flux-based weight-window technique has been introduced to improve both the magnitude and the distribution of the relative errors. Finally, a coupled code system for the simulation of steady-state reactor physics problems has been developed. Besides the problem of effective feedback data interchange between the codes, the treatment of temperature dependence of the continuous energy nuclear data has been investigated.

A Steady State and Quasi-Steady Interface Between the Generalized Fluid System Simulation Program and the SINDA/G Thermal Analysis Program

NASA Technical Reports Server (NTRS)

Schallhorn, Paul; Majumdar, Alok; Tiller, Bruce

2001-01-01

A general purpose, one dimensional fluid flow code is currently being interfaced with the thermal analysis program SINDA/G. The flow code, GFSSP, is capable of analyzing steady state and transient flow in a complex network. The flow code is capable of modeling several physical phenomena including compressibility effects, phase changes, body forces (such as gravity and centrifugal) and mixture thermodynamics for multiple species. The addition of GFSSP to SINDA/G provides a significant improvement in convective heat transfer modeling for SINDA/G. The interface development is conducted in multiple phases. This paper describes the first phase of the interface which allows for steady and quasisteady (unsteady solid, steady fluid) conjugate heat transfer modeling.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using themore » WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)« less

Three-dimensional time-dependent STAR reactor kinetics analyses coupled with RETRAN and MCPWR system response

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.

1989-11-01

The operation of a nuclear power plant must be regularly supported by various reactor dynamics and thermal-hydraulic analyses, which may include final safety analysis report (FSAR) design-basis calculations, and conservative and best-estimate analyses. The development and improvement of computer codes and analysis methodologies provide many advantages, including the ability to evaluate the effect of modeling simplifications and assumptions made in previous reactor kinetics and thermal-hydraulic calculations. This paper describes the results of using the RETRAN, MCPWR, and STAR codes in a tandem, predictive-corrective manner for three pressurized water reactor (PWR) transients: (a) loss of feedwater (LOF) anticipated transient without scrammore » (ATWS), (b) station blackout ATWS, and (c) loss of total reactor coolant system (RCS) flow with a scram.« less

Verification of the predictive capabilities of the 4C code cryogenic circuit model

NASA Astrophysics Data System (ADS)

Zanino, R.; Bonifetto, R.; Hoa, C.; Richard, L. Savoldi

2014-01-01

The 4C code was developed to model thermal-hydraulics in superconducting magnet systems and related cryogenic circuits. It consists of three coupled modules: a quasi-3D thermal-hydraulic model of the winding; a quasi-3D model of heat conduction in the magnet structures; an object-oriented a-causal model of the cryogenic circuit. In the last couple of years the code and its different modules have undergone a series of validation exercises against experimental data, including also data coming from the supercritical He loop HELIOS at CEA Grenoble. However, all this analysis work was done each time after the experiments had been performed. In this paper a first demonstration is given of the predictive capabilities of the 4C code cryogenic circuit module. To do that, a set of ad-hoc experimental scenarios have been designed, including different heating and control strategies. Simulations with the cryogenic circuit module of 4C have then been performed before the experiment. The comparison presented here between the code predictions and the results of the HELIOS measurements gives the first proof of the excellent predictive capability of the 4C code cryogenic circuit module.

Interfacing a General Purpose Fluid Network Flow Program with the SINDA/G Thermal Analysis Program

NASA Technical Reports Server (NTRS)

Schallhorn, Paul; Popok, Daniel

1999-01-01

A general purpose, one dimensional fluid flow code is currently being interfaced with the thermal analysis program Systems Improved Numerical Differencing Analyzer/Gaski (SINDA/G). The flow code, Generalized Fluid System Simulation Program (GFSSP), is capable of analyzing steady state and transient flow in a complex network. The flow code is capable of modeling several physical phenomena including compressibility effects, phase changes, body forces (such as gravity and centrifugal) and mixture thermodynamics for multiple species. The addition of GFSSP to SINDA/G provides a significant improvement in convective heat transfer modeling for SINDA/G. The interface development is conducted in multiple phases. This paper describes the first phase of the interface which allows for steady and quasi-steady (unsteady solid, steady fluid) conjugate heat transfer modeling.

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

DOE Office of Scientific and Technical Information (OSTI.GOV)

O'Brien, James E.; Sabharwall, Piyush; Yoon, Su -Jong

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs)more » at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation needs. The experimental database will guide development of appropriate predictive methods and be available for code verification and validation (V&V) related to these systems.« less

Assessment of the TRACE Reactor Analysis Code Against Selected PANDA Transient Data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zavisca, M.; Ghaderi, M.; Khatib-Rahbar, M.

2006-07-01

The TRACE (TRAC/RELAP Advanced Computational Engine) code is an advanced, best-estimate thermal-hydraulic program intended to simulate the transient behavior of light-water reactor systems, using a two-fluid (steam and water, with non-condensable gas), seven-equation representation of the conservation equations and flow-regime dependent constitutive relations in a component-based model with one-, two-, or three-dimensional elements, as well as solid heat structures and logical elements for the control system. The U.S. Nuclear Regulatory Commission is currently supporting the development of the TRACE code and its assessment against a variety of experimental data pertinent to existing and evolutionary reactor designs. This paper presents themore » results of TRACE post-test prediction of P-series of experiments (i.e., tests comprising the ISP-42 blind and open phases) conducted at the PANDA large-scale test facility in 1990's. These results show reasonable agreement with the reported test results, indicating good performance of the code and relevant underlying thermal-hydraulic and heat transfer models. (authors)« less

VICTORIA: A mechanistic model for radionuclide behavior in the reactor coolant system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schaperow, J.H.; Bixler, N.E.

1996-12-31

VICTORIA is the U.S. Nuclear Regulatory Commission`s (NRC`s) mechanistic, best-estimate code for analysis of fission product release from the core and subsequent transport in the reactor vessel and reactor coolant system. VICTORIA requires thermal-hydraulic data (i.e., temperatures, pressures, and velocities) as input. In the past, these data have been taken from the results of calculations from thermal-hydraulic codes such as SCDAP/RELAP5, MELCOR, and MAAP. Validation and assessment of VICTORIA 1.0 have been completed. An independent peer review of VICTORIA, directed by Brookhaven National Laboratory and supported by experts in the areas of fuel release, fission product chemistry, and aerosol physics,more » has been undertaken. This peer review, which will independently assess the code`s capabilities, is nearing completion with the peer review committee`s final report expected in Dec 1996. A limited amount of additional development is expected as a result of the peer review. Following this additional development, the NRC plans to release VICTORIA 1.1 and an updated and improved code manual. Future plans mainly involve use of the code for plant calculations to investigate specific safety issues as they arise. Also, the code will continue to be used in support of the Phebus experiments.« less

Thermal-hydraulic modeling needs for passive reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kelly, J.M.

1997-07-01

The U.S. Nuclear Regulatory Commission has received an application for design certification from the Westinghouse Electric Corporation for an Advanced Light Water Reactor design known as the AP600. As part of the design certification process, the USNRC uses its thermal-hydraulic system analysis codes to independently audit the vendor calculations. The focus of this effort has been the small break LOCA transients that rely upon the passive safety features of the design to depressurize the primary system sufficiently so that gravity driven injection can provide a stable source for long term cooling. Of course, large break LOCAs have also been considered,more » but as the involved phenomena do not appear to be appreciably different from those of current plants, they were not discussed in this paper. Although the SBLOCA scenario does not appear to threaten core coolability - indeed, heatup is not even expected to occur - there have been concerns as to the performance of the passive safety systems. For example, the passive systems drive flows with small heads, consequently requiring more precision in the analysis compared to active systems methods for passive plants as compared to current plants with active systems. For the analysis of SBLOCAs and operating transients, the USNRC uses the RELAP5 thermal-hydraulic system analysis code. To assure the applicability of RELAP5 to the analysis of these transients for the AP600 design, a four year long program of code development and assessment has been undertaken.« less

New Multi-group Transport Neutronics (PHISICS) Capabilities for RELAP5-3D and its Application to Phase I of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Cristian Rabiti; Andrea Alfonsi

2012-10-01

PHISICS is a neutronics code system currently under development at the Idaho National Laboratory (INL). Its goal is to provide state of the art simulation capability to reactor designers. The different modules for PHISICS currently under development are a nodal and semi-structured transport core solver (INSTANT), a depletion module (MRTAU) and a cross section interpolation (MIXER) module. The INSTANT module is the most developed of the mentioned above. Basic functionalities are ready to use, but the code is still in continuous development to extend its capabilities. This paper reports on the effort of coupling the nodal kinetics code package PHISICSmore » (INSTANT/MRTAU/MIXER) to the thermal hydraulics system code RELAP5-3D, to enable full core and system modeling. This will enable the possibility to model coupled (thermal-hydraulics and neutronics) problems with more options for 3D neutron kinetics, compared to the existing diffusion theory neutron kinetics module in RELAP5-3D (NESTLE). In the second part of the paper, an overview of the OECD/NEA MHTGR-350 MW benchmark is given. This benchmark has been approved by the OECD, and is based on the General Atomics 350 MW Modular High Temperature Gas Reactor (MHTGR) design. The benchmark includes coupled neutronics thermal hydraulics exercises that require more capabilities than RELAP5-3D with NESTLE offers. Therefore, the MHTGR benchmark makes extensive use of the new PHISICS/RELAP5-3D coupling capabilities. The paper presents the preliminary results of the three steady state exercises specified in Phase I of the benchmark using PHISICS/RELAP5-3D.« less

Implicit time-integration method for simultaneous solution of a coupled non-linear system

NASA Astrophysics Data System (ADS)

Watson, Justin Kyle

Historically large physical problems have been divided into smaller problems based on the physics involved. This is no different in reactor safety analysis. The problem of analyzing a nuclear reactor for design basis accidents is performed by a handful of computer codes each solving a portion of the problem. The reactor thermal hydraulic response to an event is determined using a system code like TRAC RELAP Advanced Computational Engine (TRACE). The core power response to the same accident scenario is determined using a core physics code like Purdue Advanced Core Simulator (PARCS). Containment response to the reactor depressurization in a Loss Of Coolant Accident (LOCA) type event is calculated by a separate code. Sub-channel analysis is performed with yet another computer code. This is just a sample of the computer codes used to solve the overall problems of nuclear reactor design basis accidents. Traditionally each of these codes operates independently from each other using only the global results from one calculation as boundary conditions to another. Industry's drive to uprate power for reactors has motivated analysts to move from a conservative approach to design basis accident towards a best estimate method. To achieve a best estimate calculation efforts have been aimed at coupling the individual physics models to improve the accuracy of the analysis and reduce margins. The current coupling techniques are sequential in nature. During a calculation time-step data is passed between the two codes. The individual codes solve their portion of the calculation and converge to a solution before the calculation is allowed to proceed to the next time-step. This thesis presents a fully implicit method of simultaneous solving the neutron balance equations, heat conduction equations and the constitutive fluid dynamics equations. It discusses the problems involved in coupling different physics phenomena within multi-physics codes and presents a solution to these problems. The thesis also outlines the basic concepts behind the nodal balance equations, heat transfer equations and the thermal hydraulic equations, which will be coupled to form a fully implicit nonlinear system of equations. The coupling of separate physics models to solve a larger problem and improve accuracy and efficiency of a calculation is not a new idea, however implementing them in an implicit manner and solving the system simultaneously is. Also the application to reactor safety codes is new and has not be done with thermal hydraulics and neutronics codes on realistic applications in the past. The coupling technique described in this thesis is applicable to other similar coupled thermal hydraulic and core physics reactor safety codes. This technique is demonstrated using coupled input decks to show that the system is solved correctly and then verified by using two derivative test problems based on international benchmark problems the OECD/NRC Three mile Island (TMI) Main Steam Line Break (MSLB) problem (representative of pressurized water reactor analysis) and the OECD/NRC Peach Bottom (PB) Turbine Trip (TT) benchmark (representative of boiling water reactor analysis).

RETRANO3 benchmarks for Beaver Valley plant transients and FSAR analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Beaumont, E.T.; Feltus, M.A.

1993-01-01

Any best-estimate code (e.g., RETRANO3) results must be validated against plant data and final safety analysis report (FSAR) predictions. The need for two independent means of benchmarking is necessary to ensure that the results were not biased toward a particular data set and to have a certain degree of accuracy. The code results need to be compared with previous results and show improvements over previous code results. Ideally, the two best means of benchmarking a thermal hydraulics code are comparing results from previous versions of the same code along with actual plant data. This paper describes RETRAN03 benchmarks against RETRAN02more » results, actual plant data, and FSAR predictions. RETRAN03, the Electric Power Research Institute's latest version of the RETRAN thermal-hydraulic analysis codes, offers several upgrades over its predecessor, RETRAN02 Mod5. RETRAN03 can use either implicit or semi-implicit numerics, whereas RETRAN02 Mod5 uses only semi-implicit numerics. Another major upgrade deals with slip model options. RETRAN03 added several new models, including a five-equation model for more accurate modeling of two-phase flow. RETPAN02 Mod5 should give similar but slightly more conservative results than RETRAN03 when executed with RETRAN02 Mod5 options.« less

Current and anticipated uses of thermal-hydraulic codes in Spain

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pelayo, F.; Reventos, F.

1997-07-01

Spanish activities in the field of Applied Thermal-Hydraulics are steadily increasing as the codes are becoming practicable enough to efficiently sustain engineering decision in the Nuclear Power industry. Before reaching this point, a lot of effort has been devoted to achieve this goal. This paper briefly describes this process, points at the current applications and draws conclusions on the limitations. Finally it establishes the applications where the use of T-H codes would be worth in the future, this in turn implies further development of the codes to widen the scope of application and improve the general performance. Due to themore » different uses of the codes, the applications mainly come from the authority, industry, universities and research institutions. The main conclusion derived from this paper establishes that further code development is justified if the following requisites are considered: (1) Safety relevance of scenarios not presently covered is established. (2) A substantial gain in margins or the capability to use realistic assumptions is obtained. (3) A general consensus on the licensability and methodology for application is reached. The role of Regulatory Body is stressed, as the most relevant outcome of the project may be related to the evolution of the licensing frame.« less

Nuclear thermal propulsion engine system design analysis code development

NASA Astrophysics Data System (ADS)

Pelaccio, Dennis G.; Scheil, Christine M.; Petrosky, Lyman J.; Ivanenok, Joseph F.

1992-01-01

A Nuclear Thermal Propulsion (NTP) Engine System Design Analyis Code has recently been developed to characterize key NTP engine system design features. Such a versatile, standalone NTP system performance and engine design code is required to support ongoing and future engine system and vehicle design efforts associated with proposed Space Exploration Initiative (SEI) missions of interest. Key areas of interest in the engine system modeling effort were the reactor, shielding, and inclusion of an engine multi-redundant propellant pump feed system design option. A solid-core nuclear thermal reactor and internal shielding code model was developed to estimate the reactor's thermal-hydraulic and physical parameters based on a prescribed thermal output which was integrated into a state-of-the-art engine system design model. The reactor code module has the capability to model graphite, composite, or carbide fuels. Key output from the model consists of reactor parameters such as thermal power, pressure drop, thermal profile, and heat generation in cooled structures (reflector, shield, and core supports), as well as the engine system parameters such as weight, dimensions, pressures, temperatures, mass flows, and performance. The model's overall analysis methodology and its key assumptions and capabilities are summarized in this paper.

The SAS4A/SASSYS-1 Safety Analysis Code System, Version 5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fanning, T. H.; Brunett, A. J.; Sumner, T.

The SAS4A/SASSYS-1 computer code is developed by Argonne National Laboratory for thermal, hydraulic, and neutronic analysis of power and flow transients in liquidmetal- cooled nuclear reactors (LMRs). SAS4A was developed to analyze severe core disruption accidents with coolant boiling and fuel melting and relocation, initiated by a very low probability coincidence of an accident precursor and failure of one or more safety systems. SASSYS-1, originally developed to address loss-of-decay-heat-removal accidents, has evolved into a tool for margin assessment in design basis accident (DBA) analysis and for consequence assessment in beyond-design-basis accident (BDBA) analysis. SAS4A contains detailed, mechanistic models of transientmore » thermal, hydraulic, neutronic, and mechanical phenomena to describe the response of the reactor core, its coolant, fuel elements, and structural members to accident conditions. The core channel models in SAS4A provide the capability to analyze the initial phase of core disruptive accidents, through coolant heat-up and boiling, fuel element failure, and fuel melting and relocation. Originally developed to analyze oxide fuel clad with stainless steel, the models in SAS4A have been extended and specialized to metallic fuel with advanced alloy cladding. SASSYS-1 provides the capability to perform a detailed thermal/hydraulic simulation of the primary and secondary sodium coolant circuits and the balance-ofplant steam/water circuit. These sodium and steam circuit models include component models for heat exchangers, pumps, valves, turbines, and condensers, and thermal/hydraulic models of pipes and plena. SASSYS-1 also contains a plant protection and control system modeling capability, which provides digital representations of reactor, pump, and valve controllers and their response to input signal changes.« less

Pump-stopping water hammer simulation based on RELAP5

NASA Astrophysics Data System (ADS)

Yi, W. S.; Jiang, J.; Li, D. D.; Lan, G.; Zhao, Z.

2013-12-01

RELAP5 was originally designed to analyze complex thermal-hydraulic interactions that occur during either postulated large or small loss-of-coolant accidents in PWRs. However, as development continued, the code was expanded to include many of the transient scenarios that might occur in thermal-hydraulic systems. The fast deceleration of the liquid results in high pressure surges, thus the kinetic energy is transformed into the potential energy, which leads to the temporary pressure increase. This phenomenon is called water hammer. Generally water hammer can occur in any thermal-hydraulic systems and it is extremely dangerous for the system when the pressure surges become considerably high. If this happens and when the pressure exceeds the critical pressure that the pipe or the fittings along the pipeline can burden, it will result in the failure of the whole pipeline integrity. The purpose of this article is to introduce the RELAP5 to the simulation and analysis of water hammer situations. Based on the knowledge of the RELAP5 code manuals and some relative documents, the authors utilize RELAP5 to set up an example of water-supply system via an impeller pump to simulate the phenomena of the pump-stopping water hammer. By the simulation of the sample case and the subsequent analysis of the results that the code has provided, we can have a better understand of the knowledge of water hammer as well as the quality of the RELAP5 code when it's used in the water-hammer fields. In the meantime, By comparing the results of the RELAP5 based model with that of other fluid-transient analysis software say, PIPENET. The authors make some conclusions about the peculiarity of RELAP5 when transplanted into water-hammer research and offer several modelling tips when use the code to simulate a water-hammer related case.

Assessment and Application of the ROSE Code for Reactor Outage Thermal-Hydraulic and Safety Analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K.; Dai, L.-C

The currently available tools, such as RELAP5, RETRAN, and others, cannot easily and correctly perform the task of analyzing the system behavior during plant outages. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as midloop operation (MLO) with loss of residual heat removal (RHR), has been developed. Important thermal-hydraulic processes involved during MLO with loss of RHR can be properly simulated by the newly developed reactor outage simulation and evaluation (ROSE) code. The two-region approach with a modified two-fluid model has been adopted to be the theoretical basis of the ROSE code.To verify the analytical modelmore » in the first step, posttest calculations against the integral midloop experiments with loss of RHR have been performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility test data is demonstrated. To further mature the ROSE code in simulating a full-sized pressurized water reactor, assessment against the WGOTHIC code and the Maanshan momentary-loss-of-RHR event has been undertaken. The successfully assessed ROSE code is then applied to evaluate the abnormal operation procedure (AOP) with loss of RHR during MLO (AOP 537.4) for the Maanshan plant. The ROSE code also has been successfully transplanted into the Maanshan training simulator to support operator training. How the simulator was upgraded by the ROSE code for MLO will be presented in the future.« less

2007 international meeting on Reduced Enrichment for Research and Test Reactors (RERTR). Abstracts and available papers presented at the meeting

DOE Office of Scientific and Technical Information (OSTI.GOV)

NONE

2008-07-15

The Meeting papers discuss research and test reactor fuel performance, manufacturing and testing. Some of the main topics are: conversion from HEU to LEU in different reactors and corresponding problems and activities; flux performance and core lifetime analysis with HEU and LEU fuels; physics and safety characteristics; measurement of gamma field parameters in core with LEU fuel; nondestructive analysis of RERTR fuel; thermal hydraulic analysis; fuel interactions; transient analyses and thermal hydraulics for HEU and LEU cores; microstructure research reactor fuels; post irradiation analysis and performance; computer codes and other related problems.

Test case for VVER-1000 complex modeling using MCU and ATHLET

NASA Astrophysics Data System (ADS)

Bahdanovich, R. B.; Bogdanova, E. V.; Gamtsemlidze, I. D.; Nikonov, S. P.; Tikhomirov, G. V.

2017-01-01

The correct modeling of processes occurring in the fuel core of the reactor is very important. In the design and operation of nuclear reactors it is necessary to cover the entire range of reactor physics. Very often the calculations are carried out within the framework of only one domain, for example, in the framework of structural analysis, neutronics (NT) or thermal hydraulics (TH). However, this is not always correct, as the impact of related physical processes occurring simultaneously, could be significant. Therefore it is recommended to spend the coupled calculations. The paper provides test case for the coupled neutronics-thermal hydraulics calculation of VVER-1000 using the precise neutron code MCU and system engineering code ATHLET. The model is based on the fuel assembly (type 2M). Test case for calculation of power distribution, fuel and coolant temperature, coolant density, etc. has been developed. It is assumed that the test case will be used for simulation of VVER-1000 reactor and in the calculation using other programs, for example, for codes cross-verification. The detailed description of the codes (MCU, ATHLET), geometry and material composition of the model and an iterative calculation scheme is given in the paper. Script in PERL language was written to couple the codes.

1DTempPro: analyzing temperature profiles for groundwater/surface-water exchange

USGS Publications Warehouse

Voytek, Emily B.; Drenkelfuss, Anja; Day-Lewis, Frederick D.; Healy, Richard; Lane, John W.; Werkema, Dale D.

2014-01-01

A new computer program, 1DTempPro, is presented for the analysis of vertical one-dimensional (1D) temperature profiles under saturated flow conditions. 1DTempPro is a graphical user interface to the U.S. Geological Survey code Variably Saturated 2-Dimensional Heat Transport (VS2DH), which numerically solves the flow and heat-transport equations. Pre- and postprocessor features allow the user to calibrate VS2DH models to estimate vertical groundwater/surface-water exchange and also hydraulic conductivity for cases where hydraulic head is known.

Parametric analyses of DEMO Divertor using two dimensional transient thermal hydraulic modelling

NASA Astrophysics Data System (ADS)

Domalapally, Phani; Di Caro, Marco

2018-05-01

Among the options considered for cooling of the Plasma facing components of the DEMO reactor, water cooling is a conservative option because of its high heat removal capability. In this work a two-dimensional transient thermal hydraulic code is developed to support the design of the divertor for the projected DEMO reactor with water as a coolant. The mathematical model accounts for transient 2D heat conduction in the divertor section. Temperature-dependent properties are used for more accurate analysis. Correlations for single phase flow forced convection, partially developed subcooled nucleate boiling, fully developed subcooled nucleate boiling and film boiling are used to calculate the heat transfer coefficients on the channel side considering the swirl flow, wherein different correlations found in the literature are compared against each other. Correlation for the Critical Heat Flux is used to estimate its limit for a given flow conditions. This paper then investigates the results of the parametric analysis performed, whereby flow velocity, diameter of the coolant channel, thickness of the coolant pipe, thickness of the armor material, inlet temperature and operating pressure affect the behavior of the divertor under steady or transient heat fluxes. This code will help in understanding the basic parameterś effect on the behavior of the divertor, to achieve a better design from a thermal hydraulic point of view.

Coupled field effects in BWR stability simulations using SIMULATE-3K

DOE Office of Scientific and Technical Information (OSTI.GOV)

Borkowski, J.; Smith, K.; Hagrman, D.

1996-12-31

The SIMULATE-3K code is the transient analysis version of the Studsvik advanced nodal reactor analysis code, SIMULATE-3. Recent developments have focused on further broadening the range of transient applications by refinement of core thermal-hydraulic models and on comparison with boiling water reactor (BWR) stability measurements performed at Ringhals unit 1, during the startups of cycles 14 through 17.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems.

PubMed

Mahadevan, Vijay S; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-08-06

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

PubMed Central

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; Jain, Rajeev; Obabko, Aleksandr; Smith, Michael; Fischer, Paul

2014-01-01

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework. PMID:24982250

Interfacing the Generalized Fluid System Simulation Program with the SINDA/G Thermal Program

NASA Technical Reports Server (NTRS)

Schallhorn, Paul; Palmiter, Christopher; Farmer, Jeffery; Lycans, Randall; Tiller, Bruce

2000-01-01

A general purpose, one dimensional fluid flow code has been interfaced with the thermal analysis program SINDA/G. The flow code, GFSSP, is capable of analyzing steady state and transient flow in a complex network. The flow code is capable of modeling several physical phenomena including compressibility effects, phase changes, body forces (such as gravity and centrifugal) and mixture thermodynamics for multiple species. The addition of GFSSP to SINDA/G provides a significant improvement in convective heat transfer modeling for SINDA/G. The interface development was conducted in two phases. This paper describes the first (which allows for steady and quasi-steady - unsteady solid, steady fluid - conjugate heat transfer modeling). The second (full transient conjugate heat transfer modeling) phase of the interface development will be addressed in a later paper. Phase 1 development has been benchmarked to an analytical solution with excellent agreement. Additional test cases for each development phase demonstrate desired features of the interface. The results of the benchmark case, three additional test cases and a practical application are presented herein.

FY17 Status Report on NEAMS Neutronics Activities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, C. H.; Jung, Y. S.; Smith, M. A.

2017-09-30

Under the U.S. DOE NEAMS program, the high-fidelity neutronics code system has been developed to support the multiphysics modeling and simulation capability named SHARP. The neutronics code system includes the high-fidelity neutronics code PROTEUS, the cross section library and preprocessing tools, the multigroup cross section generation code MC2-3, the in-house meshing generation tool, the perturbation and sensitivity analysis code PERSENT, and post-processing tools. The main objectives of the NEAMS neutronics activities in FY17 are to continue development of an advanced nodal solver in PROTEUS for use in nuclear reactor design and analysis projects, implement a simplified sub-channel based thermal-hydraulic (T/H)more » capability into PROTEUS to efficiently compute the thermal feedback, improve the performance of PROTEUS-MOCEX using numerical acceleration and code optimization, improve the cross section generation tools including MC2-3, and continue to perform verification and validation tests for PROTEUS.« less

MELCOR computer code manuals

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the U.S. Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

TRAC-P1: an advanced best estimate computer program for PWR LOCA analysis. I. Methods, models, user information, and programming details

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-05-01

The Transient Reactor Analysis Code (TRAC) is being developed at the Los Alamos Scientific Laboratory (LASL) to provide an advanced ''best estimate'' predictive capability for the analysis of postulated accidents in light water reactors (LWRs). TRAC-Pl provides this analysis capability for pressurized water reactors (PWRs) and for a wide variety of thermal-hydraulic experimental facilities. It features a three-dimensional treatment of the pressure vessel and associated internals; two-phase nonequilibrium hydrodynamics models; flow-regime-dependent constitutive equation treatment; reflood tracking capability for both bottom flood and falling film quench fronts; and consistent treatment of entire accident sequences including the generation of consistent initial conditions.more » The TRAC-Pl User's Manual is composed of two separate volumes. Volume I gives a description of the thermal-hydraulic models and numerical solution methods used in the code. Detailed programming and user information is also provided. Volume II presents the results of the developmental verification calculations.« less

Conceptual design and cost analysis of hydraulic output unit for 15 kW free-piston Stirling engine

NASA Technical Reports Server (NTRS)

White, M. A.

1982-01-01

A long-life hydraulic converter with unique features was conceptually designed to interface with a specified 15 kW(e) free-piston Stirling engine in a solar thermal dish application. Hydraulic fluid at 34.5 MPa (5000 psi) is produced to drive a conventional hydraulic motor and rotary alternator. Efficiency of the low-maintenance converter design was calculated at 93.5% for a counterbalanced version and 97.0% without the counterbalance feature. If the converter were coupled to a Stirling engine with design parameters more typcial of high-technology Stirling engines, counterbalanced converter efficiency could be increased to 99.6%. Dynamic computer simulation studies were conducted to evaluate performance and system sensitivities. Production costs of the complete Stirling hydraulic/electric power system were evaluated at $6506 which compared with $8746 for an alternative Stirling engine/linear alternator system.

The International Experimental Thermal Hydraulic Systems database – TIETHYS: A new NEA validation tool

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, Upendra S.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary ofmore » appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems (TIETHYS) database and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. https://www.oecd-nea.org/tiethysweb/« less

A New Capability for Nuclear Thermal Propulsion Design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Amiri, Benjamin W.; Nuclear and Radiological Engineering Department, University of Florida, Gainesville, FL 32611; Kapernick, Richard J.

2007-01-30

This paper describes a new capability for Nuclear Thermal Propulsion (NTP) design that has been developed, and presents the results of some analyses performed with this design tool. The purpose of the tool is to design to specified mission and material limits, while maximizing system thrust to weight. The head end of the design tool utilizes the ROCket Engine Transient Simulation (ROCETS) code to generate a system design and system design requirements as inputs to the core analysis. ROCETS is a modular system level code which has been used extensively in the liquid rocket engine industry for many years. Themore » core design tool performs high-fidelity reactor core nuclear and thermal-hydraulic design analysis. At the heart of this process are two codes TMSS-NTP and NTPgen, which together greatly automate the analysis, providing the capability to rapidly produce designs that meet all specified requirements while minimizing mass. A PERL based command script, called CORE DESIGNER controls the execution of these two codes, and checks for convergence throughout the process. TMSS-NTP is executed first, to produce a suite of core designs that meet the specified reactor core mechanical, thermal-hydraulic and structural requirements. The suite of designs consists of a set of core layouts and, for each core layout specific designs that span a range of core fuel volumes. NTPgen generates MCNPX models for each of the core designs from TMSS-NTP. Iterative analyses are performed in NTPgen until a reactor design (fuel volume) is identified for each core layout that meets cold and hot operation reactivity requirements and that is zoned to meet a radial core power distribution requirement.« less

INL Experimental Program Roadmap for Thermal Hydraulic Code Validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Glenn McCreery; Hugh McIlroy

2007-09-01

Advanced computer modeling and simulation tools and protocols will be heavily relied on for a wide variety of system studies, engineering design activities, and other aspects of the Next Generation Nuclear Power (NGNP) Very High Temperature Reactor (VHTR), the DOE Global Nuclear Energy Partnership (GNEP), and light-water reactors. The goal is for all modeling and simulation tools to be demonstrated accurate and reliable through a formal Verification and Validation (V&V) process, especially where such tools are to be used to establish safety margins and support regulatory compliance, or to design a system in a manner that reduces the role ofmore » expensive mockups and prototypes. Recent literature identifies specific experimental principles that must be followed in order to insure that experimental data meet the standards required for a “benchmark” database. Even for well conducted experiments, missing experimental details, such as geometrical definition, data reduction procedures, and manufacturing tolerances have led to poor Benchmark calculations. The INL has a long and deep history of research in thermal hydraulics, especially in the 1960s through 1980s when many programs such as LOFT and Semiscle were devoted to light-water reactor safety research, the EBRII fast reactor was in operation, and a strong geothermal energy program was established. The past can serve as a partial guide for reinvigorating thermal hydraulic research at the laboratory. However, new research programs need to fully incorporate modern experimental methods such as measurement techniques using the latest instrumentation, computerized data reduction, and scaling methodology. The path forward for establishing experimental research for code model validation will require benchmark experiments conducted in suitable facilities located at the INL. This document describes thermal hydraulic facility requirements and candidate buildings and presents examples of suitable validation experiments related to VHTRs, sodium-cooled fast reactors, and light-water reactors. These experiments range from relatively low-cost benchtop experiments for investigating individual phenomena to large electrically-heated integral facilities for investigating reactor accidents and transients.« less

Systematic void fraction studies with RELAP5, FRANCESCA and HECHAN

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stosic, Z.; Preusche, G.

1996-08-01

In enhancing the scope of standard thermal-hydraulic codes applications beyond its capabilities, i.e. coupling with a one and/or three-dimensional kinetics core model, the void fraction, transferred from thermal-hydraulics to the core model, plays a determining role in normal operating range and high core flow, as the generated heat and axial power profiles are direct functions of void distribution in the core. Hence, it is very important to know if the void quality models in the programs which have to be coupled are compatible to allow the interactive exchange of data which are based on these constitutive void-quality relations. The presentedmore » void fraction study is performed in order to give the basis for the conclusion whether a transient core simulation using the RELAP5 void fractions can calculate the axial power shapes adequately. Because of that, the void fractions calculated with RELAP5 are compared with those calculated by BWR safety code for licensing--FRANCESCA and the best estimate model for pre- and post-dryout calculation in BWR heated channel--HECHAN. In addition, a comparison with standard experimental void-quality benchmark tube data is performed for the HECHAN code.« less

NATCRCTR: One-dimensional thermal-hydraulics analysis code for natural-circulation TRIGA reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Rubinaccio, G.

1996-12-31

The Pennsylvania State University nuclear engineering department is evaluating the upgrade of the Reed College (Portland, Oregon) TRIGA reactor from 250 kW to 1 MW in two areas: thermal-hydraulics and steady-state neutronics analysis. This analysis was initiated as a cooperative effort between Penn State and Reed College as a training project for two International Atomic Energy Agency (IAEA) fellows from Ghana. The two Ghanaian IAEA fellows were assisted by G. Rubinaccio, an undergraduate, who undertook the task of writing the new computer programs for the thermal-hydraulic and physics evaluation as a three-credit special design project course. The Reed College TRIGA,more » which has a fixed graphite radial reflector, is cooled by natural circulation, without external cross-flow; whereas, the Penn State Breazeale Reactor has significant crossflow into its sides. To model the Reed TRIGA, the NATCRCTR program has been developed from first principles using the following assumptions: 1. The core is surrounded by the fixed reflector structure, which acts as a one-dimensional channel. 2. The core inlet temperature distribution is constant at the core bottom. 3. The axial heat flux distribution is a chopped cosine shape. 4. The heat transfer in the fuel is primarily in the radial directions. 5. A small gap between the fuel and cladding exists. The NATCRCTR code is used to find the peak centerline fuel, gap, and cladding surface temperatures, based on assumed flux and engineering peaking factors.« less

Specifications for a coupled neutronics thermal-hydraulics SFR test case

NASA Astrophysics Data System (ADS)

Tassone, A.; Smirnov, A. D.; Tikhomirov, G. V.

2017-01-01

Coupling neutronics/thermal-hydraulics calculations for the design of nuclear reactors are a growing trend in the scientific community. This approach allows to properly represent the mutual feedbacks between the neutronic distribution and the thermal-hydraulics properties of the materials composing the reactor, details which are often lost when separate analysis are performed. In this work, a test case for a generation IV sodium-cooled fast reactor (SFR), based on the ASTRID concept developed by CEA, is proposed. Two sub-assemblies (SA) characterized by different fuel enrichment and layout are considered. Specifications for the test case are provided including geometrical data, material compositions, thermo-physical properties and coupling scheme details. Serpent and ANSYS-CFX are used as reference in the description of suitable inputs for the performing of the benchmark, but the use of other code combinations for the purpose of validation of the results is encouraged. The expected outcome of the test case are the axial distribution of volumetric power generation term (q‴), density and temperature for the fuel, the cladding and the coolant.

Updates on HRF Payloads Operations in Columbus ATCS

NASA Technical Reports Server (NTRS)

DePalo, Savino; Wright, Bruce D.; La,e Robert E.; Challis, Simon; Davenport, Robert; Pietrafesa, Donata

2011-01-01

The NASA developed Human Research Facility 1 (HRF1) and Human Research Facility (HRF2) experiment racks have been operating in the European Space Agency (ESA) Columbus module of the International Space Station (ISS) since Summer 2008. The two racks are of the same design. Since the start of operations, unexpected pressure spikes were observed in the Columbus module's thermal-hydraulic system during the racks activation sequence. The root cause of these spikes was identified in the activation command sequence in the Rack Interface Controller (RIC), which controls the flow of thermal-hydraulic system fluid through the rack. A new Common RIC Software (CRS) release fixed the bug and was uploaded on both racks in late 2009. This paper gives a short introduction to the topic, describes the Columbus module countermeasures to mitigate the spikes, describes the ground validation test of the new software, and describes the flight checks performed before and after the final upload. Finally, the new on-orbit test designed to further simplify the racks hydraulic management is presented.

Diffusive deposition of aerosols in Phebus containment during FPT-2 test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kontautas, A.; Urbonavicius, E.

2012-07-01

At present the lumped-parameter codes is the main tool to investigate the complex response of the containment of Nuclear Power Plant in case of an accident. Continuous development and validation of the codes is required to perform realistic investigation of the processes that determine the possible source term of radioactive products to the environment. Validation of the codes is based on the comparison of the calculated results with the measurements performed in experimental facilities. The most extensive experimental program to investigate fission product release from the molten fuel, transport through the cooling circuit and deposition in the containment is performedmore » in PHEBUS test facility. Test FPT-2 performed in this facility is considered for analysis of processes taking place in containment. Earlier performed investigations using COCOSYS code showed that the code could be successfully used for analysis of thermal-hydraulic processes and deposition of aerosols, but there was also noticed that diffusive deposition on the vertical walls does not fit well with the measured results. In the CPA module of ASTEC code there is implemented different model for diffusive deposition, therefore the PHEBUS containment model was transferred from COCOSYS code to ASTEC-CPA to investigate the influence of the diffusive deposition modelling. Analysis was performed using PHEBUS containment model of 16 nodes. The calculated thermal-hydraulic parameters are in good agreement with measured results, which gives basis for realistic simulation of aerosol transport and deposition processes. Performed investigations showed that diffusive deposition model has influence on the aerosol deposition distribution on different surfaces in the test facility. (authors)« less

Improvements and applications of COBRA-TF for stand-alone and coupled LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, M.; Cuervo, D.; Ivanov, K.

2006-07-01

The advanced thermal-hydraulic subchannel code COBRA-TF has been recently improved and applied for stand-alone and coupled LWR core calculations at the Pennsylvania State Univ. in cooperation with AREVA NP GmbH (Germany)) and the Technical Univ. of Madrid. To enable COBRA-TF for academic and industrial applications including safety margins evaluations and LWR core design analyses, the code programming, numerics, and basic models were revised and substantially improved. The code has undergone through an extensive validation, verification, and qualification program. (authors)

New PDC cutters improve drilling efficiency

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mensa-Wilmot, G.

1997-10-27

New polycrystalline diamond compact (PDC) cutters increase penetration rates and cumulative footage through improved abrasion, impact, interface strength, thermal stability, and fatigue characteristics. Studies of formation characterization, vibration analysis, hydraulic layouts, and bit selection continue to improve and expand PDC bit applications. The paper discusses development philosophy, performance characteristics and requirements, Types A, B, and C cutters, and combinations.

Interfacing comprehensive rotorcraft analysis with advanced aeromechanics and vortex wake models

NASA Astrophysics Data System (ADS)

Liu, Haiying

This dissertation describes three aspects of the comprehensive rotorcraft analysis. First, a physics-based methodology for the modeling of hydraulic devices within multibody-based comprehensive models of rotorcraft systems is developed. This newly proposed approach can predict the fully nonlinear behavior of hydraulic devices, and pressure levels in the hydraulic chambers are coupled with the dynamic response of the system. The proposed hydraulic device models are implemented in a multibody code and calibrated by comparing their predictions with test bench measurements for the UH-60 helicopter lead-lag damper. Predicted peak damping forces were found to be in good agreement with measurements, while the model did not predict the entire time history of damper force to the same level of accuracy. The proposed model evaluates relevant hydraulic quantities such as chamber pressures, orifice flow rates, and pressure relief valve displacements. This model could be used to design lead-lag dampers with desirable force and damping characteristics. The second part of this research is in the area of computational aeroelasticity, in which an interface between computational fluid dynamics (CFD) and computational structural dynamics (CSD) is established. This interface enables data exchange between CFD and CSD with the goal of achieving accurate airloads predictions. In this work, a loose coupling approach based on the delta-airloads method is developed in a finite-element method based multibody dynamics formulation, DYMORE. To validate this aerodynamic interface, a CFD code, OVERFLOW-2, is loosely coupled with a CSD program, DYMORE, to compute the airloads of different flight conditions for Sikorsky UH-60 aircraft. This loose coupling approach has good convergence characteristics. The predicted airloads are found to be in good agreement with the experimental data, although not for all flight conditions. In addition, the tight coupling interface between the CFD program, OVERFLOW-2, and the CSD program, DYMORE, is also established. The ability to accurately capture the wake structure around a helicopter rotor is crucial for rotorcraft performance analysis. In the third part of this thesis, a new representation of the wake vortex structure based on Non-Uniform Rational B-Spline (NURBS) curves and surfaces is proposed to develop an efficient model for prescribed and free wakes. NURBS curves and surfaces are able to represent complex shapes with remarkably little data. The proposed formulation has the potential to reduce the computational cost associated with the use of Helmholtz's law and the Biot-Savart law when calculating the induced flow field around the rotor. An efficient free-wake analysis will considerably decrease the computational cost of comprehensive rotorcraft analysis, making the approach more attractive to routine use in industrial settings.

MELCOR computer code manuals: Primer and user`s guides, Version 1.8.3 September 1994. Volume 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Summers, R.M.; Cole, R.K. Jr.; Smith, R.C.

1995-03-01

MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor nuclear power plants. MELCOR is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission as a second-generation plant risk assessment tool and the successor to the Source Term Code Package. A broad spectrum of severe accident phenomena in both boiling and pressurized water reactors is treated in MELCOR in a unified framework. These include: thermal-hydraulic response in the reactor coolant system, reactor cavity, containment, and confinement buildings; core heatup, degradation, and relocation; core-concrete attack; hydrogen production, transport, andmore » combustion; fission product release and transport; and the impact of engineered safety features on thermal-hydraulic and radionuclide behavior. Current uses of MELCOR include estimation of severe accident source terms and their sensitivities and uncertainties in a variety of applications. This publication of the MELCOR computer code manuals corresponds to MELCOR 1.8.3, released to users in August, 1994. Volume 1 contains a primer that describes MELCOR`s phenomenological scope, organization (by package), and documentation. The remainder of Volume 1 contains the MELCOR Users` Guides, which provide the input instructions and guidelines for each package. Volume 2 contains the MELCOR Reference Manuals, which describe the phenomenological models that have been implemented in each package.« less

Coupled calculation of the radiological release and the thermal-hydraulic behavior of a 3-loop PWR after a SGTR by means of the code RELAP5

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Hove, W.; Van Laeken, K.; Bartsoen, L.

1995-09-01

To enable a more realistic and accurate calculation of the radiological consequences of a SGTR, a fission product transport model was developed. As the radiological releases strongly depend on the thermal-hydraulic transient, the model was included in the RELAP5 input decks of the Belgian NPPs. This enables the coupled calculation of the thermal-hydraulic transient and the radiological release. The fission product transport model tracks the concentration of the fission products in the primary circuit, in each of the SGs as well as in the condenser. This leads to a system of 6 coupled, first order ordinary differential equations with timemore » dependent coefficients. Flashing, scrubbing, atomisation and dry out of the break flow are accounted for. Coupling with the thermal-hydraulic calculation and correct modelling of the break position enables an accurate calculation of the mixture level above the break. Pre- and post-accident spiking in the primary circuit are introduced. The transport times in the FW-system and the SG blowdown system are also taken into account, as is the decontaminating effect of the primary make-up system and of the SG blowdown system. Physical input parameters such as the partition coefficients, half life times and spiking coefficients are explicitly introduced so that the same model can be used for iodine, caesium and noble gases.« less

Modeling of Thermal Barrier Coatings

NASA Technical Reports Server (NTRS)

Ferguson, B. L.; Petrus, G. J.; Krauss, T. M.

1992-01-01

The project examined the effectiveness of studying the creep behavior of thermal barrier coating system through the use of a general purpose, large strain finite element program, NIKE2D. Constitutive models implemented in this code were applied to simulate thermal-elastic and creep behavior. Four separate ceramic-bond coat interface geometries were examined in combination with a variety of constitutive models and material properties. The reason for focusing attention on the ceramic-bond coat interface is that prior studies have shown that cracking occurs in the ceramic near interface features which act as stress concentration points. The model conditions examined include: (1) two bond coat coefficient of thermal expansion curves; (2) the creep coefficient and creep exponent of the bond coat for steady state creep; (3) the interface geometry; and (4) the material model employed to represent the bond coat, ceramic, and superalloy base.

1DTempPro: analyzing temperature profiles for groundwater/surface-water exchange.

PubMed

Voytek, Emily B; Drenkelfuss, Anja; Day-Lewis, Frederick D; Healy, Richard; Lane, John W; Werkema, Dale

2014-01-01

A new computer program, 1DTempPro, is presented for the analysis of vertical one-dimensional (1D) temperature profiles under saturated flow conditions. 1DTempPro is a graphical user interface to the U.S. Geological Survey code Variably Saturated 2-Dimensional Heat Transport (VS2DH), which numerically solves the flow and heat-transport equations. Pre- and postprocessor features allow the user to calibrate VS2DH models to estimate vertical groundwater/surface-water exchange and also hydraulic conductivity for cases where hydraulic head is known. Published 2013. This article is a U.S. Government work and is in the public domain in the USA.

Transient analysis of ”2 inch Direct Vessel Injection line break” in SPES-2 facility by using TRACE code

NASA Astrophysics Data System (ADS)

D'Amico, S.; Lombardo, C.; Moscato, I.; Polidori, M.; Vella, G.

2015-11-01

In the past few decades a lot of theoretical and experimental researches have been done to understand the physical phenomena characterizing nuclear accidents. In particular, after the Three Miles Island accident, several reactors have been designed to handle successfully LOCA events. This paper presents a comparison between experimental and numerical results obtained for the “2 inch Direct Vessel Injection line break” in SPES-2. This facility is an integral test facility built in Piacenza at the SIET laboratories and simulating the primary circuit, the relevant parts of the secondary circuits and the passive safety systems typical of the AP600 nuclear power plant. The numerical analysis here presented was performed by using TRACE and CATHARE thermal-hydraulic codes with the purpose of evaluating their prediction capability. The main results show that the TRACE model well predicts the overall behaviour of the plant during the transient, in particular it is able to simulate the principal thermal-hydraulic phenomena related to all passive safety systems. The performance of the presented CATHARE noding has suggested some possible improvements of the model.

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

A fully-implicit high-order system thermal-hydraulics model for advanced non-LWR safety analyses

DOE PAGES

Hu, Rui

2016-11-19

An advanced system analysis tool is being developed for advanced reactor safety analysis. This paper describes the underlying physics and numerical models used in the code, including the governing equations, the stabilization schemes, the high-order spatial and temporal discretization schemes, and the Jacobian Free Newton Krylov solution method. The effects of the spatial and temporal discretization schemes are investigated. Additionally, a series of verification test problems are presented to confirm the high-order schemes. Furthermore, it is demonstrated that the developed system thermal-hydraulics model can be strictly verified with the theoretical convergence rates, and that it performs very well for amore » wide range of flow problems with high accuracy, efficiency, and minimal numerical diffusions.« less

CFD and Neutron codes coupling on a computational platform

NASA Astrophysics Data System (ADS)

Cerroni, D.; Da Vià, R.; Manservisi, S.; Menghini, F.; Scardovelli, R.

2017-01-01

In this work we investigate the thermal-hydraulics behavior of a PWR nuclear reactor core, evaluating the power generation distribution taking into account the local temperature field. The temperature field, evaluated using a self-developed CFD module, is exchanged with a neutron code, DONJON-DRAGON, which updates the macroscopic cross sections and evaluates the new neutron flux. From the updated neutron flux the new peak factor is evaluated and the new temperature field is computed. The exchange of data between the two codes is obtained thanks to their inclusion into the computational platform SALOME, an open-source tools developed by the collaborative project NURESAFE. The numerical libraries MEDmem, included into the SALOME platform, are used in this work, for the projection of computational fields from one problem to another. The two problems are driven by a common supervisor that can access to the computational fields of both systems, in every time step, the temperature field, is extracted from the CFD problem and set into the neutron problem. After this iteration the new power peak factor is projected back into the CFD problem and the new time step can be computed. Several computational examples, where both neutron and thermal-hydraulics quantities are parametrized, are finally reported in this work.

Heat Transfer Computations of Internal Duct Flows With Combined Hydraulic and Thermal Developing Length

NASA Technical Reports Server (NTRS)

Wang, C. R.; Towne, C. E.; Hippensteele, S. A.; Poinsatte, P. E.

1997-01-01

This study investigated the Navier-Stokes computations of the surface heat transfer coefficients of a transition duct flow. A transition duct from an axisymmetric cross section to a non-axisymmetric cross section, is usually used to connect the turbine exit to the nozzle. As the gas turbine inlet temperature increases, the transition duct is subjected to the high temperature at the gas turbine exit. The transition duct flow has combined development of hydraulic and thermal entry length. The design of the transition duct required accurate surface heat transfer coefficients. The Navier-Stokes computational method could be used to predict the surface heat transfer coefficients of a transition duct flow. The Proteus three-dimensional Navier-Stokes numerical computational code was used in this study. The code was first studied for the computations of the turbulent developing flow properties within a circular duct and a square duct. The code was then used to compute the turbulent flow properties of a transition duct flow. The computational results of the surface pressure, the skin friction factor, and the surface heat transfer coefficient were described and compared with their values obtained from theoretical analyses or experiments. The comparison showed that the Navier-Stokes computation could predict approximately the surface heat transfer coefficients of a transition duct flow.

High-resolution coupled physics solvers for analysing fine-scale nuclear reactor design problems

DOE PAGES

Mahadevan, Vijay S.; Merzari, Elia; Tautges, Timothy; ...

2014-06-30

An integrated multi-physics simulation capability for the design and analysis of current and future nuclear reactor models is being investigated, to tightly couple neutron transport and thermal-hydraulics physics under the SHARP framework. Over several years, high-fidelity, validated mono-physics solvers with proven scalability on petascale architectures have been developed independently. Based on a unified component-based architecture, these existing codes can be coupled with a mesh-data backplane and a flexible coupling-strategy-based driver suite to produce a viable tool for analysts. The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in ordermore » to reduce the overall numerical uncertainty while leveraging available computational resources. Finally, the coupling methodology and software interfaces of the framework are presented, along with verification studies on two representative fast sodium-cooled reactor demonstration problems to prove the usability of the SHARP framework.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Christon, Mark A.; Baksoi, Jozsef; Barnett, Nathan

This report describes the work carried out for completion of the Thermal Hydraulics Methods (THM) Level 2 Milestone THM.CFD.P5.01 for the Consortium for Advanced Simulation of Light Water Reactors (CASL). This milestone focused primarily on the initial integration of Hydra-TH in VERA. The primary objective for this milestone was the integration of Hydra-TH as a standalone executable in VERA. A series of code extensions/modifications have been made to Hydra-TH to facilitate integration of Hydra-TH in VERA and to permit future tighter integration and physics coupling. A total of 61 serial and 64 parallel regression tests have been supplied with Hydra-TH.more » These tests are are being executed in the TriBITS environment. Once the VERA team enables the full suite of tests, the results can be posted to the VERA CDash site. Future work will consider the use of the LIME 2.0 interface for tighter integration in VERA with additional efforts focused on multiphysics coupling with radiation transport, fuel performance, and solid/structural mechanics.« less

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

DOE Office of Scientific and Technical Information (OSTI.GOV)

Blyth, Taylor S.; Avramova, Maria

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics- based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR)more » cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal- hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.« less

Development and Implementation of CFD-Informed Models for the Advanced Subchannel Code CTF

NASA Astrophysics Data System (ADS)

Blyth, Taylor S.

The research described in this PhD thesis contributes to the development of efficient methods for utilization of high-fidelity models and codes to inform low-fidelity models and codes in the area of nuclear reactor core thermal-hydraulics. The objective is to increase the accuracy of predictions of quantities of interests using high-fidelity CFD models while preserving the efficiency of low-fidelity subchannel core calculations. An original methodology named Physics-based Approach for High-to-Low Model Information has been further developed and tested. The overall physical phenomena and corresponding localized effects, which are introduced by the presence of spacer grids in light water reactor (LWR) cores, are dissected in corresponding four building basic processes, and corresponding models are informed using high-fidelity CFD codes. These models are a spacer grid-directed cross-flow model, a grid-enhanced turbulent mixing model, a heat transfer enhancement model, and a spacer grid pressure loss model. The localized CFD-models are developed and tested using the CFD code STAR-CCM+, and the corresponding global model development and testing in sub-channel formulation is performed in the thermal-hydraulic subchannel code CTF. The improved CTF simulations utilize data-files derived from CFD STAR-CCM+ simulation results covering the spacer grid design desired for inclusion in the CTF calculation. The current implementation of these models is examined and possibilities for improvement and further development are suggested. The validation experimental database is extended by including the OECD/NRC PSBT benchmark data. The outcome is an enhanced accuracy of CTF predictions while preserving the computational efficiency of a low-fidelity subchannel code.

MATCHED-INDEX-OF-REFRACTION FLOW FACILITY FOR FUNDAMENTAL AND APPLIED RESEARCH

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piyush Sabharwall; Carl Stoots; Donald M. McEligot

2014-11-01

Significant challenges face reactor designers with regard to thermal hydraulic design and associated modeling for advanced reactor concepts. Computational thermal hydraulic codes solve only a piece of the core. There is a need for a whole core dynamics system code with local resolution to investigate and understand flow behavior with all the relevant physics and thermo-mechanics. The matched index of refraction (MIR) flow facility at Idaho National Laboratory (INL) has a unique capability to contribute to the development of validated computational fluid dynamics (CFD) codes through the use of state-of-the-art optical measurement techniques, such as Laser Doppler Velocimetry (LDV) andmore » Particle Image Velocimetry (PIV). PIV is a non-intrusive velocity measurement technique that tracks flow by imaging the movement of small tracer particles within a fluid. At the heart of a PIV calculation is the cross correlation algorithm, which is used to estimate the displacement of particles in some small part of the image over the time span between two images. Generally, the displacement is indicated by the location of the largest peak. To quantify these measurements accurately, sophisticated processing algorithms correlate the locations of particles within the image to estimate the velocity (Ref. 1). Prior to use with reactor deign, the CFD codes have to be experimentally validated, which requires rigorous experimental measurements to produce high quality, multi-dimensional flow field data with error quantification methodologies. Computational thermal hydraulic codes solve only a piece of the core. There is a need for a whole core dynamics system code with local resolution to investigate and understand flow behavior with all the relevant physics and thermo-mechanics. Computational techniques with supporting test data may be needed to address the heat transfer from the fuel to the coolant during the transition from turbulent to laminar flow, including the possibility of an early laminarization of the flow (Refs. 2 and 3) (laminarization is caused when the coolant velocity is theoretically in the turbulent regime, but the heat transfer properties are indicative of the coolant velocity being in the laminar regime). Such studies are complicated enough that computational fluid dynamics (CFD) models may not converge to the same conclusion. Thus, experimentally scaled thermal hydraulic data with uncertainties should be developed to support modeling and simulation for verification and validation activities. The fluid/solid index of refraction matching technique allows optical access in and around geometries that would otherwise be impossible while the large test section of the INL system provides better spatial and temporal resolution than comparable facilities. Benchmark data for assessing computational fluid dynamics can be acquired for external flows, internal flows, and coupled internal/external flows for better understanding of physical phenomena of interest. The core objective of this study is to describe MIR and its capabilities, and mention current development areas for uncertainty quantification, mainly the uncertainty surface method and cross-correlation method. Using these methods, it is anticipated to establish a suitable approach to quantify PIV uncertainty for experiments performed in the MIR.« less

Thermal-hydraulic posttest analysis for the ANL/MCTF 360/sup 0/ model heat-exchanger water test under mixed convection. [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, C.I.; Sha, W.T.; Kasza, K.E.

As a result of the uncertainties in the understanding of the influence of thermal-buoyancy effects on the flow and heat transfer in Liquid Metal Fast Breeder Reactor heat exchangers and steam generators under off-normal operating conditions, an extensive experimental program is being conducted at Argonne National Laboratory to eliminate these uncertainties. Concurrently, a parallel analytical effort is also being pursued to develop a three-dimensional transient computer code (COMMIX-IHX) to study and predict heat exchanger performance under mixed, forced, and free convection conditions. This paper presents computational results from a heat exchanger simulation and compares them with the results from amore » test case exhibiting strong thermal buoyancy effects. Favorable agreement between experiment and code prediction is obtained.« less

Posttest calculations of bundle quench test CORA-13 with ATHLET-CD

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bestele, J.; Trambauer, K.; Schubert, J.D.

Gesellschaft fuer Anlagen- und Reaktorsicherheit is developing, in cooperation with the Institut fuer Kernenergetik und Energiesysteme, Stuttgart, the system code Analysis of Thermalhydraulics of Leaks and Transients with Core Degradation (ATHLET-CD). The code consists of detailed models of the thermal hydraulics of the reactor coolant system. This thermo-fluid dynamics module is coupled with modules describing the early phase of the core degradation, like cladding deformation, oxidation and melt relocation, and the release and transport of fission products. The assessment of the code is being done by the analysis of separate effect tests, integral tests, and plant events. The code willmore » be applied to the verification of severe accident management procedures. The out-of-pile test CORA-13 was conducted by Forschungszentrum Karlsruhe in their CORA test facility. The test consisted of two phases, a heatup phase and a quench phase. At the beginning of the quench phase, a sharp peak in the hydrogen generation rate was observed. Both phases of the test have been calculated with the system code ATHLET-CD. Special efforts have been made to simulate the heat losses and the flow distribution in the test facility and the thermal hydraulics during the quench phase. In addition to previous calculations, the material relocation and the quench phase have been modeled. The temperature increase during the heatup phase, the starting time of the temperature escalation, and the maximum temperatures have been calculated correctly. At the beginning of the quench phase, an increased hydrogen generation rate has been calculated as measured in the experiment.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K; Sung, Yixing; Kucukboyaci, Vefa

The Virtual Environment for Reactor Applications core simulator (VERA-CS) being developed by the Consortium for the Advanced Simulation of Light Water Reactors (CASL) includes coupled neutronics, thermal-hydraulics, and fuel temperature components with an isotopic depletion capability. The neutronics capability employed is based on MPACT, a three-dimensional (3-D) whole core transport code. The thermal-hydraulics and fuel temperature models are provided by the COBRA-TF (CTF) subchannel code. As part of the CASL development program, the VERA-CS (MPACT/CTF) code system was applied to model and simulate reactor core response with respect to departure from nucleate boiling ratio (DNBR) at the limiting time stepmore » of a postulated pressurized water reactor (PWR) main steamline break (MSLB) event initiated at the hot zero power (HZP), either with offsite power available and the reactor coolant pumps in operation (high-flow case) or without offsite power where the reactor core is cooled through natural circulation (low-flow case). The VERA-CS simulation was based on core boundary conditions from the RETRAN-02 system transient calculations and STAR-CCM+ computational fluid dynamics (CFD) core inlet distribution calculations. The evaluation indicated that the VERA-CS code system is capable of modeling and simulating quasi-steady state reactor core response under the steamline break (SLB) accident condition, the results are insensitive to uncertainties in the inlet flow distributions from the CFD simulations, and the high-flow case is more DNB limiting than the low-flow case.« less

Code manual for CONTAIN 2.0: A computer code for nuclear reactor containment analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Murata, K.K.; Williams, D.C.; Griffith, R.O.

1997-12-01

The CONTAIN 2.0 computer code is an integrated analysis tool used for predicting the physical conditions, chemical compositions, and distributions of radiological materials inside a containment building following the release of material from the primary system in a light-water reactor accident. It can also predict the source term to the environment. CONTAIN 2.0 is intended to replace the earlier CONTAIN 1.12, which was released in 1991. The purpose of this Code Manual is to provide full documentation of the features and models in CONTAIN 2.0. Besides complete descriptions of the models, this Code Manual provides a complete description of themore » input and output from the code. CONTAIN 2.0 is a highly flexible and modular code that can run problems that are either quite simple or highly complex. An important aspect of CONTAIN is that the interactions among thermal-hydraulic phenomena, aerosol behavior, and fission product behavior are taken into account. The code includes atmospheric models for steam/air thermodynamics, intercell flows, condensation/evaporation on structures and aerosols, aerosol behavior, and gas combustion. It also includes models for reactor cavity phenomena such as core-concrete interactions and coolant pool boiling. Heat conduction in structures, fission product decay and transport, radioactive decay heating, and the thermal-hydraulic and fission product decontamination effects of engineered safety features are also modeled. To the extent possible, the best available models for severe accident phenomena have been incorporated into CONTAIN, but it is intrinsic to the nature of accident analysis that significant uncertainty exists regarding numerous phenomena. In those cases, sensitivity studies can be performed with CONTAIN by means of user-specified input parameters. Thus, the code can be viewed as a tool designed to assist the knowledge reactor safety analyst in evaluating the consequences of specific modeling assumptions.« less

Aerothermo-Structural Analysis of Low Cost Composite Nozzle/Inlet Components

NASA Technical Reports Server (NTRS)

Shivakumar, Kuwigai; Challa, Preeli; Sree, Dave; Reddy, D.

1999-01-01

This research is a cooperative effort among the Turbomachinery and Propulsion Division of NASA Glenn, CCMR of NC A&T State University, and the Tuskegee University. The NC A&T is the lead center and Tuskegee University is the participating institution. Objectives of the research were to develop an integrated aerodynamic, thermal and structural analysis code for design of aircraft engine components, such as, nozzles and inlets made of textile composites; conduct design studies on typical inlets for hypersonic transportation vehicles and setup standards test examples and finally manufacture a scaled down composite inlet. These objectives are accomplished through the following seven tasks: (1) identify the relevant public domain codes for all three types of analysis; (2) evaluate the codes for the accuracy of results and computational efficiency; (3) develop aero-thermal and thermal structural mapping algorithms; (4) integrate all the codes into one single code; (5) write a graphical user interface to improve the user friendliness of the code; (6) conduct test studies for rocket based combined-cycle engine inlet; and finally (7) fabricate a demonstration inlet model using textile preform composites. Tasks one, two and six are being pursued. Selected and evaluated NPARC for flow field analysis, CSTEM for in-depth thermal analysis of inlets and nozzles and FRAC3D for stress analysis. These codes have been independently verified for accuracy and performance. In addition, graphical user interface based on micromechanics analysis for laminated as well as textile composites was developed. Demonstration of this code will be made at the conference. A rocket based combined cycle engine was selected for test studies. Flow field analysis of various inlet geometries were studied. Integration of codes is being continued. The codes developed are being applied to a candidate example of trailblazer engine proposed for space transportation. A successful development of the code will provide a simpler, faster and user-friendly tool for conducting design studies of aircraft and spacecraft engines, applicable in high speed civil transport and space missions.

Coupled Physics Environment (CouPE) library - Design, Implementation, and Release

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mahadevan, Vijay S.

Over several years, high fidelity, validated mono-­physics solvers with proven scalability on peta-­scale architectures have been developed independently. Based on a unified component-­based architecture, these existing codes can be coupled with a unified mesh-­data backplane and a flexible coupling-­strategy-­based driver suite to produce a viable tool for analysts. In this report, we present details on the design decisions and developments on CouPE, an acronym that stands for Coupled Physics Environment that orchestrates a coupled physics solver through the interfaces exposed by MOAB array-­based unstructured mesh, both of which are part of SIGMA (Scalable Interfaces for Geometry and Mesh-­Based Applications) toolkit.more » The SIGMA toolkit contains libraries that enable scalable geometry and unstructured mesh creation and handling in a memory and computationally efficient implementation. The CouPE version being prepared for a full open-­source release along with updated documentation will contain several useful examples that will enable users to start developing their applications natively using the native MOAB mesh and couple their models to existing physics applications to analyze and solve real world problems of interest. An integrated multi-­physics simulation capability for the design and analysis of current and future nuclear reactor models is also being investigated as part of the NEAMS RPL, to tightly couple neutron transport, thermal-­hydraulics and structural mechanics physics under the SHARP framework. This report summarizes the efforts that have been invested in CouPE to bring together several existing physics applications namely PROTEUS (neutron transport code), Nek5000 (computational fluid-dynamics code) and Diablo (structural mechanics code). The goal of the SHARP framework is to perform fully resolved coupled physics analysis of a reactor on heterogeneous geometry, in order to reduce the overall numerical uncertainty while leveraging available computational resources. The design of CouPE along with motivations that led to implementation choices are also discussed. The first release of the library will be different from the current version of the code that integrates the components in SHARP and explanation on the need for forking the source base will also be provided. Enhancements in the functionality and improved user guides will be available as part of the release. CouPE v0.1 is scheduled for an open-­source release in December 2014 along with SIGMA v1.1 components that provide support for language-agnostic mesh loading, traversal and query interfaces along with scalable solution transfer of fields between different physics codes. The coupling methodology and software interfaces of the library are presented, along with verification studies on two representative fast sodium-­cooled reactor demonstration problems to prove the usability of the CouPE library.« less

Study of premixing phase of steam explosion with JASMINE code in ALPHA program

DOE Office of Scientific and Technical Information (OSTI.GOV)

Moriyama, Kiyofumi; Yamano, Norihiro; Maruyama, Yu

Premixing phase of steam explosion has been studied in ALPHA Program at Japan Atomic Energy Research Institute (JAERI). An analytical model to simulate the premixing phase, JASMINE (JAERI Simulator for Multiphase Interaction and Explosion), has been developed based on a multi-dimensional multi-phase thermal hydraulics code MISTRAL (by Fuji Research Institute Co.). The original code was extended to simulate the physics in the premixing phenomena. The first stage of the code validation was performed by analyzing two mixing experiments with solid particles and water: the isothermal experiment by Gilbertson et al. (1992) and the hot particle experiment by Angelini et al.more » (1993) (MAGICO). The code predicted reasonably well the experiments. Effectiveness of the TVD scheme employed in the code was also demonstrated.« less

Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

DOE PAGES

Hu, Po; Wilson, Paul

2014-01-01

The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in themore » code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.« less

Assessment of the MHD capability in the ATHENA code using data from the ALEX facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roth, P.A.

1989-03-01

The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code is a system transient analysis code with multi-loop, multi-fluid capabilities, which is available to the fusion community at the National Magnetic Fusion Energy Computing Center (NMFECC). The work reported here assesses the ATHENA magnetohydrodynamic (MHD) pressure drop model for liquid metals flowing through a strong magnetic field. An ATHENA model was developed for two simple geometry, adiabatic test sections used in the Argonne Liquid Metal Experiment (ALEX) at Argonne National Laboratory (ANL). The pressure drops calculated by ATHENA agreed well with the experimental results from the ALEX facility.

Interface requirements for coupling a containment code to a reactor system thermal hydraulic codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Baratta, A.J.

1997-07-01

To perform a complete analysis of a reactor transient, not only the primary system response but the containment response must also be accounted for. Such transients and accidents as a loss of coolant accident in both pressurized water and boiling water reactors and inadvertent operation of safety relief valves all challenge the containment and may influence flows because of containment feedback. More recently, the advanced reactor designs put forth by General Electric and Westinghouse in the US and by Framatome and Seimens in Europe rely on the containment to act as the ultimate heat sink. Techniques used by analysts andmore » engineers to analyze the interaction of the containment and the primary system were usually iterative in nature. Codes such as RELAP or RETRAN were used to analyze the primary system response and CONTAIN or CONTEMPT the containment response. The analysis was performed by first running the system code and representing the containment as a fixed pressure boundary condition. The flows were usually from the primary system to the containment initially and generally under choked conditions. Once the mass flows and timing are determined from the system codes, these conditions were input into the containment code. The resulting pressures and temperatures were then calculated and the containment performance analyzed. The disadvantage of this approach becomes evident when one performs an analysis of a rapid depressurization or a long term accident sequence in which feedback from the containment can occur. For example, in a BWR main steam line break transient, the containment heats up and becomes a source of energy for the primary system. Recent advances in programming and computer technology are available to provide an alternative approach. The author and other researchers have developed linkage codes capable of transferring data between codes at each time step allowing discrete codes to be coupled together.« less

RERTR-9 Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez

2011-05-01

The RERTR-9 experiment was designed to test the effect of modified fuel/clad interfaces in monolithic fuel plates and to demonstrate that the addition of Si to the matrix material in dispersion plates continued to be effective at high loading (~8.5 g U/cc). Several monolithic fuel plates were fabricated by Hot Isostatic Pressing (HIP) and Friction Bonding (FB) with thin layers of Si inserted and by HIP with a Zr diffusion barrier between the fuel and cladding. Si was applied to the interface by thermal spray of Al Si mixtures and by the insertion of thin Si-rich Al alloy foil betweenmore » the fuel/clad interface. The dispersion fuel plates were fabricated by semi-standard rolling techniques (the reduction by rolling was lowered to limit fabrication defects). Matrix materials consisted of Al-Si alloys and mixtures with various levels of Si. The following report summarizes the life of the RERTR-9A/B experiment through end of irradiation, including as-run neutronic analysis, thermal analysis and hydraulic testing results.« less

Factors controlling the configuration of the fresh-saline water interface in the Dead Sea coastal aquifers: Synthesis of TDEM surveys and numerical groundwater modeling

USGS Publications Warehouse

Yechieli, Y.; Kafri, U.; Goldman, M.; Voss, C.I.

2001-01-01

TDEM (time domain electromagnetic) traverses in the Dead Sea (DS) coastal aquifer help to delineate the configuration of the interrelated fresh-water and brine bodies and the interface in between. A good linear correlation exists between the logarithm of TDEM resistivity and the chloride concentration of groundwater, mostly in the higher salinity range, close to that of the DS brine. In this range, salinity is the most important factor controlling resistivity. The configuration of the fresh-saline water interface is dictated by the hydraulic gradient, which is controlled by a number of hydrological factors. Three types of irregularities in the configuration of fresh-water and saline-water bodies were observed in the study area: 1. Fresh-water aquifers underlying more saline ones ("Reversal") in a multi-aquifer system. 2. "Reversal" and irregular residual saline-water bodies related to historical, frequently fluctuating DS base level and respective interfaces, which have not undergone complete flushing. A rough estimate of flushing rates may be obtained based on knowledge of the above fluctuations. The occurrence of salt beds is also a factor affecting the interface configuration. 3. The interface steepens towards and adjacent to the DS Rift fault zone. Simulation analysis with a numerical, variable-density flow model, using the US Geological Survey's SUTRA code, indicates that interface steep- ening may result from a steep water-level gradient across the zone, possibly due to a low hydraulic conductivity in the immediate vicinity of the fault.

Numerical Simulation and Analyses of the Loss of Feedwater Transient at the Unit 4 of Kola NPP

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stevanovic, Vladimir D.; Stosic, Zoran V.; Kiera, Michael

2002-07-01

A three-dimensional numerical simulation of the loss-of-feed water transient at the horizontal steam generator of the Kola nuclear power plant is performed. Presented numerical results show transient change of integral steam generator parameters, such as steam generation rate, water mass inventory, outlet reactor coolant temperature, as well as detailed distribution of shell side thermal-hydraulic parameters: swell and collapsed levels, void fraction distributions, mass flux vectors, etc. Numerical results are compared with measurements at the Kola NPP. The agreement is satisfactory, while differences are close to or below the measurement uncertainties. Obtained numerical results are the first ones that give completemore » insight into the three-dimensional and transient horizontal steam generator thermal-hydraulics. Also, the presented results serve as benchmark tests for the assessment and further improvement of one-dimensional models of horizontal steam generator built with safety codes. (authors)« less

Heat, Mass and Aerosol Transfers in Spray Conditions for Containment Application

NASA Astrophysics Data System (ADS)

Porcheron, Emmanuel; Lemaitre, Pascal; Nuboer, Amandine; Vendel, Jacques

TOSQAN is an experimental program undertaken by the Institut de Radioprotection et de Surété Nucleaire (IRSN) in order to perform thermal hydraulic containment studies. The TOSQAN facility is a large enclosure devoted to simulating typical accidental thermal hydraulic flow conditions in nuclear Pressurized Water Reactor (PWR) containment. The TOSQAN facility, which is highly instrumented with non-intrusive optical diagnostics, is particularly adapted to nuclear safety CFD code validation. The present work is devoted to studying the interaction of a water spray injection used as a mitigation means in order to reduce the gas pressure and temperature in the containment, to produce gases mixing and washout of fission products. In order to have a better understanding of heat and mass transfers between spray droplets and the gas mixture, and to analyze mixing effects due to spray activation, we performed detailed characterization of the two-phase flow.

Interface Shape and Convection During Solidification and Melting of Succinonitrile

NASA Technical Reports Server (NTRS)

Degroh, Henry C., III; Lindstrom, Tiffany

1994-01-01

An experimental study was conducted of the crystal growth of succinonitrile during solidification, melting, and no-growth conditions using a horizontal Bridgman furnace and square glass ampoule. For use as input boundary conditions to numerical codes, thermal profiles on the outside of the ampoule at five locations around its periphery were measured along the ampoule's length. Temperatures inside the ampoule were also measured. The shapes of the s/l interface in various two dimensional planes were quantitatively determined. Though interfaces were nondendritic and noncellular, they were not flat, but were highly curved and symmetric in only one unique longitudinal y-z plane (at x=O). The shapes of the interface were dominated by the primary longitudinal flow cell characteristic of shallow cavity flow in horizontal Bridgman; this flow cell was driven by the imposed furnace temperature gradient and caused a 'radical' thermal gradient such that the upper half of the ampoule was hotter than the bottom half. We believe that due to the strong convection, the release of latent heat does not significantly influence the thermal conditions near the interface. We hope that the interface shape and thermal data presented in this paper can be used to optimize crystal growth processes and validate numerical models.

Benchmark Simulations of the Thermal-Hydraulic Responses during EBR-II Inherent Safety Tests using SAM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hu, Rui; Sumner, Tyler S.

2016-04-17

An advanced system analysis tool SAM is being developed for fast-running, improved-fidelity, and whole-plant transient analyses at Argonne National Laboratory under DOE-NE’s Nuclear Energy Advanced Modeling and Simulation (NEAMS) program. As an important part of code development, companion validation activities are being conducted to ensure the performance and validity of the SAM code. This paper presents the benchmark simulations of two EBR-II tests, SHRT-45R and BOP-302R, whose data are available through the support of DOE-NE’s Advanced Reactor Technology (ART) program. The code predictions of major primary coolant system parameter are compared with the test results. Additionally, the SAS4A/SASSYS-1 code simulationmore » results are also included for a code-to-code comparison.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bolstad, J.W.; Haarman, R.A.

The results of two transients involving the loss of a steam generator in a single-pass, steam generator, pressurized water reactor have been analyzed using a state-of-the-art, thermal-hydraulic computer code. Computed results include the formation of a steam bubble in the core while the pressurizer is solid. Calculations show that continued injection of high pressure water would have stopped the scenario. These are similar to the happenings at Three Mile Island.

Joint three-dimensional inversion of coupled groundwater flow and heat transfer based on automatic differentiation: sensitivity calculation, verification, and synthetic examples

NASA Astrophysics Data System (ADS)

Rath, V.; Wolf, A.; Bücker, H. M.

2006-10-01

Inverse methods are useful tools not only for deriving estimates of unknown parameters of the subsurface, but also for appraisal of the thus obtained models. While not being neither the most general nor the most efficient methods, Bayesian inversion based on the calculation of the Jacobian of a given forward model can be used to evaluate many quantities useful in this process. The calculation of the Jacobian, however, is computationally expensive and, if done by divided differences, prone to truncation error. Here, automatic differentiation can be used to produce derivative code by source transformation of an existing forward model. We describe this process for a coupled fluid flow and heat transport finite difference code, which is used in a Bayesian inverse scheme to estimate thermal and hydraulic properties and boundary conditions form measured hydraulic potentials and temperatures. The resulting derivative code was validated by comparison to simple analytical solutions and divided differences. Synthetic examples from different flow regimes demonstrate the use of the inverse scheme, and its behaviour in different configurations.

A Software Upgrade of the NASA Aeroheating Code "MINIVER"

NASA Technical Reports Server (NTRS)

Louderback, Pierce Mathew

2013-01-01

Computational Fluid Dynamics (CFD) is a powerful and versatile tool simulating fluid and thermal environments of launch and re-entry vehicles alike. Where it excels in power and accuracy, however, it lacks in speed. An alternative tool for this purpose is known as MINIVER, an aeroheating code widely used by NASA and within the aerospace industry. Capable of providing swift, reasonably accurate approximations of the fluid and thermal environment of launch vehicles, MINIVER is used where time is of the essence and accuracy need not be exact. However, MINIVER is an old, aging tool: running on a user-unfriendly, legacy command-line interface, it is difficult for it to keep pace with more modem software tools. Florida Institute of Technology was tasked with the construction of a new Graphical User Interface (GUI) that implemented the legacy version's capabilities and enhanced them with new tools and utilities. This thesis provides background to the legacy version of the program, the progression and final version of a modem user interface, and benchmarks to demonstrate its usefulness.

Needs and opportunities for CFD-code validation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, B.L.

1996-06-01

The conceptual design for the ESS target consists of a horizontal cylinder containing a liquid metal - mercury is considered in the present study - which circulates by forced convection and carries away the waste heat generated by the spallation reactions. The protons enter the target via a beam window, which must withstand the thermal, mechanical and radiation loads to which it is subjected. For a beam power of 5MW, it is estimated that about 3.3MW of waste heat would be deposited in the target material and associated structures. it is intended to confirm, by detailed thermal-hydraulics calculations, that amore » convective flow of the liquid metal target material can effectively remove the waste heat. The present series of Computational Fluid Dynamics (CFD) calculations has indicated that a single-inlet Target design leads to excessive local overheating, but a multiple-inlet design, is coolable. With this option, inlet flow streams, two from the sides and one from below, merge over the target window, cooling the window itself in crossflow and carrying away the heat generated volumetrically in the mercury with a strong axial flow down the exit channel. The three intersecting streams form a complex, three-dimensional, swirling flow field in which critical heat transfer processes are taking place. In order to produce trustworthy code simulations, it is necessary that the mesh resolution is adequate for the thermal-hydraulic conditions encountered and that the physical models used by the code are appropriate to the fluid dynamic environment. The former relies on considerable user experience in the application of the code, and the latter assurance is best gained in the context of controlled benchmark activities where measured data are available. Such activities will serve to quantify the accuracy of given models and to identify potential problem area for the numerical simulation which may not be obvious from global heat and mass balance considerations.« less

10. Floor Layout of Thermal Hydraulics Laboratory, from The Thermal ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

10. Floor Layout of Thermal Hydraulics Laboratory, from The Thermal Hydraulics Laboratory at Hanford. General Electric Company, Hanford Atomic Products Operation, Richland, Washington, 1961. - D-Reactor Complex, Deaeration Plant-Refrigeration Buildings, Area 100-D, Richland, Benton County, WA

Characterization of thermal-hydraulic and ignition phenomena in prototypic, full-length boiling water reactor spent fuel pool assemblies after a complete loss-of-coolant accident.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lindgren, Eric Richard; Durbin, Samuel G

2007-04-01

The objective of this project was to provide basic thermal-hydraulic data associated with a SFP complete loss-of-coolant accident. The accident conditions of interest for the SFP were simulated in a full-scale prototypic fashion (electrically-heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate code validation (primarily MELCOR) and reduce questions associated with interpretation of the experimental results. It was necessary to simulate a cluster of assemblies to represent a higher decay (younger) assembly surrounded by older, lower-power assemblies. Specifically, this program providedmore » data and analysis confirming: (1) MELCOR modeling of inter-assembly radiant heat transfer, (2) flow resistance modeling and the natural convective flow induced in a fuel assembly as it heats up in air, (3) the potential for and nature of thermal transient (i.e., Zircaloy fire) propagation, and (4) mitigation strategies concerning fuel assembly management.« less

Steady-State Thermal-Hydraulics Analyses for the Conversion of BR2 to Low Enriched Uranium Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Bergeron, A.; Dionne, B.

The code PLTEMP/ANL version 4.2 was used to perform the steady-state thermal-hydraulic analyses of the BR2 research reactor for conversion from Highly-Enriched to Low Enriched Uranium fuel (HEU and LEU, respectively). Calculations were performed to evaluate different fuel assemblies with respect to the onset of nucleate boiling (ONB), flow instability (FI), critical heat flux (CHF) and fuel temperature at beginning of cycle conditions. The fuel assemblies were characteristic of fresh fuel (0% burnup), highest heat flux (16% burnup), highest power (32% burnup) and highest burnup (46% burnup). Results show that the high heat flux fuel element is limiting for ONB,more » FI, and CHF, for both HEU and LEU fuel, but that the high power fuel element produces similar margin in a few cases. The maximum fuel temperature similarly occurs in both the high heat flux and high power fuel assemblies for both HEU and LEU fuel. A sensitivity study was also performed to evaluate the variation in fuel temperature due to uncertainties in the thermal conductivity degradation associated with burnup.« less

A HISTORICAL PERSPECTIVE OF NUCLEAR THERMAL HYDRAULICS

DOE Office of Scientific and Technical Information (OSTI.GOV)

D’Auria, F; Rohatgi, Upendra S.

The nuclear thermal-hydraulics discipline was developed following the needs for nuclear power plants (NPPs) and, to a more limited extent, research reactors (RR) design and safety. As in all other fields where analytical methods are involved, nuclear thermal-hydraulics took benefit of the development of computers. Thermodynamics, rather than fluid dynamics, is at the basis of the development of nuclear thermal-hydraulics together with the experiments in complex two-phase situations, namely, geometry, high thermal density, and pressure.

Thermal-hydraulic behaviors of vapor-liquid interface due to arrival of a pressure wave

DOE Office of Scientific and Technical Information (OSTI.GOV)

Inoue, Akira; Fujii, Yoshifumi; Matsuzaki, Mitsuo

In the vapor explosion, a pressure wave (shock wave) plays a fundamental role for triggering, propagation and enhancement of the explosion. Energy of the explosion is related to the magnitude of heat transfer rate from hot liquid to cold volatile one. This is related to an increasing rate of interface area and to an amount of transient heat flux between the liquids. In this study, the characteristics of transient heat transfer and behaviors of vapor film both on the platinum tube and on the hot melt tin drop, under same boundary conditions have been investigated. It is considered that theremore » exists a fundamental mechanism of the explosion in the initial expansion process of the hot liquid drop immediately after arrival of pressure wave. The growth rate of the vapor film is much faster on the hot liquid than that on the solid surface. Two kinds of roughness were observed, one due to the Taylor instability, by rapid growth of the explosion bubble, and another, nucleation sites were observed at the vapor-liquid interface. Based on detailed observation of early stage interface behaviors after arrival of a pressure wave, the thermal fragmentation mechanism is proposed.« less

Parameter Measurement Methods for Interfacing Hydraulic Systems with Microelectronic Instruments and Controllers.

DTIC Science & Technology

1983-11-01

successfully. I- Accession For NTIS -GO iiiONa DTIC TAB t Unannounced - Justificatio Distribution/ I Availability Codes vail and/or DIst Special IA-11...terms of initial signal power. An active sensor must be excited externally. Such a sensor receives its power from an external source and merely modulates...electrons in the material to gain L enough energy to be emitted. The voltage source causes a positive potential to be felt on the collector, thus causing the

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uchibori, Akihiro; Kurihara, Akikazu; Ohshima, Hiroyuki

A multiphysics analysis system for sodium-water reaction phenomena in a steam generator of sodium-cooled fast reactors was newly developed. The analysis system consists of the mechanistic numerical analysis codes, SERAPHIM, TACT, and RELAP5. The SERAPHIM code calculates the multicomponent multiphase flow and sodium-water chemical reaction caused by discharging of pressurized water vapor. Applicability of the SERAPHIM code was confirmed through the analyses of the experiment on water vapor discharging in liquid sodium. The TACT code was developed to calculate heat transfer from the reacting jet to the adjacent tube and to predict the tube failure occurrence. The numerical models integratedmore » into the TACT code were verified through some related experiments. The RELAP5 code evaluates thermal hydraulic behavior of water inside the tube. The original heat transfer correlations were corrected for the tube rapidly heated by the reacting jet. The developed system enables evaluation of the wastage environment and the possibility of the failure propagation.« less

Test case specifications for coupled neutronics-thermal hydraulics calculation of Gas-cooled Fast Reactor

NASA Astrophysics Data System (ADS)

Osuský, F.; Bahdanovich, R.; Farkas, G.; Haščík, J.; Tikhomirov, G. V.

2017-01-01

The paper is focused on development of the coupled neutronics-thermal hydraulics model for the Gas-cooled Fast Reactor. It is necessary to carefully investigate coupled calculations of new concepts to avoid recriticality scenarios, as it is not possible to ensure sub-critical state for a fast reactor core under core disruptive accident conditions. Above mentioned calculations are also very suitable for development of new passive or inherent safety systems that can mitigate the occurrence of the recriticality scenarios. In the paper, the most promising fuel material compositions together with a geometry model are described for the Gas-cooled fast reactor. Seven fuel pin and fuel assembly geometry is proposed as a test case for coupled calculation with three different enrichments of fissile material in the form of Pu-UC. The reflective boundary condition is used in radial directions of the test case and vacuum boundary condition is used in axial directions. During these condition, the nuclear system is in super-critical state and to achieve a stable state (which is numerical representation of operational conditions) it is necessary to decrease the reactivity of the system. The iteration scheme is proposed, where SCALE code system is used for collapsing of a macroscopic cross-section into few group representation as input for coupled code NESTLE.

Three-dimensional fuel pin model validation by prediction of hydrogen distribution in cladding and comparison with experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Aly, A.; Avramova, Maria; Ivanov, Kostadin

To correctly describe and predict this hydrogen distribution there is a need for multi-physics coupling to provide accurate three-dimensional azimuthal, radial, and axial temperature distributions in the cladding. Coupled high-fidelity reactor-physics codes with a sub-channel code as well as with a computational fluid dynamics (CFD) tool have been used to calculate detailed temperature distributions. These high-fidelity coupled neutronics/thermal-hydraulics code systems are coupled further with the fuel-performance BISON code with a kernel (module) for hydrogen. Both hydrogen migration and precipitation/dissolution are included in the model. Results from this multi-physics analysis is validated utilizing calculations of hydrogen distribution using models informed bymore » data from hydrogen experiments and PIE data.« less

A new method for mapping variability in vertical seepage flux in streambeds

NASA Astrophysics Data System (ADS)

Chen, Xunhong; Song, Jinxi; Cheng, Cheng; Wang, Deming; Lackey, Susan O.

2009-05-01

A two-step approach was used to measure the flux across the water-sediment interface in river channels. A hollow tube was pressed into the streambed and an in situ sediment column of the streambed was created inside the tube. The hydraulic gradient between the two ends of the sediment column was measured. The vertical hydraulic conductivity of the sediment column was determined using a falling-head permeameter test in the river. Given the availability of the hydraulic gradient and vertical hydraulic conductivity of the streambed, Darcy’s law was used to calculate the specific discharge. This approach was applied to the Elkhorn River and one tributary in northeastern Nebraska, USA. The results suggest that the magnitude of the vertical flux varied greatly within a short distance. Furthermore, the flux can change direction from downward to upward between two locations only several meters apart. This spatial pattern of variation probably represents the inflow and outflow within the hyporheic zone, not the regional ambient flow systems. In this study, a thermal infrared camera was also used to detect the discharge locations of groundwater in the streambed. After the hydraulic gradient and the vertical hydraulic conductivity were estimated from the groundwater spring, the discharge rate was calculated.

Optimization of coupled multiphysics methodology for safety analysis of pebble bed modular reactor

NASA Astrophysics Data System (ADS)

Mkhabela, Peter Tshepo

The research conducted within the framework of this PhD thesis is devoted to the high-fidelity multi-physics (based on neutronics/thermal-hydraulics coupling) analysis of Pebble Bed Modular Reactor (PBMR), which is a High Temperature Reactor (HTR). The Next Generation Nuclear Plant (NGNP) will be a HTR design. The core design and safety analysis methods are considerably less developed and mature for HTR analysis than those currently used for Light Water Reactors (LWRs). Compared to LWRs, the HTR transient analysis is more demanding since it requires proper treatment of both slower and much longer transients (of time scale in hours and days) and fast and short transients (of time scale in minutes and seconds). There is limited operation and experimental data available for HTRs for validation of coupled multi-physics methodologies. This PhD work developed and verified reliable high fidelity coupled multi-physics models subsequently implemented in robust, efficient, and accurate computational tools to analyse the neutronics and thermal-hydraulic behaviour for design optimization and safety evaluation of PBMR concept The study provided a contribution to a greater accuracy of neutronics calculations by including the feedback from thermal hydraulics driven temperature calculation and various multi-physics effects that can influence it. Consideration of the feedback due to the influence of leakage was taken into account by development and implementation of improved buckling feedback models. Modifications were made in the calculation procedure to ensure that the xenon depletion models were accurate for proper interpolation from cross section tables. To achieve this, the NEM/THERMIX coupled code system was developed to create the system that is efficient and stable over the duration of transient calculations that last over several tens of hours. Another achievement of the PhD thesis was development and demonstration of full-physics, three-dimensional safety analysis methodology for the PBMR to provide reference solutions. Investigation of different aspects of the coupled methodology and development of efficient kinetics treatment for the PBMR were carried out, which accounts for all feedback phenomena in an efficient manner. The OECD/NEA PBMR-400 coupled code benchmark was used as a test matrix for the proposed investigations. The integrated thermal-hydraulics and neutronics (multi-physics) methods were extended to enable modeling of a wider range of transients pertinent to the PBMR. First, the effect of the spatial mapping schemes (spatial coupling) was studied and quantified for different types of transients, which resulted in implementation of improved mapping methodology based on user defined criteria. The second aspect that was studied and optimized is the temporal coupling and meshing schemes between the neutronics and thermal-hydraulics time step selection algorithms. The coupled code convergence was achieved supplemented by application of methods to accelerate it. Finally, the modeling of all feedback phenomena in PBMRs was investigated and a novel treatment of cross-section dependencies was introduced for improving the representation of cross-section variations. The added benefit was that in the process of studying and improving the coupled multi-physics methodology more insight was gained into the physics and dynamics of PBMR, which will help also to optimize the PBMR design and improve its safety. One unique contribution of the PhD research is the investigation of the importance of the correct representation of the three-dimensional (3-D) effects in the PBMR analysis. The performed studies demonstrated that explicit 3-D modeling of control rod movement is superior and removes the errors associated with the grey curtain (2-D homogenized) approximation.

Developmental assessment of the Fort St. Vrain version of the Composite HTGR Analysis Program (CHAP-2)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stroh, K.R.

1980-01-01

The Composite HTGR Analysis Program (CHAP) consists of a model-independent systems analysis mainframe named LASAN and model-dependent linked code modules, each representing a component, subsystem, or phenomenon of an HTGR plant. The Fort St. Vrain (FSV) version (CHAP-2) includes 21 coded modules that model the neutron kinetics and thermal response of the core; the thermal-hydraulics of the reactor primary coolant system, secondary steam supply system, and balance-of-plant; the actions of the control system and plant protection system; the response of the reactor building; and the relative hazard resulting from fuel particle failure. FSV steady-state and transient plant data are beingmore » used to partially verify the component modeling and dynamic smulation techniques used to predict plant response to postulated accident sequences.« less

Scale effect of slip boundary condition at solid–liquid interface

PubMed Central

Nagayama, Gyoko; Matsumoto, Takenori; Fukushima, Kohei; Tsuruta, Takaharu

2017-01-01

Rapid advances in microelectromechanical systems have stimulated the development of compact devices, which require effective cooling technologies (e.g., microchannel cooling). However, the inconsistencies between experimental and classical theoretical predictions for the liquid flow in microchannel remain unclarified. Given the larger surface/volume ratio of microchannel, the surface effects increase as channel scale decreases. Here we show the scale effect of the boundary condition at the solid–liquid interface on single-phase convective heat transfer characteristics in microchannels. We demonstrate that the deviation from classical theory with a reduction in hydraulic diameters is due to the breakdown of the continuum solid–liquid boundary condition. The forced convective heat transfer characteristics of single-phase laminar flow in a parallel-plate microchannel are investigated. Using the theoretical Poiseuille and Nusselt numbers derived under the slip boundary condition at the solid–liquid interface, we estimate the slip length and thermal slip length at the interface. PMID:28256536

Pretest aerosol code comparisons for LWR aerosol containment tests LA1 and LA2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wright, A.L.; Wilson, J.H.; Arwood, P.C.

The Light-Water-Reactor (LWR) Aerosol Containment Experiments (LACE) are being performed in Richland, Washington, at the Hanford Engineering Development Laboratory (HEDL) under the leadership of an international project board and the Electric Power Research Institute. These tests have two objectives: (1) to investigate, at large scale, the inherent aerosol retention behavior in LWR containments under simulated severe accident conditions, and (2) to provide an experimental data base for validating aerosol behavior and thermal-hydraulic computer codes. Aerosol computer-code comparison activities are being coordinated at the Oak Ridge National Laboratory. For each of the six LACE tests, ''pretest'' calculations (for code-to-code comparisons) andmore » ''posttest'' calculations (for code-to-test data comparisons) are being performed. The overall goals of the comparison effort are (1) to provide code users with experience in applying their codes to LWR accident-sequence conditions and (2) to evaluate and improve the code models.« less

Deepak Condenser Model (DeCoM)

NASA Technical Reports Server (NTRS)

Patel, Deepak

2013-01-01

Development of the DeCoM comes from the requirement of analyzing the performance of a condenser. A component of a loop heat pipe (LHP), the condenser, is interfaced with the radiator in order to reject heat. DeCoM simulates the condenser, with certain input parameters. Systems Improved Numerical Differencing Analyzer (SINDA), a thermal analysis software, calculates the adjoining component temperatures, based on the DeCoM parameters and interface temperatures to the radiator. Application of DeCoM is (at the time of this reporting) restricted to small-scale analysis, without the need for in-depth LHP component integrations. To efficiently develop a model to simulate the LHP condenser, DeCoM was developed to meet this purpose with least complexity. DeCoM is a single-condenser, single-pass simulator for analyzing its behavior. The analysis is done based on the interactions between condenser fluid, the wall, and the interface between the wall and the radiator. DeCoM is based on conservation of energy, two-phase equations, and flow equations. For two-phase, the Lockhart- Martinelli correlation has been used in order to calculate the convection value between fluid and wall. Software such as SINDA (for thermal analysis analysis) and Thermal Desktop (for modeling) are required. DeCoM also includes the ability to implement a condenser into a thermal model with the capability of understanding the code process and being edited to user-specific needs. DeCoM requires no license, and is an open-source code. Advantages to DeCoM include time dependency, reliability, and the ability for the user to view the code process and edit to their needs.

A Thermal Management Systems Model for the NASA GTX RBCC Concept

NASA Technical Reports Server (NTRS)

Traci, Richard M.; Farr, John L., Jr.; Laganelli, Tony; Walker, James (Technical Monitor)

2002-01-01

The Vehicle Integrated Thermal Management Analysis Code (VITMAC) was further developed to aid the analysis, design, and optimization of propellant and thermal management concepts for advanced propulsion systems. The computational tool is based on engineering level principles and models. A graphical user interface (GUI) provides a simple and straightforward method to assess and evaluate multiple concepts before undertaking more rigorous analysis of candidate systems. The tool incorporates the Chemical Equilibrium and Applications (CEA) program and the RJPA code to permit heat transfer analysis of both rocket and air breathing propulsion systems. Key parts of the code have been validated with experimental data. The tool was specifically tailored to analyze rocket-based combined-cycle (RBCC) propulsion systems being considered for space transportation applications. This report describes the computational tool and its development and verification for NASA GTX RBCC propulsion system applications.

Assessment of PWR Steam Generator modelling in RELAP5/MOD2. International Agreement Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Putney, J.M.; Preece, R.J.

1993-06-01

An assessment of Steam Generator (SG) modelling in the PWR thermal-hydraulic code RELAP5/MOD2 is presented. The assessment is based on a review of code assessment calculations performed in the UK and elsewhere, detailed calculations against a series of commissioning tests carried out on the Wolf Creek PWR and analytical investigations of the phenomena involved in normal and abnormal SG operation. A number of modelling deficiencies are identified and their implications for PWR safety analysis are discussed -- including methods for compensating for the deficiencies through changes to the input deck. Consideration is also given as to whether the deficiencies willmore » still be present in the successor code RELAP5/MOD3.« less

Assessment of the MHD capability in the ATHENA code using data from the ALEX (Argonne Liquid Metal Experiment) facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roth, P.A.

1988-10-28

The ATHENA (Advanced Thermal Hydraulic Energy Network Analyzer) code is a system transient analysis code with multi-loop, multi-fluid capabilities, which is available to the fusion community at the National Magnetic Fusion Energy Computing Center (NMFECC). The work reported here assesses the ATHENA magnetohydrodynamic (MHD) pressure drop model for liquid metals flowing through a strong magnetic field. An ATHENA model was developed for two simple geometry, adiabatic test sections used in the Argonne Liquid Metal Experiment (ALEX) at Argonne National Laboratory (ANL). The pressure drops calculated by ATHENA agreed well with the experimental results from the ALEX facility. 13 refs., 4more » figs., 2 tabs.« less

Comparing contribution of flexural and planar modes to thermodynamic properties

NASA Astrophysics Data System (ADS)

Mann, Sarita; Rani, Pooja; Jindal, V. K.

2017-05-01

Graphene, the most studied and explored 2D structure has unusual thermal properties such as negative thermal expansion, high thermal conductivity etc. We have already studied the thermal expansion behavior and various thermodynamic properties of pure graphene like heat capacity, entropy and free energy. The results of thermal expansion and various thermodynamic properties match well with available theoretical studies. For a deeper understanding of these properties, we analyzed the contribution of each phonon branch towards the total value of the individual property. To compute these properties, the dynamical matrix was calculated using VASP code where the density functional perturbation theory (DFPT) is employed under quasi-harmonic approximation in interface with phonopy code. It is noticed that transverse mode has major contribution to negative thermal expansion and all branches have almost same contribution towards the various thermodynamic properties with the contribution of ZA mode being the highest.

RELAP5 Model of the First Wall/Blanket Primary Heat Transfer System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Popov, Emilian L; Yoder Jr, Graydon L; Kim, Seokho H

2010-06-01

ITER inductive power operation is modeled and simulated using a system level computer code to evaluate the behavior of the Primary Heat Transfer System (PHTS) and predict parameter operational ranges. The control algorithm strategy and derivation are summarized in this report as well. A major feature of ITER is pulsed operation. The plasma does not burn continuously, but the power is pulsed with large periods of zero power between pulses. This feature requires active temperature control to maintain a constant blanket inlet temperature and requires accommodation of coolant thermal expansion during the pulse. In view of the transient nature ofmore » the power (plasma) operation state a transient system thermal-hydraulics code was selected: RELAP5. The code has a well-documented history for nuclear reactor transient analyses, it has been benchmarked against numerous experiments, and a large user database of commonly accepted modeling practices exists. The process of heat deposition and transfer in the blanket modules is multi-dimensional and cannot be accurately captured by a one-dimensional code such as RELAP5. To resolve this, a separate CFD calculation of blanket thermal power evolution was performed using the 3-D SC/Tetra thermofluid code. A 1D-3D co-simulation more realistically models FW/blanket internal time-dependent thermal inertia while eliminating uncertainties in the time constant assumed in a 1-D system code. Blanket water outlet temperature and heat release histories for any given ITER pulse operation scenario are calculated. These results provide the basis for developing time dependent power forcing functions which are used as input in the RELAP5 calculations.« less

Thermally Actuated Hydraulic Pumps

NASA Technical Reports Server (NTRS)

Jones, Jack; Ross, Ronald; Chao, Yi

2008-01-01

Thermally actuated hydraulic pumps have been proposed for diverse applications in which direct electrical or mechanical actuation is undesirable and the relative slowness of thermal actuation can be tolerated. The proposed pumps would not contain any sliding (wearing) parts in their compressors and, hence, could have long operational lifetimes. The basic principle of a pump according to the proposal is to utilize the thermal expansion and contraction of a wax or other phase-change material in contact with a hydraulic fluid in a rigid chamber. Heating the chamber and its contents from below to above the melting temperature of the phase-change material would cause the material to expand significantly, thus causing a substantial increase in hydraulic pressure and/or a substantial displacement of hydraulic fluid out of the chamber. Similarly, cooling the chamber and its contents from above to below the melting temperature of the phase-change material would cause the material to contract significantly, thus causing a substantial decrease in hydraulic pressure and/or a substantial displacement of hydraulic fluid into the chamber. The displacement of the hydraulic fluid could be used to drive a piston. The figure illustrates a simple example of a hydraulic jack driven by a thermally actuated hydraulic pump. The pump chamber would be a cylinder containing encapsulated wax pellets and containing radial fins to facilitate transfer of heat to and from the wax. The plastic encapsulation would serve as an oil/wax barrier and the remaining interior space could be filled with hydraulic oil. A filter would retain the encapsulated wax particles in the pump chamber while allowing the hydraulic oil to flow into and out of the chamber. In one important class of potential applications, thermally actuated hydraulic pumps, exploiting vertical ocean temperature gradients for heating and cooling as needed, would be used to vary hydraulic pressures to control buoyancy in undersea research vessels. Heretofore, electrically actuated hydraulic pumps have been used for this purpose. By eliminating the demand for electrical energy for pumping, the use of the thermally actuated hydraulic pumps could prolong the intervals between battery charges, thus making it possible to greatly increase the durations of undersea exploratory missions.

Modeling of two-phase flow instabilities during startup transients utilizing RAMONA-4B methodology

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paniagua, J.; Rohatgi, U.S.; Prasad, V.

1996-10-01

RAMONA-4B code is currently under development for simulating thermal hydraulic instabilities that can occur in Boiling Water Reactors (BWRs) and the Simplified Boiling Water Reactor (SBWR). As one of the missions of RAMONA-4B is to simulate SBWR startup transients, where geysering or condensation-induced instability may be encountered, the code needs to be assessed for this application. This paper outlines the results of the assessments of the current version of RAMONA-4B and the modifications necessary for simulating the geysering or condensation-induced instability. The test selected for assessment are the geysering tests performed by Prof Aritomi (1993).

Heating of a fully saturated darcian half-space: Pressure generation, fluid expulsion, and phase change

USGS Publications Warehouse

Delaney, P.

1984-01-01

Analytical solutions are developed for the pressurization, expansion, and flow of one- and two-phase liquids during heating of fully saturated and hydraulically open Darcian half-spaces subjected to a step rise in temperature at its surface. For silicate materials, advective transfer is commonly unimportant in the liquid region; this is not always the case in the vapor region. Volume change is commonly more important than heat of vaporization in determining the position of the liquid-vapor interface, assuring that the temperatures cannot be determined independently of pressures. Pressure increases reach a maximum near the leading edge of the thermal front and penetrate well into the isothermal region of the body. Mass flux is insensitive to the hydraulic properties of the half-space. ?? 1984.

The COSIMA experiments and their verification, a data base for the validation of two phase flow computer codes

NASA Astrophysics Data System (ADS)

Class, G.; Meyder, R.; Stratmanns, E.

1985-12-01

The large data base for validation and development of computer codes for two-phase flow, generated at the COSIMA facility, is reviewed. The aim of COSIMA is to simulate the hydraulic, thermal, and mechanical conditions in the subchannel and the cladding of fuel rods in pressurized water reactors during the blowout phase of a loss of coolant accident. In terms of fuel rod behavior, it is found that during blowout under realistic conditions only small strains are reached. For cladding rupture extremely high rod internal pressures are necessary. The behavior of fuel rod simulators and the effect of thermocouples attached to the cladding outer surface are clarified. Calculations performed with the codes RELAP and DRUFAN show satisfactory agreement with experiments. This can be improved by updating the phase separation models in the codes.

Code Development and Assessment for Reactor Outage Thermal-Hydraulic and Safety Analysis - Midloop Operation with Loss of Residual Heat Removal

DOE Office of Scientific and Technical Information (OSTI.GOV)

Liang, Thomas K.S.; Ko, F.-K

Although only a few percent of residual power remains during plant outages, the associated risk of core uncovery and corresponding fuel overheating has been identified to be relatively high, particularly under midloop operation (MLO) in pressurized water reactors. However, to analyze the system behavior during outages, the tools currently available, such as RELAP5, RETRAN, etc., cannot easily perform the task. Therefore, a medium-sized program aiming at reactor outage simulation and evaluation, such as MLO with the loss of residual heat removal (RHR), was developed. All important thermal-hydraulic processes involved during MLO with the loss of RHR will be properly simulatedmore » by the newly developed reactor outage simulation and evaluation (ROSE) code. Important processes during MLO with loss of RHR involve a pressurizer insurge caused by the hot-leg flooding, reflux condensation, liquid holdup inside the steam generator, loop-seal clearance, core-level depression, etc. Since the accuracy of the pressure distribution from the classical nodal momentum approach will be degraded when the system is stratified and under atmospheric pressure, the two-region approach with a modified two-fluid model will be the theoretical basis of the new program to analyze the nuclear steam supply system during plant outages. To verify the analytical model in the first step, posttest calculations against the closed integral midloop experiments with loss of RHR were performed. The excellent simulation capacity of the ROSE code against the Institute of Nuclear Energy Research Integral System Test Facility (IIST) test data is demonstrated.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G; Powers, Jeffrey J; Clarno, Kevin T

The Consortium for Advanced Simulation of Light Water Reactors (CASL) aims to provide high-fidelity, multiphysics simulations of light water reactors (LWRs) by coupling a variety of codes within the Virtual Environment for Reactor Analysis (VERA). One of the primary goals of CASL is to predict local cladding failure through pellet-clad interaction (PCI). This capability is currently being pursued through several different approaches, such as with Tiamat, which is a simulation tool within VERA that more tightly couples the MPACT neutron transport solver, the CTF thermal hydraulics solver, and the MOOSE-based Bison-CASL fuel performance code. However, the process in this papermore » focuses on running fuel performance calculations with Bison-CASL to predict PCI using the multicycle output data from coupled neutron transport/thermal hydraulics simulations. In recent work within CASL, Watts Bar Unit 1 has been simulated over 12 cycles using the VERA core simulator capability based on MPACT and CTF. Using the output from these simulations, Bison-CASL results can be obtained without rerunning all 12 cycles, while providing some insight into PCI indicators. Multi-cycle Bison-CASL results are presented and compared against results from the FRAPCON fuel performance code. There are several quantities of interest in considering PCI and subsequent fuel rod failures, such as the clad hoop stress and maximum centerline fuel temperature, particularly as a function of time. Bison-CASL performs single-rod simulations using representative power and temperature distributions, providing high-resolution results for these and a number of other quantities. This will assist in identifying fuels rods as potential failure locations for use in further analyses.« less

Hydraulic performance of compacted clay liners under simulated daily thermal cycles.

PubMed

Aldaeef, A A; Rayhani, M T

2015-10-01

Compacted clay liners (CCLs) are commonly used as hydraulic barriers in several landfill applications to isolate contaminants from the surrounding environment and minimize the escape of leachate from the landfill. Prior to waste placement in landfills, CCLs are often exposed to temperature fluctuations which can affect the hydraulic performance of the liner. Experimental research was carried out to evaluate the effects of daily thermal cycles on the hydraulic performance of CCLs under simulated landfill conditions. Hydraulic conductivity tests were conducted on different soil specimens after being exposed to various thermal and dehydration cycles. An increase in the CCL hydraulic conductivity of up to one order of magnitude was recorded after 30 thermal cycles for soils with low plasticity index (PI = 9.5%). However, medium (PI = 25%) and high (PI = 37.2%) plasticity soils did not show significant hydraulic deviation due to their self-healing potential. Overlaying the CCL with a cover layer minimized the effects of daily thermal cycles, and maintained stable hydraulic performance in the CCLs even after exposure to 60 thermal cycles. Wet-dry cycles had a significant impact on the hydraulic aspect of low plasticity CCLs. However, medium and high plasticity CCLs maintained constant hydraulic performance throughout the test intervals. The study underscores the importance of protecting the CCL from exposure to atmosphere through covering it by a layer of geomembrane or an interim soil layer. Copyright © 2015 Elsevier Ltd. All rights reserved.

A Novel Multi-Scale Domain Overlapping CFD/STH Coupling Methodology for Multi-Dimensional Flows Relevant to Nuclear Applications

NASA Astrophysics Data System (ADS)

Grunloh, Timothy P.

The objective of this dissertation is to develop a 3-D domain-overlapping coupling method that leverages the superior flow field resolution of the Computational Fluid Dynamics (CFD) code STAR-CCM+ and the fast execution of the System Thermal Hydraulic (STH) code TRACE to efficiently and accurately model thermal hydraulic transport properties in nuclear power plants under complex conditions of regulatory and economic importance. The primary contribution is the novel Stabilized Inertial Domain Overlapping (SIDO) coupling method, which allows for on-the-fly correction of TRACE solutions for local pressures and velocity profiles inside multi-dimensional regions based on the results of the CFD simulation. The method is found to outperform the more frequently-used domain decomposition coupling methods. An STH code such as TRACE is designed to simulate large, diverse component networks, requiring simplifications to the fluid flow equations for reasonable execution times. Empirical correlations are therefore required for many sub-grid processes. The coarse grids used by TRACE diminish sensitivity to small scale geometric details such as Reactor Pressure Vessel (RPV) internals. A CFD code such as STAR-CCM+ uses much finer computational meshes that are sensitive to the geometric details of reactor internals. In turbulent flows, it is infeasible to fully resolve the flow solution, but the correlations used to model turbulence are at a low level. The CFD code can therefore resolve smaller scale flow processes. The development of a 3-D coupling method was carried out with the intention of improving predictive capabilities of transport properties in the downcomer and lower plenum regions of an RPV in reactor safety calculations. These regions are responsible for the multi-dimensional mixing effects that determine the distribution at the core inlet of quantities with reactivity implications, such as fluid temperature and dissolved neutron absorber concentration.

Development of a new code to solve hydro-mechanical coupling, shear failure and tensile failure due to hydraulic fracturing operations.

NASA Astrophysics Data System (ADS)

María Gómez Castro, Berta; De Simone, Silvia; Carrera, Jesús

2016-04-01

Nowadays, there are still some unsolved relevant questions which must be faced if we want to proceed to the hydraulic fracturing in a safe way. How much will the fracture propagate? This is one of the most important questions that have to be solved in order to avoid the formation of pathways leading to aquifer targets and atmospheric release. Will the fracture failure provoke a microseismic event? Probably this is the biggest fear that people have in fracking. The aim of this work (developed as a part of the EU - FracRisk project) is to understand the hydro-mechanical coupling that controls the shear of existing fractures and their propagation during a hydraulic fracturing operation, in order to identify the key parameters that dominate these processes and answer the mentioned questions. This investigation focuses on the development of a new C++ code which simulates hydro-mechanical coupling, shear movement and propagation of a fracture. The framework employed, called Kratos, uses the Finite Element Method and the fractures are represented with an interface element which is zero thickness. This means that both sides of the element lie together in the initial configuration (it seems a 1D element in a 2D domain, and a 2D element in a 3D domain) and separate as the adjacent matrix elements deform. Since we are working in hard, fragile rocks, we can assume an elastic matrix and impose irreversible displacements in fractures when rock failure occurs. The formulation used to simulate shear and tensile failures is based on the analytical solution proposed by Okada, 1992 and it is part of an iterative process. In conclusion, the objective of this work is to employ the new code developed to analyze the main uncertainties related with the hydro-mechanical behavior of fractures derived from the hydraulic fracturing operations.

Multidimensional computer simulation of Stirling cycle engines

NASA Technical Reports Server (NTRS)

Hall, C. A.; Porsching, T. A.; Medley, J.; Tew, R. C.

1990-01-01

The computer code ALGAE (algorithms for the gas equations) treats incompressible, thermally expandable, or locally compressible flows in complicated two-dimensional flow regions. The solution method, finite differencing schemes, and basic modeling of the field equations in ALGAE are applicable to engineering design settings of the type found in Stirling cycle engines. The use of ALGAE to model multiple components of the space power research engine (SPRE) is reported. Videotape computer simulations of the transient behavior of the working gas (helium) in the heater-regenerator-cooler complex of the SPRE demonstrate the usefulness of such a program in providing information on thermal and hydraulic phenomena in multiple component sections of the SPRE.

Thermal and flow analysis of the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% effort of Title 1)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Steinke, R.G.; Mueller, C.; Knight, T.D.

1998-03-01

The computational fluid dynamics code CFX4.2 was used to evaluate steady-state thermal-hydraulic conditions in the Fluor Daniel, Inc., Nuclear Material Storage Facility renovation design (initial 30% of Title 1). Thirteen facility cases were evaluated with varying temperature dependence, drywell-array heat-source magnitude and distribution, location of the inlet tower, and no-flow curtains in the drywell-array vault. Four cases of a detailed model of the inlet-tower top fixture were evaluated to show the effect of the canopy-cruciform fixture design on the air pressure and flow distributions.

Efficient Geometry and Data Handling for Large-Scale Monte Carlo - Thermal-Hydraulics Coupling

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard

2014-06-01

Detailed coupling of thermal-hydraulics calculations to Monte Carlo reactor criticality calculations requires each axial layer of each fuel pin to be defined separately in the input to the Monte Carlo code in order to assign to each volume the temperature according to the result of the TH calculation, and if the volume contains coolant, also the density of the coolant. This leads to huge input files for even small systems. In this paper a methodology for dynamical assignment of temperatures with respect to cross section data is demonstrated to overcome this problem. The method is implemented in MCNP5. The method is verified for an infinite lattice with 3x3 BWR-type fuel pins with fuel, cladding and moderator/coolant explicitly modeled. For each pin 60 axial zones are considered with different temperatures and coolant densities. The results of the axial power distribution per fuel pin are compared to a standard MCNP5 run in which all 9x60 cells for fuel, cladding and coolant are explicitly defined and their respective temperatures determined from the TH calculation. Full agreement is obtained. For large-scale application the method is demonstrated for an infinite lattice with 17x17 PWR-type fuel assemblies with 25 rods replaced by guide tubes. Again all geometrical detailed is retained. The method was used in a procedure for coupled Monte Carlo and thermal-hydraulics iterations. Using an optimised iteration technique, convergence was obtained in 11 iteration steps.

A thermal NO(x) prediction model - Scalar computation module for CFD codes with fluid and kinetic effects

NASA Technical Reports Server (NTRS)

Mcbeath, Giorgio; Ghorashi, Bahman; Chun, Kue

1993-01-01

A thermal NO(x) prediction model is developed to interface with a CFD, k-epsilon based code. A converged solution from the CFD code is the input to the postprocessing model for prediction of thermal NO(x). The model uses a decoupled analysis to estimate the equilibrium level of (NO(x))e which is the constant rate limit. This value is used to estimate the flame (NO(x)) and in turn predict the rate of formation at each node using a two-step Zeldovich mechanism. The rate is fixed on the NO(x) production rate plot by estimating the time to reach equilibrium by a differential analysis based on the reaction: O + N2 = NO + N. The rate is integrated in the nonequilibrium time space based on the residence time at each node in the computational domain. The sum of all nodal predictions yields the total NO(x) level.

77 FR 9707 - Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Thermal-Hydraulics...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-17

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards Meeting of the ACRS Subcommittee on Thermal-Hydraulics Phenomena; Revision to February 22, 2012, ACRS Meeting Federal Register Notice The Federal Register Notice for the ACRS Subcommittee meeting on Thermal-Hydraulics Phenomena...

75 FR 69140 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-11-10

... To Support Specific Success Criteria in the Standardized Plant Analysis Risk Models--Surry and Peach... INFORMATION: NUREG-1953, ``Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the... document entitled: NUREG-1953, ``Confirmatory Thermal- Hydraulic Analysis to Support Specific Success...

Multiscale Modeling and Multifunctional Composites

DTIC Science & Technology

2013-07-17

dλ α µ α= − − = +E Eθ θ (9) 6 where α is the coefficient of thermal expansion , and ,e d...longitudinal and transverse coefficient of thermal expansion , respectively. The piezoelectric constants are related by (Bahei-El-Din, 2009) 31 31 33 33 31...is coded into the user defined subroutine UEXPAN of the ABAQUS finite element program. This serves as the interface between the global finite element

The STAT7 Code for Statistical Propagation of Uncertainties In Steady-State Thermal Hydraulics Analysis of Plate-Fueled Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dunn, Floyd E.; Hu, Lin-wen; Wilson, Erik

The STAT code was written to automate many of the steady-state thermal hydraulic safety calculations for the MIT research reactor, both for conversion of the reactor from high enrichment uranium fuel to low enrichment uranium fuel and for future fuel re-loads after the conversion. A Monte-Carlo statistical propagation approach is used to treat uncertainties in important parameters in the analysis. These safety calculations are ultimately intended to protect against high fuel plate temperatures due to critical heat flux or departure from nucleate boiling or onset of flow instability; but additional margin is obtained by basing the limiting safety settings onmore » avoiding onset of nucleate boiling. STAT7 can simultaneously analyze all of the axial nodes of all of the fuel plates and all of the coolant channels for one stripe of a fuel element. The stripes run the length of the fuel, from the bottom to the top. Power splits are calculated for each axial node of each plate to determine how much of the power goes out each face of the plate. By running STAT7 multiple times, full core analysis has been performed by analyzing the margin to ONB for each axial node of each stripe of each plate of each element in the core.« less

Comparison of computational results of the SABRE LMFBR pin bundle blockage code with data from well-instrumented out-of-pile test bundles (THORS bundles 3A and 5A)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J.F.

The Subchannel Analysis of Blockages in Reactor Elements (SABRE) computer code, developed by the United Kingdom Atomic Energy Authority, is currently the only practical tool available for performing detailed analyses of velocity and temperature fields in the recirculating flow regions downstream of blockages in liquid-metal fast breeder reactor (LMFBR) pin bundles. SABRE is a subchannel analysis code; that is, it accurately represents the complex geometry of nuclear fuel pins arranged on a triangular lattice. The results of SABRE computational models are compared here with temperature data from two out-of-pile 19-pin test bundles from the Thermal-Hydraulic Out-of-Reactor Safety (THORS) Facility atmore » Oak Ridge National Laboratory. One of these bundles has a small central flow blockage (bundle 3A), while the other has a large edge blockage (bundle 5A). Values that give best agreement with experiment for the empirical thermal mixing correlation factor, FMIX, in SABRE are suggested. These values of FMIX are Reynolds-number dependent, however, indicating that the coded turbulent mixing correlation is not appropriate for wire-wrap pin bundles.« less

Coupling procedure for TRANSURANUS and KTF codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jimenez, J.; Alglave, S.; Avramova, M.

2012-07-01

The nuclear industry aims to ensure safe and economic operation of each single fuel rod introduced in the reactor core. This goal is even more challenging nowadays due to the current strategy of going for higher burn-up (fuel cycles of 18 or 24 months) and longer residence time. In order to achieve that goal, fuel modeling is the key to predict the fuel rod behavior and lifetime under thermal and pressure loads, corrosion and irradiation. In this context, fuel performance codes, such as TRANSURANUS, are used to improve the fuel rod design. The modeling capabilities of the above mentioned toolsmore » can be significantly improved if they are coupled with a thermal-hydraulic code in order to have a better description of the flow conditions within the rod bundle. For LWR applications, a good representation of the two phase flow within the fuel assembly is necessary in order to have a best estimate calculation of the heat transfer inside the bundle. In this paper we present the coupling methodology of TRANSURANUS with KTF (Karlsruhe Two phase Flow subchannel code) as well as selected results of the coupling proof of principle. (authors)« less

THE RETC CODE FOR QUANTIFYING THE HYDRAULIC FUNCTIONS OF UNSATURATED SOILS

EPA Science Inventory

This report describes the RETC computer code for analyzing the soil water retention and hydraulic conductivity functions of unsaturated soils. These hydraulic properties are key parameters in any quantitative description of water flow into and through the unsaturated zone of soil...

Comparison of COBRA III-C and SABRE-1 (wire-wrap version) computational results with steady-state data from a 19-pin internally guard heated sodium-cooled bundle with a six-channel central blockage (THORS Bundle 3C). [LMFBR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dearing, J F; Nelson, W R; Rose, S D

Computational thermal-hydraulic models of a 19-pin, electrically heated, wire-wrap liquid-metal fast breeder reactor test bundle were developed using two well-known subchannel analysis codes, COBRA III-C and SABRE-1 (wire-wrap version). These two codes use similar subchannel control volumes for the finite difference conservation equations but vary markedly in solution strategy and modeling capability. In particular, the empirical wire-wrap-forced diversion crossflow models are different. Surprisingly, however, crossflow velocity predictions of the two codes are very similar. Both codes show generally good agreement with experimental temperature data from a test in which a large radial temperature gradient was imposed. Differences between data andmore » code results are probably caused by experimental pin bowing, which is presently the limiting factor in validating coded empirical models.« less

Hydraulic Experiments for Determination of In-situ Hydraulic Conductivity of Submerged Sediments

PubMed Central

Lee, Bong-Joo; Lee, Ji-Hoon; Yoon, Heesung; Lee, Eunhee

2015-01-01

A new type of in-situ hydraulic permeameter was developed to determine vertical hydraulic conductivity (VHC) of saturated sediments from hydraulic experiments using Darcy's law. The system allows water to move upward through the porous media filled in the permeameter chamber driven into sediments at water-sediment interface. Darcy flux and hydraulic gradient can be measured using the system, and the VHC can be determined from the relationship between them using Darcy's law. Evaluations in laboratory and in field conditions were performed to see if the proposed permeameter give reliable and valid measures of the VHC even where the vertical flow at water-sediment interface and fluctuation of water stage exist without reducing the accuracy of the derived VHC. Results from the evaluation tests indicate that the permeameter proposed in this study can be used to measure VHC of saturated sandy sediments at water-sediment interface in stream and marine environment with high accuracy. PMID:25604984

Performance Evaluation of Nose Cap to Silica Tile Joint of RLV-TD under the Simulated Flight Environment using Plasma Wind Tunnel Facility

NASA Astrophysics Data System (ADS)

Pillai, Aravindakshan; Krishnaraj, K.; Sreenivas, N.; Nair, Praveen

2017-12-01

Indian Space Research Organisation, India has successfully flight tested the reusable launch vehicle through launching of a demonstration flight known as RLV-TD HEX mission. This mission has given a platform for exposing the thermal protection system to the real hypersonic flight thermal conditions and thereby validated the design. In this vehicle, the nose cap region is thermally protected by carbon-carbon followed by silica tiles with a gap in between them for thermal expansion. The gap is filled with silica fibre. Base material on which the C-C is placed is made of molybdenum. Silica tile with strain isolation pad is bonded to aluminium structure. These interfaces with a variety of materials are characterised with different coefficients of thermal expansion joined together. In order to evaluate and qualify this joint, model tests were carried out in Plasma Wind Tunnel facility under the simultaneous simulation of heat flux and shear levels as expected in flight. The thermal and flow parameters around the model are determined and made available for the thermal analysis using in-house CFD code. Two tests were carried out. The measured temperatures at different locations were benign in both these tests and the SiC coating on C-C and the interface were also intact. These tests essentially qualified the joint interface between C-C and molybdenum bracket and C-C to silica tile interface of RLV-TD.

Project Summary. THE RETC CODE FOR QUANTIFYING THE HYDRAULIC FUNCTIONS OF UNSATURATED SOILS

EPA Science Inventory

This summary describes the RETC computer code for analyzing the soil water retention and hydraulic conductivity functions of unsaturated soils. These hydraulic properties are key parameters in any quantitative description of water flow into and through the unsaturated zone of soi...

TRAC-PF1 code verification with data from the OTIS test facility. [Once-Through Intergral System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Childerson, M.T.; Fujita, R.K.

1985-01-01

A computer code (TRAC-PF1/MOD1) developed for predicting transient thermal and hydraulic integral nuclear steam supply system (NSSS) response was benchmarked. Post-small break loss-of-coolant accident (LOCA) data from a scaled, experimental facility, designated the One-Through Integral System (OTIS), were obtained for the Babcock and Wilcox NSSS and compared to TRAC predictions. The OTIS tests provided a challenging small break LOCA data set for TRAC verification. The major phases of a small break LOCA observed in the OTIS tests included pressurizer draining and loop saturation, intermittent reactor coolant system circulation, boiler-condenser mode, and the initial stages of refill. The TRAC code wasmore » successful in predicting OTIS loop conditions (system pressures and temperatures) after modification of the steam generator model. In particular, the code predicted both pool and auxiliary-feedwater initiated boiler-condenser mode heat transfer.« less

Process management using component thermal-hydraulic function classes

DOEpatents

Morman, James A.; Wei, Thomas Y. C.; Reifman, Jaques

1999-01-01

A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced.

Process management using component thermal-hydraulic function classes

DOEpatents

Morman, J.A.; Wei, T.Y.C.; Reifman, J.

1999-07-27

A process management expert system where following malfunctioning of a component, such as a pump, for determining system realignment procedures such as for by-passing the malfunctioning component with on-line speeds to maintain operation of the process at full or partial capacity or to provide safe shut down of the system while isolating the malfunctioning component. The expert system uses thermal-hydraulic function classes at the component level for analyzing unanticipated as well as anticipated component malfunctions to provide recommended sequences of operator actions. Each component is classified according to its thermal-hydraulic function, and the generic and component-specific characteristics for that function. Using the diagnosis of the malfunctioning component and its thermal hydraulic class, the expert system analysis is carried out using generic thermal-hydraulic first principles. One aspect of the invention employs a qualitative physics-based forward search directed primarily downstream from the malfunctioning component in combination with a subsequent backward search directed primarily upstream from the serviced component. Generic classes of components are defined in the knowledge base according to the three thermal-hydraulic functions of mass, momentum and energy transfer and are used to determine possible realignment of component configurations in response to thermal-hydraulic function imbalance caused by the malfunctioning component. Each realignment to a new configuration produces the accompanying sequence of recommended operator actions. All possible new configurations are examined and a prioritized list of acceptable solutions is produced. 5 figs.

Eulerian and Lagrangian Plasma Jet Modeling for the Plasma Liner Experiment

NASA Astrophysics Data System (ADS)

Hatcher, Richard; Cassibry, Jason; Stanic, Milos; Loverich, John; Hakim, Ammar

2011-10-01

The Plasma Liner Experiment (PLX) aims to demonstrate the feasibility of using spherically-convergent plasma jets to from an imploding plasma liner. Our group has modified two hydrodynamic simulation codes to include radiative loss, tabular equations of state (EOS), and thermal transport. Nautilus, created by TechX Corporation, is a finite-difference Eulerian code which solves the MHD equations formulated as systems of hyperbolic conservation laws. The other is SPHC, a smoothed particle hydrodynamics code produced by Stellingwerf Consulting. Use of the Lagrangian fluid particle approach of SPH is motivated by the ability to accurately track jet interfaces, the plasma vacuum boundary, and mixing of various layers, but Eulerian codes have been in development for much longer and have better shock capturing. We validate these codes against experimental measurements of jet propagation, expansion, and merging of two jets. Precursor jets are observed to form at the jet interface. Conditions that govern evolution of two and more merging jets are explored.

Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Fast Salt Reactor (MFSR)

NASA Astrophysics Data System (ADS)

Laureau, A.; Rubiolo, P. R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

2014-06-01

Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated.

Evaporation from soils subjected to natural boundary conditions at the land-atmospheric interface

NASA Astrophysics Data System (ADS)

Smits, K.; Illngasekare, T.; Ngo, V.; Cihan, A.

2012-04-01

Bare soil evaporation is a key process for water exchange between the land and the atmosphere and an important component of the water balance in semiarid and arid regions. However, there is no agreement on the best methodology to determine evaporation under different boundary conditions at the land surface. This becomes critical in developing models that couples land to the atmosphere. Because it is difficult to measure evaporation from soil, with the exception of using lysimeters, numerous formulations have been proposed to establish a relationship between the rate of evaporation and soil moisture and/or soil temperature and thermal properties. Different formulations vary in how they partition available energy. A need exists to systematically compare existing methods to experimental data under highly controlled conditions not achievable in the field. The goal of this work is to perform controlled experiments under transient conditions of soil moisture, temperature and wind at the land/atmospheric interface to test different conceptual and mathematical formulations for the soil surface boundary conditions to develop appropriate numerical models to be used in simulations. In this study, to better understand the coupled water-vapor-heat flow processes in the shallow subsurface near the land surface, we modified a previously developed theory by Smits et al. [2011] that allows non-equilibrium liquid/gas phase change with gas phase vapor diffusion to better account for dry soil conditions. The model did not implement fitting parameters such as a vapor enhancement factor that is commonly introduced into the vapor diffusion coefficient as an arbitrary multiplication factor. In order to experimentally test the numerical formulations/code, we performed a two-dimensional physical model experiment under varying boundary conditions using test sand for which the hydraulic and thermal properties were well characterized. Precision data under well-controlled transient heat and wind boundary conditions was generated and results from numerical simulations were compared with experimental data. Results demonstrate that the boundary condition approaches varied in their ability to capture stage 1- and stage 2- evaporation. Results also demonstrated the importance of properly characterizing soil thermal properties and accounting for dry soil conditions. The contribution of film flow to hydraulic conductivity for the layer above the drying front is dominant compared to that of capillary flow, demonstrating the importance of including film flow in modeling efforts for dry soils, especially for fine grained soils. Comparisons of different formulations of the surface boundary condition validate the need for joint evaluation of heat and mass transfer for better modeling accuracy. This knowledge is applicable to many current hydrologic and environmental problems to include climate modeling and the simulation of contaminant transport and volatilization in the shallow subsurface. Smits, K. M., A. Cihan, T. Sakaki, and T. H. Illangasekare (2011). Evaporation from soils under thermal boundary conditions: Experimental and modeling investigation to compare equilibrium- and nonequilibrium-based approaches, Water Resour. Res., 47, W05540, doi:10.1029/2010WR009533.

A Method to Estimate the Hydraulic Conductivity of the Ground by TRT Analysis.

PubMed

Liuzzo Scorpo, Alberto; Nordell, Bo; Gehlin, Signhild

2017-01-01

The knowledge of hydraulic properties of aquifers is important in many engineering applications. Careful design of ground-coupled heat exchangers requires that the hydraulic characteristics and thermal properties of the aquifer must be well understood. Knowledge of groundwater flow rate and aquifer thermal properties is the basis for proper design of such plants. Different methods have been developed in order to estimate hydraulic conductivity by evaluating the transport of various tracers (chemical, heat etc.); thermal response testing (TRT) is a specific type of heat tracer that allows including the hydraulic properties in an effective thermal conductivity value. Starting from these considerations, an expeditious, graphical method was proposed to estimate the hydraulic conductivity of the aquifer, using TRT data and plausible assumption. Suggested method, which is not yet verified or proven to be reliable, should be encouraging further studies and development in this direction. © 2016, National Ground Water Association.

Parallelization of TWOPORFLOW, a Cartesian Grid based Two-phase Porous Media Code for Transient Thermo-hydraulic Simulations

NASA Astrophysics Data System (ADS)

Trost, Nico; Jiménez, Javier; Imke, Uwe; Sanchez, Victor

2014-06-01

TWOPORFLOW is a thermo-hydraulic code based on a porous media approach to simulate single- and two-phase flow including boiling. It is under development at the Institute for Neutron Physics and Reactor Technology (INR) at KIT. The code features a 3D transient solution of the mass, momentum and energy conservation equations for two inter-penetrating fluids with a semi-implicit continuous Eulerian type solver. The application domain of TWOPORFLOW includes the flow in standard porous media and in structured porous media such as micro-channels and cores of nuclear power plants. In the latter case, the fluid domain is coupled to a fuel rod model, describing the heat flow inside the solid structure. In this work, detailed profiling tools have been utilized to determine the optimization potential of TWOPORFLOW. As a result, bottle-necks were identified and reduced in the most feasible way, leading for instance to an optimization of the water-steam property computation. Furthermore, an OpenMP implementation addressing the routines in charge of inter-phase momentum-, energy- and mass-coupling delivered good performance together with a high scalability on shared memory architectures. In contrast to that, the approach for distributed memory systems was to solve sub-problems resulting by the decomposition of the initial Cartesian geometry. Thread communication for the sub-problem boundary updates was accomplished by the Message Passing Interface (MPI) standard.

Coal-rock interface detector

NASA Technical Reports Server (NTRS)

Rose, S. D.; Crouch, C. E.; Jones, E. W. (Inventor)

1979-01-01

A coal-rock interface detector is presented which employs a radioactive source and radiation sensor. The source and sensor are separately and independently suspended and positioned against a mine surface of hydraulic pistons, which are biased from an air cushioned source of pressurized hydraulic fluid.

Neutronic Calculation Analysis for CN HCCB TBM-Set

NASA Astrophysics Data System (ADS)

Cao, Qixiang; Zhao, Fengchao; Zhao, Zhou; Wu, Xinghua; Li, Zaixin; Wang, Xiaoyu; Feng, Kaiming

2015-07-01

Using the Monte Carlo transport code MCNP, neutronic calculation analysis for China helium cooled ceramic breeder test blanket module (CN HCCB TBM) and the associated shield block (together called TBM-set) has been carried out based on the latest design of HCCB TBM-set and C-lite model. Key nuclear responses of HCCB TBM-set, such as the neutron flux, tritium production rate, nuclear heating and radiation damage, have been obtained and discussed. These nuclear performance data can be used as the basic input data for other analyses of HCCB TBM-set, such as thermal-hydraulics, thermal-mechanics and safety analysis. supported by the Major State Basic Research Development Program of China (973 Program) (No. 2013GB108000)

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

NASA Astrophysics Data System (ADS)

Salim, Mohd Faiz; Roslan, Ridha; Ibrahim, Mohd Rizal Mamat @

2014-02-01

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.

A User’s Guide to the PLTEMP/ANL Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Olson, A. P.; Kalimullah, M.; Feldman, E. E.

2016-07-25

PLTEMP/ANL V4.2 is a program that obtains a steady-state flow and temperature solution for a nuclear reactor core, or for a single fuel assembly. It is based on an evolutionary sequence of codes originally used for plate temperatures, hence “PLTEMP”, developed at Argonne National Laboratory over several decades. Fueled and non-fueled regions are modeled. Each fuel assembly consists of one or more plates or tubes separated by coolant channels. The fuel plates may have one to five layers of different materials, each with heat generation. The width of a fuel plate may be divided into multiple longitudinal stripes, each withmore » its own axial power shape. The temperature solution is effectively 2-dimensional. It begins with a one-dimensional solution across all coolant channels and fuel plates or tubes within a given fuel assembly, at the entrance to the assembly. The temperature solution is repeated for each axial node along the length of the fuel assembly. The geometry may be either slab or radial, corresponding to fuel assemblies made of a series of flat (or slightly curved) plates, or of nested tubes. A variety of thermal-hydraulic correlations are available with which to determine safety margins such as onset-of-nucleate boiling ratio(ONBR), departure from nucleate boiling ratio (DNBR), and onset of flow instability ratio (OFIR). Coolant properties for either light or heavy water are obtained from FORTRAN functions rather than from tables. The code is intended for thermal-hydraulic analysis of research reactor performance in the sub-cooled boiling regime. Both turbulent and laminar flow regimes can be modeled. Options to calculate both forced flow and natural circulation are available. A general search capability is available (Appendix XII) to greatly reduce the reactor analyst’s time.« less

Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets.

PubMed

Abedi, Ebrahim; Ebrahimkhani, Marzieh; Davari, Amin; Mirvakili, Seyed Mohammad; Tabasi, Mohsen; Maragheh, Mohammad Ghannadi

2016-12-01

Efficient and safe production of molybdenum-99 ( 99 Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99 Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99 Mo production is accessible at significant activity values in TRR current core configuration. Copyright © 2016 Elsevier Ltd. All rights reserved.

Thermal Hydraulic Computational Fluid Dynamics Simulations and Experimental Investigation of Deformed Fuel Assemblies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mays, Brian; Jackson, R. Brian

2017-03-08

The project, Toward a Longer Life Core: Thermal Hydraulic CFD Simulations and Experimental Investigation of Deformed Fuel Assemblies, DOE Project code DE-NE0008321, was a verification and validation project for flow and heat transfer through wire wrapped simulated liquid metal fuel assemblies that included both experiments and computational fluid dynamics simulations of those experiments. This project was a two year collaboration between AREVA, TerraPower, Argonne National Laboratory and Texas A&M University. Experiments were performed by AREVA and Texas A&M University. Numerical simulations of these experiments were performed by TerraPower and Argonne National Lab. Project management was performed by AREVA Federal Services.more » The first of a kind project resulted in the production of both local point temperature measurements and local flow mixing experiment data paired with numerical simulation benchmarking of the experiments. The project experiments included the largest wire-wrapped pin assembly Mass Index of Refraction (MIR) experiment in the world, the first known wire-wrapped assembly experiment with deformed duct geometries and the largest numerical simulations ever produced for wire-wrapped bundles.« less

An information theoretic approach to use high-fidelity codes to calibrate low-fidelity codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lewis, Allison, E-mail: lewis.allison10@gmail.com; Smith, Ralph; Williams, Brian

For many simulation models, it can be prohibitively expensive or physically infeasible to obtain a complete set of experimental data to calibrate model parameters. In such cases, one can alternatively employ validated higher-fidelity codes to generate simulated data, which can be used to calibrate the lower-fidelity code. In this paper, we employ an information-theoretic framework to determine the reduction in parameter uncertainty that is obtained by evaluating the high-fidelity code at a specific set of design conditions. These conditions are chosen sequentially, based on the amount of information that they contribute to the low-fidelity model parameters. The goal is tomore » employ Bayesian experimental design techniques to minimize the number of high-fidelity code evaluations required to accurately calibrate the low-fidelity model. We illustrate the performance of this framework using heat and diffusion examples, a 1-D kinetic neutron diffusion equation, and a particle transport model, and include initial results from the integration of the high-fidelity thermal-hydraulics code Hydra-TH with a low-fidelity exponential model for the friction correlation factor.« less

A Comprehensive Validation Approach Using The RAVEN Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J

2015-06-01

The RAVEN computer code , developed at the Idaho National Laboratory, is a generic software framework to perform parametric and probabilistic analysis based on the response of complex system codes. RAVEN is a multi-purpose probabilistic and uncertainty quantification platform, capable to communicate with any system code. A natural extension of the RAVEN capabilities is the imple- mentation of an integrated validation methodology, involving several different metrics, that represent an evolution of the methods currently used in the field. The state-of-art vali- dation approaches use neither exploration of the input space through sampling strategies, nor a comprehensive variety of metrics neededmore » to interpret the code responses, with respect experimental data. The RAVEN code allows to address both these lacks. In the following sections, the employed methodology, and its application to the newer developed thermal-hydraulic code RELAP-7, is reported.The validation approach has been applied on an integral effect experiment, representing natu- ral circulation, based on the activities performed by EG&G Idaho. Four different experiment configurations have been considered and nodalized.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Volume 3, Sessions 12-16

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

This document, Volume 3, includes papers presented at the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics (NURETH-7) September 10--15, 1995 at Saratoga Springs, N.Y. The following subjects are discussed: Progress in analytical and experimental work on the fundamentals of nuclear thermal-hydraulics, the development of advanced mathematical and numerical methods, ad the application of advancements in the field in the development of novel reactor concepts. Also combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. Selected abstracts have been indexed separately for inclusion in the Energy Science and Technology Database.

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Verification of Modelica-Based Models with Analytical Solutions for Tritium Diffusion

DOE PAGES

Rader, Jordan D.; Greenwood, Michael Scott; Humrickhouse, Paul W.

2018-03-20

Here, tritium transport in metal and molten salt fluids combined with diffusion through high-temperature structural materials is an important phenomenon in both magnetic confinement fusion (MCF) and molten salt reactor (MSR) applications. For MCF, tritium is desirable to capture for fusion fuel. For MSRs, uncaptured tritium potentially can be released to the environment. In either application, quantifying the time- and space-dependent tritium concentration in the working fluid(s) and structural components is necessary.Whereas capability exists specifically for calculating tritium transport in such systems (e.g., using TMAP for fusion reactors), it is desirable to unify the calculation of tritium transport with othermore » system variables such as dynamic fluid and structure temperature combined with control systems such as those that might be found in a system code. Some capability for radioactive trace substance transport exists in thermal-hydraulic systems codes (e.g., RELAP5-3D); however, this capability is not coupled to species diffusion through solids. Combined calculations of tritium transport and thermal-hydraulic solution have been demonstrated with TRIDENT but only for a specific type of MSR.Researchers at Oak Ridge National Laboratory have developed a set of Modelica-based dynamic system modeling tools called TRANsient Simulation Framework Of Reconfigurable Models (TRANSFORM) that were used previously to model advanced fission reactors and associated systems. In this system, the augmented TRANSFORM library includes dynamically coupled fluid and solid trace substance transport and diffusion. Results from simulations are compared against analytical solutions for verification.« less

Analytical methods in the high conversion reactor core design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zeggel, W.; Oldekop, W.; Axmann, J.K.

High conversion reactor (HCR) design methods have been used at the Technical University of Braunschweig (TUBS) with the technological support of Kraftwerk Union (KWU). The present state and objectives of this cooperation between KWU and TUBS in the field of HCRs have been described using existing design models and current activities aimed at further development and validation of the codes. The hard physical and thermal-hydraulic boundary conditions of pressurized water reactor (PWR) cores with a high degree of fuel utilization result from the tight packing of the HCR fuel rods and the high fissionable plutonium content of the fuel. Inmore » terms of design, the problem will be solved with rod bundles whose fuel rods are adjusted by helical spacers to the proposed small rod pitches. These HCR properties require novel computational models for neutron physics, thermal hydraulics, and fuel rod design. By means of a survey of the codes, the analytical procedure for present-day HCR core design is presented. The design programs are currently under intensive development, as design tools with a solid, scientific foundation and with essential parameters that are widely valid and are required for a promising optimization of the HCR core. Design results and a survey of future HCR development are given. In this connection, the reoptimization of the PWR core in the direction of an HCR is considered a fascinating scientific task, with respect to both economic and safety aspects.« less

Proceedings of the 7th International Meeting on Nuclear Reactor Thermal-Hydraulics NURETH-7. Sessions 17-24

DOE Office of Scientific and Technical Information (OSTI.GOV)

Block, R.C.; Feiner, F.

1995-09-01

Technical papers accepted for presentation at the Seventh International Topical Meeting on Nuclear Reactor Thermal-Hydraulics are included in the present Proceedings. Except for the invited papers in the plenary session, all other papers are contributed papers. The topics of the meeting encompass all major areas of nuclear thermal-hydraulics, including analytical and experimental works on the fundamental mechanisms of fluid flow and heat transfer, the development of advanced mathematical and numerical methods, and the application of advancements in the field in the development of novel reactor concepts. Because of the complex nature of nuclear reactors and power plants, several papers dealmore » with the combined issues of thermal-hydraulics and reactor/power-plant safety, core neutronics and/or radiation. The participation in the conference by the authors from several countries and four continents makes the Proceedings a comprehensive review of the recent progress in the field of nuclear reactor thermal-hydraulics worldwide. Individual papers have been cataloged separately.« less

TOPAZ2D heat transfer code users manual and thermal property data base

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shapiro, A.B.; Edwards, A.L.

1990-05-01

TOPAZ2D is a two dimensional implicit finite element computer code for heat transfer analysis. This user's manual provides information on the structure of a TOPAZ2D input file. Also included is a material thermal property data base. This manual is supplemented with The TOPAZ2D Theoretical Manual and the TOPAZ2D Verification Manual. TOPAZ2D has been implemented on the CRAY, SUN, and VAX computers. TOPAZ2D can be used to solve for the steady state or transient temperature field on two dimensional planar or axisymmetric geometries. Material properties may be temperature dependent and either isotropic or orthotropic. A variety of time and temperature dependentmore » boundary conditions can be specified including temperature, flux, convection, and radiation. Time or temperature dependent internal heat generation can be defined locally be element or globally by material. TOPAZ2D can solve problems of diffuse and specular band radiation in an enclosure coupled with conduction in material surrounding the enclosure. Additional features include thermally controlled reactive chemical mixtures, thermal contact resistance across an interface, bulk fluid flow, phase change, and energy balances. Thermal stresses can be calculated using the solid mechanics code NIKE2D which reads the temperature state data calculated by TOPAZ2D. A three dimensional version of the code, TOPAZ3D is available. The material thermal property data base, Chapter 4, included in this manual was originally published in 1969 by Art Edwards for use with his TRUMP finite difference heat transfer code. The format of the data has been altered to be compatible with TOPAZ2D. Bob Bailey is responsible for adding the high explosive thermal property data.« less

An MHD Code for the Study of Magnetic Structures in the Solar Wind

NASA Technical Reports Server (NTRS)

Allred, J. C.; MacNeice, P. J.

2015-01-01

We have developed a 2.5D MHD code designed to study how the solar wind influences the evolution of transient events in the solar corona and inner heliosphere. The code includes thermal conduction, coronal heating and radiative cooling. Thermal conduction is assumed to be magnetic field-aligned in the inner corona and transitions to a collisionless formulation in the outer corona. We have developed a stable method to handle field-aligned conduction around magnetic null points. The inner boundary is placed in the upper transition region, and the mass flux across the boundary is determined from 1D field-aligned characteristics and a 'radiative energy balance' condition. The 2.5D nature of this code makes it ideal for parameter studies not yet possible with 3D codes. We have made this code publicly available as a tool for the community. To this end we have developed a graphical interface to aid in the selection of appropriate options and a graphical interface that can process and visualize the data produced by the simulation. As an example, we show a simulation of a dipole field stretched into a helmet streamer by the solar wind. Plasmoids periodically erupt from the streamer, and we perform a parameter study of how the frequency and location of these eruptions changed in response to different levels of coronal heating. As a further example, we show the solar wind stretching a compact multi-polar flux system. This flux system will be used to study breakout coronal mass ejections in the presence of the solar wind.

Advanced Pellet-Cladding Interaction Modeling using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Montgomery, Robert O.; Capps, Nathan A.; Sunderland, Dion J.

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermo-mechanical-chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale code thatmore » is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

Fracture Analysis of Vessels. Oak Ridge FAVOR, v06.1, Computer Code: Theory and Implementation of Algorithms, Methods, and Correlations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williams, P. T.; Dickson, T. L.; Yin, S.

The current regulations to insure that nuclear reactor pressure vessels (RPVs) maintain their structural integrity when subjected to transients such as pressurized thermal shock (PTS) events were derived from computational models developed in the early-to-mid 1980s. Since that time, advancements and refinements in relevant technologies that impact RPV integrity assessment have led to an effort by the NRC to re-evaluate its PTS regulations. Updated computational methodologies have been developed through interactions between experts in the relevant disciplines of thermal hydraulics, probabilistic risk assessment, materials embrittlement, fracture mechanics, and inspection (flaw characterization). Contributors to the development of these methodologies include themore » NRC staff, their contractors, and representatives from the nuclear industry. These updated methodologies have been integrated into the Fracture Analysis of Vessels -- Oak Ridge (FAVOR, v06.1) computer code developed for the NRC by the Heavy Section Steel Technology (HSST) program at Oak Ridge National Laboratory (ORNL). The FAVOR, v04.1, code represents the baseline NRC-selected applications tool for re-assessing the current PTS regulations. This report is intended to document the technical bases for the assumptions, algorithms, methods, and correlations employed in the development of the FAVOR, v06.1, code.« less

Kinetic Monte Carlo simulation of dopant-defect systems under submicrosecond laser thermal processes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fisicaro, G.; Pelaz, Lourdes; Lopez, P.

2012-11-06

An innovative Kinetic Monte Carlo (KMC) code has been developed, which rules the post-implant kinetics of the defects system in the extremely far-from-the equilibrium conditions caused by the laser irradiation close to the liquid-solid interface. It considers defect diffusion, annihilation and clustering. The code properly implements, consistently to the stochastic formalism, the fast varying local event rates related to the thermal field T(r,t) evolution. This feature of our numerical method represents an important advancement with respect to current state of the art KMC codes. The reduction of the implantation damage and its reorganization in defect aggregates are studied as amore » function of the process conditions. Phosphorus activation efficiency, experimentally determined in similar conditions, has been related to the emerging damage scenario.« less

A thermal control approach for a solar electric propulsion thrust subsystem

NASA Technical Reports Server (NTRS)

Maloy, J. E.; Oglebay, J. C.

1979-01-01

A thrust subsystem thermal control design is defined for a Solar Electric Propulsion System (SEPS) proposed for the comet Halley Flyby/comet Tempel 2 rendezvous mission. A 114 node analytic model, developed and coded on the systems improved numerical differencing analyzer program, was employed. A description of the resulting thrust subsystem thermal design is presented as well as a description of the analytic model and comparisons of the predicted temperature profiles for various SEPS thermal configurations that were generated using this model. It was concluded that: (1) a BIMOD engine system thermal design can be autonomous; (2) an independent thrust subsystem thermal design is feasible; (3) the interface module electronics temperatures can be controlled by a passive radiator and supplementary heaters; (4) maintaining heat pipes above the freezing point would require an additional 322 watts of supplementary heating power for the situation where no thrusters are operating; (5) insulation is required around the power processors, and between the interface module and the avionics module, as well as in those areas which may be subjected to solar heating; and (6) insulation behind the heat pipe radiators is not necessary.

Computer Code for Nanostructure Simulation

NASA Technical Reports Server (NTRS)

Filikhin, Igor; Vlahovic, Branislav

2009-01-01

Due to their small size, nanostructures can have stress and thermal gradients that are larger than any macroscopic analogue. These gradients can lead to specific regions that are susceptible to failure via processes such as plastic deformation by dislocation emission, chemical debonding, and interfacial alloying. A program has been developed that rigorously simulates and predicts optoelectronic properties of nanostructures of virtually any geometrical complexity and material composition. It can be used in simulations of energy level structure, wave functions, density of states of spatially configured phonon-coupled electrons, excitons in quantum dots, quantum rings, quantum ring complexes, and more. The code can be used to calculate stress distributions and thermal transport properties for a variety of nanostructures and interfaces, transport and scattering at nanoscale interfaces and surfaces under various stress states, and alloy compositional gradients. The code allows users to perform modeling of charge transport processes through quantum-dot (QD) arrays as functions of inter-dot distance, array order versus disorder, QD orientation, shape, size, and chemical composition for applications in photovoltaics and physical properties of QD-based biochemical sensors. The code can be used to study the hot exciton formation/relation dynamics in arrays of QDs of different shapes and sizes at different temperatures. It also can be used to understand the relation among the deposition parameters and inherent stresses, strain deformation, heat flow, and failure of nanostructures.

acme: The Amendable Coal-Fire Modeling Exercise. A C++ Class Library for the Numerical Simulation of Coal-Fires

NASA Astrophysics Data System (ADS)

Wuttke, Manfred W.

2017-04-01

At LIAG, we use numerical models to develop and enhance understanding of coupled transport processes and to predict the dynamics of the system under consideration. Topics include geothermal heat utilization, subrosion processes, and spontaneous underground coal fires. Although the details make it inconvenient if not impossible to apply a single code implementation to all systems, their investigations go along similar paths: They all depend on the solution of coupled transport equations. We thus saw a need for a modular code system with open access for the various communities to maximize the shared synergistic effects. To this purpose we develop the oops! ( open object-oriented parallel solutions) - toolkit, a C++ class library for the numerical solution of mathematical models of coupled thermal, hydraulic and chemical processes. This is used to develop problem-specific libraries like acme( amendable coal-fire modeling exercise), a class library for the numerical simulation of coal-fires and applications like kobra (Kohlebrand, german for coal-fire), a numerical simulation code for standard coal-fire models. Basic principle of the oops!-code system is the provision of data types for the description of space and time dependent data fields, description of terms of partial differential equations (pde), their discretisation and solving methods. Coupling of different processes, described by their particular pde is modeled by an automatic timescale-ordered operator-splitting technique. acme is a derived coal-fire specific application library, depending on oops!. If specific functionalities of general interest are implemented and have been tested they will be assimilated into the main oops!-library. Interfaces to external pre- and post-processing tools are easily implemented. Thus a construction kit which can be arbitrarily amended is formed. With the kobra-application constructed with acme we study the processes and propagation of shallow coal seam fires in particular in Xinjiang, China, as well as analyze and interpret results from lab experiments.

Statistical core design methodology using the VIPRE thermal-hydraulics code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lloyd, M.W.; Feltus, M.A.

1994-12-31

This Penn State Statistical Core Design Methodology (PSSCDM) is unique because it not only includes the EPRI correlation/test data standard deviation but also the computational uncertainty for the VIPRE code model and the new composite box design correlation. The resultant PSSCDM equation mimics the EPRI DNBR correlation results well, with an uncertainty of 0.0389. The combined uncertainty yields a new DNBR limit of 1.18 that will provide more plant operational flexibility. This methodology and its associated correlation and uniqe coefficients are for a very particular VIPRE model; thus, the correlation will be specifically linked with the lumped channel and subchannelmore » layout. The results of this research and methodology, however, can be applied to plant-specific VIPRE models.« less

Crack deflection in brittle media with heterogeneous interfaces and its application in shale fracking

NASA Astrophysics Data System (ADS)

Zeng, Xiaguang; Wei, Yujie

Driven by the rapid progress in exploiting unconventional energy resources such as shale gas, there is growing interest in hydraulic fracture of brittle yet heterogeneous shales. In particular, how hydraulic cracks interact with natural weak zones in sedimentary rocks to form permeable cracking networks is of significance in engineering practice. Such a process is typically influenced by crack deflection, material anisotropy, crack-surface friction, crustal stresses, and so on. In this work, we extend the He-Hutchinson theory (He and Hutchinson, 1989) to give the closed-form formulae of the strain energy release rate of a hydraulic crack with arbitrary angles with respect to the crustal stress. The critical conditions in which the hydraulic crack deflects into weak interfaces and exhibits a dependence on crack-surface friction and crustal stress anisotropy are given in explicit formulae. We reveal analytically that, with increasing pressure, hydraulic fracture in shales may sequentially undergo friction locking, mode II fracture, and mixed mode fracture. Mode II fracture dominates the hydraulic fracturing process and the impinging angle between the hydraulic crack and the weak interface is the determining factor that accounts for crack deflection; the lower friction coefficient between cracked planes and the greater crustal stress difference favor hydraulic fracturing. In addition to shale fracking, the analytical solution of crack deflection could be used in failure analysis of other brittle media.

Main steam line break accident simulation of APR1400 using the model of ATLAS facility

NASA Astrophysics Data System (ADS)

Ekariansyah, A. S.; Deswandri; Sunaryo, Geni R.

2018-02-01

A main steam line break simulation for APR1400 as an advanced design of PWR has been performed using the RELAP5 code. The simulation was conducted in a model of thermal-hydraulic test facility called as ATLAS, which represents a scaled down facility of the APR1400 design. The main steam line break event is described in a open-access safety report document, in which initial conditions and assumptionsfor the analysis were utilized in performing the simulation and analysis of the selected parameter. The objective of this work was to conduct a benchmark activities by comparing the simulation results of the CESEC-III code as a conservative approach code with the results of RELAP5 as a best-estimate code. Based on the simulation results, a general similarity in the behavior of selected parameters was observed between the two codes. However the degree of accuracy still needs further research an analysis by comparing with the other best-estimate code. Uncertainties arising from the ATLAS model should be minimized by taking into account much more specific data in developing the APR1400 model.

A flooding induced station blackout analysis for a pressurized water reactor using the RISMC toolkit

DOE PAGES

Mandelli, Diego; Prescott, Steven; Smith, Curtis; ...

2015-05-17

In this paper we evaluate the impact of a power uprate on a pressurized water reactor (PWR) for a tsunami-induced flooding test case. This analysis is performed using the RISMC toolkit: the RELAP-7 and RAVEN codes. RELAP-7 is the new generation of system analysis codes that is responsible for simulating the thermal-hydraulic dynamics of PWR and boiling water reactor systems. RAVEN has two capabilities: to act as a controller of the RELAP-7 simulation (e.g., component/system activation) and to perform statistical analyses. In our case, the simulation of the flooding is performed by using an advanced smooth particle hydrodynamics code calledmore » NEUTRINO. The obtained results allow the user to investigate and quantify the impact of timing and sequencing of events on system safety. The impact of power uprate is determined in terms of both core damage probability and safety margins.« less

SCORE-EVET: a computer code for the multidimensional transient thermal-hydraulic analysis of nuclear fuel rod arrays. [BWR; PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Benedetti, R. L.; Lords, L. V.; Kiser, D. M.

1978-02-01

The SCORE-EVET code was developed to study multidimensional transient fluid flow in nuclear reactor fuel rod arrays. The conservation equations used were derived by volume averaging the transient compressible three-dimensional local continuum equations in Cartesian coordinates. No assumptions associated with subchannel flow have been incorporated into the derivation of the conservation equations. In addition to the three-dimensional fluid flow equations, the SCORE-EVET code ocntains: (a) a one-dimensional steady state solution scheme to initialize the flow field, (b) steady state and transient fuel rod conduction models, and (c) comprehensive correlation packages to describe fluid-to-fuel rod interfacial energy and momentum exchange. Velocitymore » and pressure boundary conditions can be specified as a function of time and space to model reactor transient conditions such as a hypothesized loss-of-coolant accident (LOCA) or flow blockage.« less

MOOSE Implementation of MAMBA

DOE Office of Scientific and Technical Information (OSTI.GOV)

Galloway, Jack; Matthews, Topher

The development of MAMBA is targeted at capturing both core wide CRUD induced power shifts (CIPS) as well as pin-­level CRUD induced localized corrosion (CILC). Both CIPS and CILC require some sort of information from thermal-­hydraulic, neutronics, and fuel performance codes, although the degree of coupling is different for the two effects. Since CIPS necessarily requires a core-­wide power distribution solve, it requires tight coupling with a neutronics code. Conversely, CIPS tends to be an individual pin phenomenon, requiring tight coupling a fuel performance code. As efforts are now focused on coupling MAMBA within the VERA suite, a natural separationmore » has surfaced in which a FORTRAN rewrite of MAMBA is optimal for VERA integration to capture CIPS behavior, while a CILC focused calculation would benefit from a tight coupling with BISON, motivating a MOOSE version of MAMBA.« less

Posttest RELAP5 simulations of the Semiscale S-UT series experiments. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Leonard, M.T.

The RELAP5/MOD1 computer code was used to perform posttest calculations, simulating six experiments, run in the Semiscale Mod-2A facility, investigating the effects of upper head injection on small break transient behavior. The results of these calculations and corresponding test data are presented in this report. An evaluation is made of the capability of RELAP5 to calculate the thermal-hydraulic response of the Mod-2A system over a spectrum of break sizes, with and without the use of upper head injection.

Influence evaluation of loading conditions during pressurized thermal shock transients based on thermal-hydraulics and structural analyses

NASA Astrophysics Data System (ADS)

Katsuyama, Jinya; Uno, Shumpei; Watanabe, Tadashi; Li, Yinsheng

2018-03-01

The thermal hydraulic (TH) behavior of coolant water is a key factor in the structural integrity assessments on reactor pressure vessels (RPVs) of pressurized water reactors (PWRs) under pressurized thermal shock (PTS) events, because the TH behavior may affect the loading conditions in the assessment. From the viewpoint of TH behavior, configuration of plant equipment and their dimensions, and operator action time considerably influence various parameters, such as the temperature and flow rate of coolant water and inner pressure. In this study, to investigate the influence of the operator action time on TH behavior during a PTS event, we developed an analysis model for a typical Japanese PWR plant, including the RPV and the main components of both primary and secondary systems, and performed TH analyses by using a system analysis code called RELAP5. We applied two different operator action times based on the Japanese and the United States (US) rules: Operators may act after 10 min (Japanese rules) and 30 min (the US rules) after the occurrence of PTS events. Based on the results of TH analysis with different operator action times, we also performed structural analyses for evaluating thermal-stress distributions in the RPV during PTS events as loading conditions in the structural integrity assessment. From the analysis results, it was clarified that differences in operator action times significantly affect TH behavior and loading conditions, as the Japanese rule may lead to lower stresses than that under the US rule because an earlier operator action caused lower pressure in the RPV.

Improvement of COBRA-TF for modeling of PWR cold- and hot-legs during reactor transients

NASA Astrophysics Data System (ADS)

Salko, Robert K.

COBRA-TF is a two-phase, three-field (liquid, vapor, droplets) thermal-hydraulic modeling tool that has been developed by the Pacific Northwest Laboratory under sponsorship of the NRC. The code was developed for Light Water Reactor analysis starting in the 1980s; however, its development has continued to this current time. COBRA-TF still finds wide-spread use throughout the nuclear engineering field, including nuclear-power vendors, academia, and research institutions. It has been proposed that extension of the COBRA-TF code-modeling region from vessel-only components to Pressurized Water Reactor (PWR) coolant-line regions can lead to improved Loss-of-Coolant Accident (LOCA) analysis. Improved modeling is anticipated due to COBRA-TF's capability to independently model the entrained-droplet flow-field behavior, which has been observed to impact delivery to the core region[1]. Because COBRA-TF was originally developed for vertically-dominated, in-vessel, sub-channel flow, extension of the COBRA-TF modeling region to the horizontal-pipe geometries of the coolant-lines required several code modifications, including: • Inclusion of the stratified flow regime into the COBRA-TF flow regime map, along with associated interfacial drag, wall drag and interfacial heat transfer correlations, • Inclusion of a horizontal-stratification force between adjacent mesh cells having unequal levels of stratified flow, and • Generation of a new code-input interface for the modeling of coolant-lines. The sheer number of COBRA-TF modifications that were required to complete this work turned this project into a code-development project as much as it was a study of thermal-hydraulics in reactor coolant-lines. The means for achieving these tasks shifted along the way, ultimately leading the development of a separate, nearly completely independent one-dimensional, two-phase-flow modeling code geared toward reactor coolant-line analysis. This developed code has been named CLAP, for Coolant-Line-Analysis Package. Versions were created that were both coupled to COBRA-TF and standalone, with the most recent version being a standalone code. This code performs a separate, simplified, 1-D solution of the conservation equations while making special considerations for coolant-line geometry and flow phenomena. The end of this project saw a functional code package that demonstrates a stable numerical solution and that has gone through a series of Validation and Verification tests using the Two-Phase Testing Facility (TPTF) experimental data[2]. The results indicate that CLAP is under-performing RELAP5-MOD3 in predicting the experimental void of the TPTF facility in some cases. There is no apparent pattern, however, to point to a consistent type of case that the code fails to predict properly (e.g., low-flow, high-flow, discharging to full vessel, or discharging to empty vessel). Pressure-profile predictions are sometimes unrealistic, which indicates that there may be a problem with test-case boundary conditions or with the coupling of continuity and momentum equations in the solution algorithm. The code does predict the flow regime correctly for all cases with the stratification-force model off. Turning the stratification model on can cause the low-flow case void profiles to over-react to the force and the flow regime to transition out of stratified flow. The code would benefit from an increased amount of Validation & Verification testing. The development of CLAP was significant, as it is a cleanly written, logical representation of the reactor coolant-line geometry. It is stable and capable of modeling basic flow physics in the reactor coolant-line. Code development and debugging required the temporary removal of the energy equation and mass-transfer terms in governing equations. The reintroduction of these terms will allow future coupling to RELAP and re-coupling with COBRA-TF. Adding in more applicable entrainment and de-entrainment models would allow the capture of more advanced physics in the coolant-line that can be expected during Loss-of-Coolant Accident. One of the package's benefits is its ability to be used as a platform for future coolant-line model development and implementation, including capturing of the important de-entrainment behavior in reactor hot-legs (steam-binding effect) and flow convection in the upper-plenum region of the vessel.

Modeling the effects of hydraulic stimulation on geothermal reservoirs

NASA Astrophysics Data System (ADS)

De Simone, Silvia; Vilarrasa, Victor; Carrera, Jesús; Alcolea, Andrés; Meier, Peter

2013-04-01

Geothermal energy represents a huge power source that can provide clean energy in potentially unlimited supply. When designing geothermal energy production from deep hot rocks, permeability is considered to control the economic efficiency of the heat extraction operations. In fact, a high permeability heat exchanger is required to achieve a cost-competitive power generation. The typical procedure entails intercepting naturally fractured rocks and enhancing their permeability by means of stimulation. Hydraulic stimulation is the most widely used method. It involves the massive injection of a large volume of water at high flow rates to increase the downhole pore pressure. This overpressure reduces the effective stresses, which tends to induce shearing along the fracture planes. In this way permeability is enhanced due to dilatancy, especially in the direction perpendicular to shear. These processes usually trigger microseismic events, which are sometimes of sufficient magnitude to be felt by the local population. This causes a negative impact on the local population and may compromise the continuation of the project. Hence, understanding the mechanisms triggering these induced micro-earthquakes is important to properly design and manage geothermal stimulation and operations so as to prevent them. We analyzed the thermo-hydro-mechanical response of a fractured deep rock mass subjected to hydraulic stimulation. Considering that seismicity is triggered when failure condition are reached, we studied the variation of the stress regime due to the hydraulic and thermal perturbations during fluid injection. Starting with a simplified model with constant permeability fault zones, more sophisticated schemes are considered to simulate the behavior of the discontinuity zones, including permeability variation associated to temperature, pressure and stress regime changes. Numerical simulations are performed using the finite element numerical code CODE_BRIGHT, which allows to solve fully coupled thermo-hydro-mechanical problems. Results allowed to estimate the impact of the hydraulic stimulation on the overall behavior.

Constitutive relations in TRAC-P1A

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Saha, P.

1980-08-01

The purpose of this document is to describe the basic thermal-hydraulic models and correlations that are in the TRAC-P1A code, as released in March 1979. It is divided into two parts, A and B. Part A describes the models in the three-dimensional vessel module of TRAC, whereas Part B focuses on the loop components that are treated by one-dimensional formulations. The report follows the format of the questions prepared by the Analysis Development Branch of USNRC and the questionnaire has been attached to this document for completeness. Concerted efforts have been made in understanding the present models in TRAC-P1A bymore » going through the FORTRAN listing of the code. Some discrepancies between the code and the TRAC-P1A manual have been found. These are pointed out in this document. Efforts have also been made to check the TRAC references for the range of applicability of the models and correlations used in the code. 26 refs., 5 figs., 1 tab.« less

Recovery Act. Development and Validation of an Advanced Stimulation Prediction Model for Enhanced Geothermal System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gutierrez, Marte

The research project aims to develop and validate an advanced computer model that can be used in the planning and design of stimulation techniques to create engineered reservoirs for Enhanced Geothermal Systems. The specific objectives of the proposal are to: 1) Develop a true three-dimensional hydro-thermal fracturing simulator that is particularly suited for EGS reservoir creation. 2) Perform laboratory scale model tests of hydraulic fracturing and proppant flow/transport using a polyaxial loading device, and use the laboratory results to test and validate the 3D simulator. 3) Perform discrete element/particulate modeling of proppant transport in hydraulic fractures, and use the resultsmore » to improve understand of proppant flow and transport. 4) Test and validate the 3D hydro-thermal fracturing simulator against case histories of EGS energy production. 5) Develop a plan to commercialize the 3D fracturing and proppant flow/transport simulator. The project is expected to yield several specific results and benefits. Major technical products from the proposal include: 1) A true-3D hydro-thermal fracturing computer code that is particularly suited to EGS, 2) Documented results of scale model tests on hydro-thermal fracturing and fracture propping in an analogue crystalline rock, 3) Documented procedures and results of discrete element/particulate modeling of flow and transport of proppants for EGS applications, and 4) Database of monitoring data, with focus of Acoustic Emissions (AE) from lab scale modeling and field case histories of EGS reservoir creation.« less

Recovery Act. Development and Validation of an Advanced Stimulation Prediction Model for Enhanced Geothermal Systems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gutierrez, Marte

2013-12-31

This research project aims to develop and validate an advanced computer model that can be used in the planning and design of stimulation techniques to create engineered reservoirs for Enhanced Geothermal Systems. The specific objectives of the proposal are to; Develop a true three-dimensional hydro-thermal fracturing simulator that is particularly suited for EGS reservoir creation; Perform laboratory scale model tests of hydraulic fracturing and proppant flow/transport using a polyaxial loading device, and use the laboratory results to test and validate the 3D simulator; Perform discrete element/particulate modeling of proppant transport in hydraulic fractures, and use the results to improve understandmore » of proppant flow and transport; Test and validate the 3D hydro-thermal fracturing simulator against case histories of EGS energy production; and Develop a plan to commercialize the 3D fracturing and proppant flow/transport simulator. The project is expected to yield several specific results and benefits. Major technical products from the proposal include; A true-3D hydro-thermal fracturing computer code that is particularly suited to EGS; Documented results of scale model tests on hydro-thermal fracturing and fracture propping in an analogue crystalline rock; Documented procedures and results of discrete element/particulate modeling of flow and transport of proppants for EGS applications; and Database of monitoring data, with focus of Acoustic Emissions (AE) from lab scale modeling and field case histories of EGS reservoir creation.« less

Comparison of the PHISICS/RELAP5-3D Ring and Block Model Results for Phase I of the OECD MHTGR-350 Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom

2014-04-01

The INL PHISICS code system consists of three modules providing improved core simulation capability: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been finalized, and as part of the code verification and validation program the exercises defined for Phase I of the OECD/NEA MHTGR 350 MW Benchmark were completed. This paper provides an overview of the MHTGR Benchmark, and presents selected results of the three steady state exercises 1-3 defined for Phase I. For Exercise 1,more » a stand-alone steady-state neutronics solution for an End of Equilibrium Cycle Modular High Temperature Reactor (MHTGR) was calculated with INSTANT, using the provided geometry, material descriptions, and detailed cross-section libraries. Exercise 2 required the modeling of a stand-alone thermal fluids solution. The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 combined the first two exercises in a coupled neutronics and thermal fluids solution, and the coupled code suite PHISICS/RELAP5-3D was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of the traditional RELAP5-3D “ring” model approach vs. a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity of the block model is illustrated with comparison results on the temperature, power density and flux distributions, and the typical under-predictions produced by the ring model approach are highlighted.« less

RAMONA-4B a computer code with three-dimensional neutron kinetics for BWR and SBWR system transient - user`s manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rohatgi, U.S.; Cheng, H.S.; Khan, H.J.

This document is the User`s Manual for the Boiling Water Reactor (BWR), and Simplified Boiling Water Reactor (SBWR) systems transient code RAMONA-4B. The code uses a three-dimensional neutron-kinetics model coupled with a multichannel, nonequilibrium, drift-flux, phase-flow model of the thermal hydraulics of the reactor vessel. The code is designed to analyze a wide spectrum of BWR core and system transients. Chapter 1 gives an overview of the code`s capabilities and limitations; Chapter 2 describes the code`s structure, lists major subroutines, and discusses the computer requirements. Chapter 3 is on code, auxillary codes, and instructions for running RAMONA-4B on Sun SPARCmore » and IBM Workstations. Chapter 4 contains component descriptions and detailed card-by-card input instructions. Chapter 5 provides samples of the tabulated output for the steady-state and transient calculations and discusses the plotting procedures for the steady-state and transient calculations. Three appendices contain important user and programmer information: lists of plot variables (Appendix A) listings of input deck for sample problem (Appendix B), and a description of the plotting program PAD (Appendix C). 24 refs., 18 figs., 11 tabs.« less

Numerical investigation of thermal-hydraulic performance of channel with protrusions by turbulent cross flow jet

NASA Astrophysics Data System (ADS)

Sahu, M. K.; Pandey, K. M.; Chatterjee, S.

2018-05-01

In this two dimensional numerical investigation, small rectangular channel with right angled triangular protrusions in the bottom wall of test section is considered. A slot nozzle is placed at the middle of top wall of channel which impinges air normal to the protruded surface. A duct flow and nozzle flow combined to form cross flow which is investigated for heat transfer enhancement of protruded channel. The governing equations for continuity, momentum, energy along with SST k-ω turbulence model are solved with finite volume based Computational fluid dynamics code ANSYS FLUENT 14.0. The range of duct Reynolds number considered for this analysis is 8357 to 51760. The ratios of pitch of protrusion to height of duct considered are 0.5, 0.64 and 0.82. The ratios of height of protrusion to height of duct considered are 0.14, 0.23 and 0.29. The effect of duct Reynolds number, pitch and height of protrusion on thermal-hydraulic performance is studied under cross flow condition. It is found that heat transfer rate is more at relatively larger pitch and small pressure drop is found in case of low height of protrusion.

Multidimensional Mixing Behavior of Steam-Water Flow in a Downcomer Annulus During LBLOCA Reflood Phase with a Direct Vessel Injection Mode

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kwon, Tae-Soon; Yun, Byong-Jo; Euh, Dong-Jin

Multidimensional thermal-hydraulic behavior in the downcomer annulus of a pressurized water reactor (PWR) vessel with a direct vessel injection mode is presented based on the experimental observation in the MIDAS (multidimensional investigation in downcomer annulus simulation) steam-water test facility. From the steady-state test results to simulate the late reflood phase of a large-break loss-of-coolant accident (LBLOCA), isothermal lines show the multidimensional phenomena of a phasic interaction between steam and water in the downcomer annulus very well. MIDAS is a steam-water separate effect test facility, which is 1/4.93 linearly scaled down to a 1400-MW(electric) PWR type of a nuclear reactor, focusedmore » on understanding multidimensional thermal-hydraulic phenomena in a downcomer annulus with various types of safety injection during the refill or reflood phase of an LBLOCA. The initial and the boundary conditions are scaled from the pretest analysis based on the preliminary calculation using the TRAC code. The superheated steam with a superheating degree of 80 K at a given downcomer pressure of 180 kPa is injected equally through three intact cold legs into the downcomer.« less

New Laboratory Observations of Thermal Pressurization Weakening

NASA Astrophysics Data System (ADS)

Badt, N.; Tullis, T. E.; Hirth, G.

2017-12-01

Dynamic frictional weakening due to pore fluid thermal pressurization has been studied under elevated confining pressure in the laboratory, using a rotary-shear apparatus having a sample with independent pore pressure and confining pressure systems. Thermal pressurization is directly controlled by the permeability of the rocks, not only for the initiation of high-speed frictional weakening but also for a subsequent sequence of high-speed sliding events. First, the permeability is evaluated at different effective pressures using a method where the pore pressure drop and the flow-through rate are compared using Darcy's Law as well as a pore fluid oscillation method, the latter method also permitting measurement of the storage capacity. Then, the samples undergo a series of high-speed frictional sliding segments at a velocity of 2.5 mm/s, under an applied confining pressure and normal stress of 45 MPa and 50 MPa, respectively, and an initial pore pressure of 25 MPa. Finally the rock permeability and storage capacity are measured again to assess the evolution of the rock's pore fluid properties. For samples with a permeability of 10-20 m2 thermal pressurization promotes a 40% decrease in strength. However, after a sequence of three high-speed sliding events, the magnitude of weakening diminishes progressively from 40% to 15%. The weakening events coincide with dilation of the sliding interface. Moreover, the decrease in the weakening degree with progressive fast-slip events suggest that the hydraulic diffusivity may increase locally near the sliding interface during thermal pressurization-enhanced slip. This could result from stress- or thermally-induced damage to the host rock, which would perhaps increase both permeability and storage capacity, and so possibly decrease the susceptibility of dynamic weakening due to thermal pressurization in subsequent high-speed sliding events.

BODYFIT-1FE: a computer code for three-dimensional steady-state/transient single-phase rod-bundle thermal-hydraulic analysis. Draft report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chen, B.C.J.; Sha, W.T.; Doria, M.L.

1980-11-01

The governing equations, i.e., conservation equations for mass, momentum, and energy, are solved as a boundary-value problem in space and an initial-value problem in time. BODYFIT-1FE code uses the technique of boundary-fitted coordinate systems where all the physical boundaries are transformed to be coincident with constant coordinate lines in the transformed space. By using this technique, one can prescribe boundary conditions accurately without interpolation. The transformed governing equations in terms of the boundary-fitted coordinates are then solved by using implicit cell-by-cell procedure with a choice of either central or upwind convective derivatives. It is a true benchmark rod-bundle code withoutmore » invoking any assumptions in the case of laminar flow. However, for turbulent flow, some empiricism must be employed due to the closure problem of turbulence modeling. The detailed velocity and temperature distributions calculated from the code can be used to benchmark and calibrate empirical coefficients employed in subchannel codes and porous-medium analyses.« less

77 FR 5063 - Advisory Committee on Reactor Safeguards (ACRS) Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-02-01

... Subcommittee on Thermal-Hydraulics Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal-Hydraulics... Regulatory Guide (1.79), ``Preoperational Testing of Emergency Core Cooling Systems for Pressurized Water... Water Reactors.'' The Subcommittee will hear presentations by and hold discussions with the NRC staff...

Posttest analysis of international standard problem 10 using RELAP4/MOD7. [PWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hsu, M.; Davis, C.B.; Peterson, A.C. Jr.

RELAP4/MOD7, a best estimate computer code for the calculation of thermal and hydraulic phenomena in a nuclear reactor or related system, is the latest version in the RELAP4 code development series. This paper evaluates the capability of RELAP4/MOD7 to calculate refill/reflood phenomena. This evaluation uses the data of International Standard Problem 10, which is based on West Germany's KWU PKL refill/reflood experiment K9A. The PKL test facility represents a typical West German four-loop, 1300 MW pressurized water reactor (PWR) in reduced scale while maintaining prototypical volume-to-power ratio. The PKL facility was designed to specifically simulate the refill/reflood phase of amore » hypothetical loss-of-coolant accident (LOCA).« less

Characterization of an alluvial aquifer with thermal tracer tomography

NASA Astrophysics Data System (ADS)

Somogyvári, Márk; Bayer, Peter

2017-04-01

In the summer of 2015, a series of thermal tracer tests was performed at the Widen field site in northeast Switzerland. At this site numerous hydraulic, tracer, geophysical and hydrogeophysical field tests have been conducted in the past to investigate a shallow alluvial aquifer. The goals of the campaign in 2015 were to design a cost-effective thermal tracer tomography setup and to validate the concept of travel time-based thermal tracer tomography under field conditions. Thermal tracer tomography uses repeated thermal tracer injections with different injection depths and distributed temperature measurements to map the hydraulic conductivity distribution of a heterogeneous aquifer. The tracer application was designed with minimal experimental time and cost. Water was heated in inflatable swimming pools using direct sunlight of the warm summer days, and it was injected as low temperature pulses in a well. Because of the small amount of injected heat, no long recovery times were required between the repeated heat tracer injections and every test started from natural thermal conditions. At Widen, four thermal tracer tests were performed during a period of three days. Temperatures were measured in one downgradient well using a distributed temperature measurement system installed at seven depth points. Totally 12 temperature breakthrough curves were collected. Travel time based tomographic inversion assumes that thermal transport is dominated by advection and the travel time of the thermal tracer can be related to the hydraulic conductivities of the aquifer. This assumption is valid in many shallow porous aquifers where the groundwater flow is fast. In our application, the travel time problem was treated by a tomographic solver, analogous to seismic tomography, to derive the hydraulic conductivity distribution. At the test site, a two-dimensional cross-well hydraulic conductivity profile was reconstructed with the travel time based inversion. The reconstructed profile corresponds well with the findings of the earlier hydraulic and geophysical experiments at the site.

Discrimination of soil hydraulic properties by combined thermal infrared and microwave remote sensing

NASA Technical Reports Server (NTRS)

Vandegriend, A. A.; Oneill, P. E.

1986-01-01

Using the De Vries models for thermal conductivity and heat capacity, thermal inertia was determined as a function of soil moisture for 12 classes of soil types ranging from sand to clay. A coupled heat and moisture balance model was used to describe the thermal behavior of the top soil, while microwave remote sensing was used to estimate the soil moisture content of the same top soil. Soil hydraulic parameters are found to be very highly correlated with the combination of soil moisture content and thermal inertia at the same moisture content. Therefore, a remotely sensed estimate of the thermal behavior of the soil from diurnal soil temperature observations and an independent remotely sensed estimate of soil moisture content gives the possibility of estimating soil hydraulic properties by remote sensing.

Development and Demonstration of Material Properties Database and Software for the Simulation of Flow Properties in Cementitious Materials

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, F.; Flach, G.

This report describes work performed by the Savannah River National Laboratory (SRNL) in fiscal year 2014 to develop a new Cementitious Barriers Project (CBP) software module designated as FLOExcel. FLOExcel incorporates a uniform database to capture material characterization data and a GoldSim model to define flow properties for both intact and fractured cementitious materials and estimate Darcy velocity based on specified hydraulic head gradient and matric tension. The software module includes hydraulic parameters for intact cementitious and granular materials in the database and a standalone GoldSim framework to manipulate the data. The database will be updated with new data asmore » it comes available. The software module will later be integrated into the next release of the CBP Toolbox, Version 3.0. This report documents the development efforts for this software module. The FY14 activities described in this report focused on the following two items that form the FLOExcel package; 1) Development of a uniform database to capture CBP data for cementitious materials. In particular, the inclusion and use of hydraulic properties of the materials are emphasized; and 2) Development of algorithms and a GoldSim User Interface to calculate hydraulic flow properties of degraded and fractured cementitious materials. Hydraulic properties are required in a simulation of flow through cementitious materials such as Saltstone, waste tank fill grout, and concrete barriers. At SRNL these simulations have been performed using the PORFLOW code as part of Performance Assessments for salt waste disposal and waste tank closure.« less

A THC Simulator for Modeling Fluid-Rock Interactions

NASA Astrophysics Data System (ADS)

Hamidi, Sahar; Galvan, Boris; Heinze, Thomas; Miller, Stephen

2014-05-01

Fluid-rock interactions play an essential role in many earth processes, from a likely influence on earthquake nucleation and aftershocks, to enhanced geothermal system, carbon capture and storage (CCS), and underground nuclear waste repositories. In THC models, two-way interactions between different processes (thermal, hydraulic and chemical) are present. Fluid flow influences the permeability of the rock especially if chemical reactions are taken into account. On one hand solute concentration influences fluid properties while, on the other hand, heat can affect further chemical reactions. Estimating heat production from a naturally fractured geothermal systems remains a complex problem. Previous works are typically based on a local thermal equilibrium assumption and rarely consider the salinity. The dissolved salt in fluid affects the hydro- and thermodynamical behavior of the system by changing the hydraulic properties of the circulating fluid. Coupled thermal-hydraulic-chemical models (THC) are important for investigating these processes, but what is needed is a coupling to mechanics to result in THMC models. Although similar models currently exist (e.g. PFLOTRAN), our objective here is to develop algorithms for implementation using the Graphics Processing Unit (GPU) computer architecture to be run on GPU clusters. To that aim, we present a two-dimensional numerical simulation of a fully coupled non-isothermal non-reactive solute flow. The thermal part of the simulation models heat transfer processes for either local thermal equilibrium or nonequilibrium cases, and coupled to a non-reactive mass transfer described by a non-linear diffusion/dispersion model. The flow process of the model includes a non-linear Darcian flow for either saturated or unsaturated scenarios. For the unsaturated case, we use the Richards' approximation for a mixture of liquid and gas phases. Relative permeability and capillary pressure are determined by the van Genuchten relations. Permeability of rock is controlled by porosity, which is itself related to effective stress. The theoretical model is solved using explicit finite differences, and runs in parallel mode with OpenMP. The code is fully modular so that any combination of current THC processes, one- and two-phase, can be chosen. Future developments will include dissolution and precipitation of chemical components in addition to chemical erosion.

Portable Life Support Subsystem Thermal Hydraulic Performance Analysis

NASA Technical Reports Server (NTRS)

Barnes, Bruce; Pinckney, John; Conger, Bruce

2010-01-01

This paper presents the current state of the thermal hydraulic modeling efforts being conducted for the Constellation Space Suit Element (CSSE) Portable Life Support Subsystem (PLSS). The goal of these efforts is to provide realistic simulations of the PLSS under various modes of operation. The PLSS thermal hydraulic model simulates the thermal, pressure, flow characteristics, and human thermal comfort related to the PLSS performance. This paper presents modeling approaches and assumptions as well as component model descriptions. Results from the models are presented that show PLSS operations at steady-state and transient conditions. Finally, conclusions and recommendations are offered that summarize results, identify PLSS design weaknesses uncovered during review of the analysis results, and propose areas for improvement to increase model fidelity and accuracy.

75 FR 51499 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-08-20

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulics Phenomena The ACRS Subcommittee on Thermal Hydraulics Phenomena will hold a meeting on September 7, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting...

75 FR 57536 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-09-21

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulic Phenomena The ACRS Subcommittee on Thermal Hydraulic Phenomena will hold a meeting on October 18, 2010, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The entire meeting...

76 FR 53979 - Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2011-08-30

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS); Meeting of the ACRS Subcommittee on Thermal Hydraulics Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal Hydraulics... Revision 4 to Regulatory Guide 1.82, ``Water Sources for Long-Term Recirculation Cooling Following a Loss...

Numerical Analysis of Coolant Flow and Heat Transfer in ITER Diagnostic First Wall

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khodak, A.; Loesser, G.; Zhai, Y.

2015-07-24

We performed numerical simulations of the ITER Diagnostic First Wall (DFW) using ANSYS workbench. During operation DFW will include solid main body as well as liquid coolant. Thus thermal and hydraulic analysis of the DFW was performed using conjugated heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously fluid dynamics analysis was performed only in the liquid part. This approach includes interface between solid and liquid part of the systemAnalysis was performed using ANSYS CFX software. CFX software allows solution of heat transfer equations in solid and liquid part, and solution ofmore » the flow equations in the liquid part. Coolant flow in the DFW was assumed turbulent and was resolved using Reynolds averaged Navier-Stokes equations with Shear Stress Transport turbulence model. Meshing was performed using CFX method available within ANSYS. The data cloud for thermal loading consisting of volumetric heating and surface heating was imported into CFX Volumetric heating source was generated using Attila software. Surface heating was obtained using radiation heat transfer analysis. Our results allowed us to identify areas of excessive heating. Proposals for cooling channel relocation were made. Additional suggestions were made to improve hydraulic performance of the cooling system.« less

Focused ultrasound thermal therapy system with ultrasound image guidance and temperature measurement feedback.

PubMed

Lin, Kao-Han; Young, Sun-Yi; Hsu, Ming-Chuan; Chan, Hsu; Chen, Yung-Yaw; Lin, Win-Li

2008-01-01

In this study, we developed a focused ultrasound (FUS) thermal therapy system with ultrasound image guidance and thermocouple temperature measurement feedback. Hydraulic position devices and computer-controlled servo motors were used to move the FUS transducer to the desired location with the measurement of actual movement by linear scale. The entire system integrated automatic position devices, FUS transducer, power amplifier, ultrasound image system, and thermocouple temperature measurement into a graphical user interface. For the treatment procedure, a thermocouple was implanted into a targeted treatment region in a tissue-mimicking phantom under ultrasound image guidance, and then the acoustic interference pattern formed by image ultrasound beam and low-power FUS beam was employed as image guidance to move the FUS transducer to have its focal zone coincident with the thermocouple tip. The thermocouple temperature rise was used to determine the sonication duration for a suitable thermal lesion as a high power was turned on and ultrasound image was used to capture the thermal lesion formation. For a multiple lesion formation, the FUS transducer was moved under the acoustic interference guidance to a new location and then it sonicated with the same power level and duration. This system was evaluated and the results showed that it could perform two-dimensional motion control to do a two-dimensional thermal therapy with a small localization error 0.5 mm. Through the user interface, the FUS transducer could be moved to heat the target region with the guidance of ultrasound image and acoustic interference pattern. The preliminary phantom experimental results demonstrated that the system could achieve the desired treatment plan satisfactorily.

Three-Dimensional Modeling of Fracture Clusters in Geothermal Reservoirs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ghassemi, Ahmad

The objective of this is to develop a 3-D numerical model for simulating mode I, II, and III (tensile, shear, and out-of-plane) propagation of multiple fractures and fracture clusters to accurately predict geothermal reservoir stimulation using the virtual multi-dimensional internal bond (VMIB). Effective development of enhanced geothermal systems can significantly benefit from improved modeling of hydraulic fracturing. In geothermal reservoirs, where the temperature can reach or exceed 350oC, thermal and poro-mechanical processes play an important role in fracture initiation and propagation. In this project hydraulic fracturing of hot subsurface rock mass will be numerically modeled by extending the virtual multiplemore » internal bond theory and implementing it in a finite element code, WARP3D, a three-dimensional finite element code for solid mechanics. The new constitutive model along with the poro-thermoelastic computational algorithms will allow modeling the initiation and propagation of clusters of fractures, and extension of pre-existing fractures. The work will enable the industry to realistically model stimulation of geothermal reservoirs. The project addresses the Geothermal Technologies Office objective of accurately predicting geothermal reservoir stimulation (GTO technology priority item). The project goal will be attained by: (i) development of the VMIB method for application to 3D analysis of fracture clusters; (ii) development of poro- and thermoelastic material sub-routines for use in 3D finite element code WARP3D; (iii) implementation of VMIB and the new material routines in WARP3D to enable simulation of clusters of fractures while accounting for the effects of the pore pressure, thermal stress and inelastic deformation; (iv) simulation of 3D fracture propagation and coalescence and formation of clusters, and comparison with laboratory compression tests; and (v) application of the model to interpretation of injection experiments (planned by our industrial partner) with reference to the impact of the variations in injection rate and temperature, rock properties, and in-situ stress.« less

Analysis of PANDA Passive Containment Cooling Steady-State Tests with the Spectra Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stempniewicz, Marek M

2000-07-15

Results of post test simulation of the PANDA passive containment cooling (PCC) steady-state tests (S-series tests), performed at the PANDA facility at the Paul Scherrer Institute, Switzerland, are presented. The simulation has been performed using the computer code SPECTRA, a thermal-hydraulic code, designed specifically for analyzing containment behavior of nuclear power plants.Results of the present calculations are compared to the measurement data as well as the results obtained earlier with the codes MELCOR, TRAC-BF1, and TRACG. The calculated PCC efficiencies are somewhat lower than the measured values. Similar underestimation of PCC efficiencies had been obtained in the past, with themore » other computer codes. To explain this difference, it is postulated that condensate coming into the tubes forms a stream of liquid in one or two tubes, leaving most of the tubes unaffected. The condensate entering the water box is assumed to fall down in the form of droplets. With these assumptions, the results calculated with SPECTRA are close to the experimental data.It is concluded that the SPECTRA code is a suitable tool for analyzing containments of advanced reactors, equipped with passive containment cooling systems.« less

Horizontal steam generator thermal-hydraulics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ubra, O.; Doubek, M.

1995-09-01

Horizontal steam generators are typical components of nuclear power plants with pressure water reactor type VVER. Thermal-hydraulic behavior of horizontal steam generators is very different from the vertical U-tube steam generator, which has been extensively studied for several years. To contribute to the understanding of the horizontal steam generator thermal-hydraulics a computer program for 3-D steady state analysis of the PGV-1000 steam generator has been developed. By means of this computer program, a detailed thermal-hydraulic and thermodynamic study of the horizontal steam generator PGV-1000 has been carried out and a set of important steam generator characteristics has been obtained. Themore » 3-D distribution of the void fraction and 3-D level profile as functions of load and secondary side pressure have been investigated and secondary side volumes and masses as functions of load and pressure have been evaluated. Some of the interesting results of calculations are presented in the paper.« less

Numerical study of dam-break induced tsunami-like bore with a hump of different slopes

NASA Astrophysics Data System (ADS)

Cheng, Du; Zhao, Xi-zeng; Zhang, Da-ke; Chen, Yong

2017-12-01

Numerical simulation of dam-break wave, as an imitation of tsunami hydraulic bore, with a hump of different slopes is performed in this paper using an in-house code, named a Constrained Interpolation Profile (CIP)-based model. The model is built on a Cartesian grid system with the Navier Stokes equations using a CIP method for the flow solver, and employs an immersed boundary method (IBM) for the treatment of solid body boundary. A more accurate interface capturing scheme, the Tangent of hyperbola for interface capturing/Slope weighting (THINC/SW) scheme, is adopted as the interface capturing method. Then, the CIP-based model is applied to simulate the dam break flow problem in a bumpy channel. Considerable attention is paid to the spilling type reflected bore, the following spilling type wave breaking, free surface profiles and water level variations over time. Computations are compared with available experimental data and other numerical results quantitatively and qualitatively. Further investigation is conducted to analyze the influence of variable slopes on the flow features of the tsunami-like bore.

1D-3D coupling for hydraulic system transient simulations

NASA Astrophysics Data System (ADS)

Wang, Chao; Nilsson, Håkan; Yang, Jiandong; Petit, Olivier

2017-01-01

This work describes a coupling between the 1D method of characteristics (MOC) and the 3D finite volume method of computational fluid dynamics (CFD). The coupling method is applied to compressible flow in hydraulic systems. The MOC code is implemented as a set of boundary conditions in the OpenFOAM open source CFD software. The coupling is realized by two linear equations originating from the characteristics equation and the Riemann constant equation, respectively. The coupling method is validated using three simple water hammer cases and several coupling configurations. The accuracy and robustness are investigated with respect to the mesh size ratio across the interface, and 3D flow features close to the interface. The method is finally applied to the transient flow caused by the closing and opening of a knife valve (gate) in a pipe, where the flow is driven by the difference in free surface elevation between two tanks. A small region surrounding the moving gate is resolved by CFD, using a dynamic mesh library, while the rest of the system is modeled by MOC. Minor losses are included in the 1D region, corresponding to the contraction of the flow from the upstream tank into the pipe, a separate stationary flow regulation valve, and a pipe bend. The results are validated with experimental data. A 1D solution is provided for comparison, using the static gate characteristics obtained from steady-state CFD simulations.

Core follow calculation with the nTRACER numerical reactor and verification using power reactor measurement data

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jung, Y. S.; Joo, H. G.; Yoon, J. I.

The nTRACER direct whole core transport code employing the planar MOC solution based 3-D calculation method, the subgroup method for resonance treatment, the Krylov matrix exponential method for depletion, and a subchannel thermal/hydraulic calculation solver was developed for practical high-fidelity simulation of power reactors. Its accuracy and performance is verified by comparing with the measurement data obtained for three pressurized water reactor cores. It is demonstrated that accurate and detailed multi-physic simulation of power reactors is practically realizable without any prior calculations or adjustments. (authors)

Advanced Pellet Cladding Interaction Modeling Using the US DOE CASL Fuel Performance Code: Peregrine

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jason Hales; Various

The US DOE’s Consortium for Advanced Simulation of LWRs (CASL) program has undertaken an effort to enhance and develop modeling and simulation tools for a virtual reactor application, including high fidelity neutronics, fluid flow/thermal hydraulics, and fuel and material behavior. The fuel performance analysis efforts aim to provide 3-dimensional capabilities for single and multiple rods to assess safety margins and the impact of plant operation and fuel rod design on the fuel thermomechanical- chemical behavior, including Pellet-Cladding Interaction (PCI) failures and CRUD-Induced Localized Corrosion (CILC) failures in PWRs. [1-3] The CASL fuel performance code, Peregrine, is an engineering scale codemore » that is built upon the MOOSE/ELK/FOX computational FEM framework, which is also common to the fuel modeling framework, BISON [4,5]. Peregrine uses both 2-D and 3-D geometric fuel rod representations and contains a materials properties and fuel behavior model library for the UO2 and Zircaloy system common to PWR fuel derived from both open literature sources and the FALCON code [6]. The primary purpose of Peregrine is to accurately calculate the thermal, mechanical, and chemical processes active throughout a single fuel rod during operation in a reactor, for both steady state and off-normal conditions.« less

RELAP5-3D Results for Phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom

2012-06-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requiresmore » participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2.« less

RELAP5-3D results for phase I (Exercise 2) of the OECD/NEA MHTGR-350 MW benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Strydom, G.; Epiney, A. S.

2012-07-01

The coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D has recently been initiated at the Idaho National Laboratory (INL) to provide a fully coupled prismatic Very High Temperature Reactor (VHTR) system modeling capability as part of the NGNP methods development program. The PHISICS code consists of three modules: INSTANT (performing 3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and a perturbation/mixer module. As part of the verification and validation activities, steady state results have been obtained for Exercise 2 of Phase I of the newly-defined OECD/NEA MHTGR-350 MW Benchmark. This exercise requiresmore » participants to calculate a steady-state solution for an End of Equilibrium Cycle 350 MW Modular High Temperature Reactor (MHTGR), using the provided geometry, material, and coolant bypass flow description. The paper provides an overview of the MHTGR Benchmark and presents typical steady state results (e.g. solid and gas temperatures, thermal conductivities) for Phase I Exercise 2. Preliminary results are also provided for the early test phase of Exercise 3 using a two-group cross-section library and the Relap5-3D model developed for Exercise 2. (authors)« less

Overview and Current Status of Analyses of Potential LEU Design Concepts for TREAT

DOE Office of Scientific and Technical Information (OSTI.GOV)

Connaway, H. M.; Kontogeorgakos, D. C.; Papadias, D. D.

2015-10-01

Neutronic and thermal-hydraulic analyses have been performed to evaluate the performance of different low-enriched uranium (LEU) fuel design concepts for the conversion of the Transient Reactor Test Facility (TREAT) from its current high-enriched uranium (HEU) fuel. TREAT is an experimental reactor developed to generate high neutron flux transients for the testing of nuclear fuels. The goal of this work was to identify an LEU design which can maintain the performance of the existing HEU core while continuing to operate safely. A wide variety of design options were considered, with a focus on minimizing peak fuel temperatures and optimizing the powermore » coupling between the TREAT core and test samples. Designs were also evaluated to ensure that they provide sufficient reactivity and shutdown margin for each control rod bank. Analyses were performed using the core loading and experiment configuration of historic M8 Power Calibration experiments (M8CAL). The Monte Carlo code MCNP was utilized for steady-state analyses, and transient calculations were performed with the point kinetics code TREKIN. Thermal analyses were performed with the COMSOL multi-physics code. Using the results of this study, a new LEU Baseline design concept is being established, which will be evaluated in detail in a future report.« less

77 FR 24745 - Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Thermal...

Federal Register 2010, 2011, 2012, 2013, 2014

2012-04-25

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS), Meeting of the ACRS Subcommittee on Thermal Hydraulic Phenomena; Notice of Meeting The ACRS Subcommittee on Thermal Hydraulic Phenomena will hold a meeting on May 8-9, 2012, Room T-2B1, 11545 Rockville Pike, Rockville, Maryland. The...

Coupled Long-Term Simulation of Reach-Scale Water and Heat Fluxes Across the River-Groundwater Interface for Retrieving Hyporheic Residence Times and Temperature Dynamics

NASA Astrophysics Data System (ADS)

Munz, Matthias; Oswald, Sascha E.; Schmidt, Christian

2017-11-01

Flow patterns in conjunction with seasonal and diurnal temperature variations control ecological and biogeochemical conditions in hyporheic sediments. In particular, hyporheic temperatures have a great impact on many temperature-sensitive microbial processes. In this study, we used 3-D coupled water flow and heat transport simulations applying the HydroGeoSphere code in combination with high-resolution observations of hydraulic heads and temperatures to quantify reach-scale water and heat flux across the river-groundwater interface and hyporheic temperature dynamics of a lowland gravel bed river. The model was calibrated in order to constrain estimates of the most sensitive model parameters. The magnitude and variations of the simulated temperatures matched the observed ones, with an average mean absolute error of 0.7°C and an average Nash Sutcliffe efficiency of 0.87. Our results indicate that nonsubmerged streambed structures such as gravel bars cause substantial thermal heterogeneity within the saturated sediment at the reach scale. Individual hyporheic flow path temperatures strongly depend on the flow path residence time, flow path depth, river, and groundwater temperature. Variations in individual hyporheic flow path temperatures were up to 7.9°C, significantly higher than the daily average (2.8°C), but still lower than the average seasonal hyporheic temperature difference (19.2°C). The distribution between flow path temperatures and residence times follows a power law relationship with exponent of about 0.37. Based on this empirical relation, we further estimated the influence of hyporheic flow path residence time and temperature on oxygen consumption which was found to partly increase by up to 29% in simulations.

Conductor analysis of the ITER FEAT poloidal field coils during a plasma scenario

NASA Astrophysics Data System (ADS)

Nicollet, S.; Hertout, P.; Duchateau, J. L.; Bleyer, A.; Bessette, D.

2002-05-01

In the framework of the ITER (International Thermonuclear Experimental Reactor) FEAT (Fusion Energy Advanced Tokamak) project, a fully superconducting PF (Poloidal Field) system has been designed in detail. The Central Solenoid and the 6 equilibrium coils constituting the PF system provide the magnetic fields which develop, shape and control the 15 MA plasma during the 1800 s of a typical plasma scenario. The 6 PF coils will be wound two-in-hand from a 45 kA niobium-titanium CICC (Cable-In-Conduit-Conductor). These coils will experience severe heat loads specially during the 400 s of the plasma burn: nuclear heating due to the 400 MW of fusion power, thermal radiation and AC losses (30 to 300 kJ). The AC losses along the PF coil pancakes are deduced from accurate magnetic field computations performed with a 3D magnetostatic code, TRAPS. The nuclear heating and the thermal radiation are assumed to be uniform over a given face of the PF coils. These heat loads are used as input to perform the thermal and hydraulic analysis with a finite element code, GANDALF. The temperature increases (0.1 to 0.4 K) are computed, the margins and performances of the conductor are evaluated.

Designing for adaptation to novelty and change: functional information, emergent feature graphics, and higher-level control.

PubMed

Hajdukiewicz, John R; Vicente, Kim J

2002-01-01

Ecological interface design (EID) is a theoretical framework that aims to support worker adaptation to change and novelty in complex systems. Previous evaluations of EID have emphasized representativeness to enhance generalizability of results to operational settings. The research presented here is complementary, emphasizing experimental control to enhance theory building. Two experiments were conducted to test the impact of functional information and emergent feature graphics on adaptation to novelty and change in a thermal-hydraulic process control microworld. Presenting functional information in an interface using emergent features encouraged experienced participants to become perceptually coupled to the interface and thereby to exhibit higher-level control and more successful adaptation to unanticipated events. The absence of functional information or of emergent features generally led to lower-level control and less success at adaptation, the exception being a minority of participants who compensated by relying on analytical reasoning. These findings may have practical implications for shaping coordination in complex systems and fundamental implications for the development of a general unified theory of coordination for the technical, human, and social sciences. Actual or potential applications of this research include the design of human-computer interfaces that improve safety in complex sociotechnical systems.

Advanced numerical methods for three dimensional two-phase flow calculations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toumi, I.; Caruge, D.

1997-07-01

This paper is devoted to new numerical methods developed for both one and three dimensional two-phase flow calculations. These methods are finite volume numerical methods and are based on the use of Approximate Riemann Solvers concepts to define convective fluxes versus mean cell quantities. The first part of the paper presents the numerical method for a one dimensional hyperbolic two-fluid model including differential terms as added mass and interface pressure. This numerical solution scheme makes use of the Riemann problem solution to define backward and forward differencing to approximate spatial derivatives. The construction of this approximate Riemann solver uses anmore » extension of Roe`s method that has been successfully used to solve gas dynamic equations. As far as the two-fluid model is hyperbolic, this numerical method seems very efficient for the numerical solution of two-phase flow problems. The scheme was applied both to shock tube problems and to standard tests for two-fluid computer codes. The second part describes the numerical method in the three dimensional case. The authors discuss also some improvements performed to obtain a fully implicit solution method that provides fast running steady state calculations. Such a scheme is not implemented in a thermal-hydraulic computer code devoted to 3-D steady-state and transient computations. Some results obtained for Pressurised Water Reactors concerning upper plenum calculations and a steady state flow in the core with rod bow effect evaluation are presented. In practice these new numerical methods have proved to be stable on non staggered grids and capable of generating accurate non oscillating solutions for two-phase flow calculations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gilkey, Lindsay

This milestone presents a demonstration of the High-to-Low (Hi2Lo) process in the VVI focus area. Validation and additional calculations with the commercial computational fluid dynamics code, STAR-CCM+, were performed using a 5x5 fuel assembly with non-mixing geometry and spacer grids. This geometry was based on the benchmark experiment provided by Westinghouse. Results from the simulations were compared to existing experimental data and to the subchannel thermal-hydraulics code COBRA-TF (CTF). An uncertainty quantification (UQ) process was developed for the STAR-CCM+ model and results of the STAR UQ were communicated to CTF. Results from STAR-CCM+ simulations were used as experimental design pointsmore » in CTF to calibrate the mixing parameter β and compared to results obtained using experimental data points. This demonstrated that CTF’s β parameter can be calibrated to match existing experimental data more closely. The Hi2Lo process for the STAR-CCM+/CTF code coupling was documented in this milestone and closely linked L3:VVI.H2LP15.01 milestone report.« less

The General-Use Nodal Network Solver (GUNNS) Modeling Package for Space Vehicle Flow System Simulation

NASA Technical Reports Server (NTRS)

Harvey, Jason; Moore, Michael

2013-01-01

The General-Use Nodal Network Solver (GUNNS) is a modeling software package that combines nodal analysis and the hydraulic-electric analogy to simulate fluid, electrical, and thermal flow systems. GUNNS is developed by L-3 Communications under the TS21 (Training Systems for the 21st Century) project for NASA Johnson Space Center (JSC), primarily for use in space vehicle training simulators at JSC. It has sufficient compactness and fidelity to model the fluid, electrical, and thermal aspects of space vehicles in real-time simulations running on commodity workstations, for vehicle crew and flight controller training. It has a reusable and flexible component and system design, and a Graphical User Interface (GUI), providing capability for rapid GUI-based simulator development, ease of maintenance, and associated cost savings. GUNNS is optimized for NASA's Trick simulation environment, but can be run independently of Trick.

Neutron radiography experiments for verification of soluble boron mixing and transport modeling under natural circulation conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Morlang, G.M.

1996-06-01

The use of neutron radiography for visualization of fluid flow through flow visualization modules has been very successful. Current experiments at the Penn State Breazeale Reactor serve to verify the mixing and transport of soluble boron under natural flow conditions as would be experienced in a pressurized water reactor. Different flow geometries have been modeled including holes, slots, and baffles. Flow modules are constructed of aluminum box material 1 1/2 inches by 4 inches in varying lengths. An experimental flow system was built which pumps fluid to a head tank and natural circulation flow occurs from the head tank throughmore » the flow visualization module to be radiographed. The entire flow system is mounted on a portable assembly to allow placement of the flow visualization module in front of the neutron beam port. A neutron-transparent fluorinert fluid is used to simulate water at different densities. Boron is modeled by gadolinium oxide powder as a tracer element, which is placed in a mixing assembly and injected into the system by remote operated electric valve, once the reactor is at power. The entire sequence is recorded on real-time video. Still photographs are made frame-by-frame from the video tape. Computers are used to digitally enhance the video and still photographs. The data obtained from the enhancement will be used for verification of simple geometry predictions using the TRAC and RELAP thermal-hydraulic codes. A detailed model of a reactor vessel inlet plenum, downcomer region, flow distribution area and core inlet is being constructed to model the AP600 plenum. Successive radiography experiments of each section of the model under identical conditions will provide a complete vessel/core model for comparison with the thermal-hydraulic codes.« less

Influence of Groundwater Hydraulic Gradient on Bank Storage Metrics.

PubMed

Welch, Chani; Harrington, Glenn A; Cook, Peter G

2015-01-01

The hydraulic gradient between aquifers and rivers is one of the most variable properties in a river/aquifer system. Detailed process understanding of bank storage under hydraulic gradients is obtained from a two-dimensional numerical model of a variably saturated aquifer slice perpendicular to a river. Exchange between the river and the aquifer occurs first at the interface with the unsaturated zone. The proportion of total water exchanged through the river bank compared to the river bed is a function of aquifer hydraulic conductivity, partial penetration, and hydraulic gradient. Total exchange may be estimated to within 50% using existing analytical solutions provided that unsaturated zone processes do not strongly influence exchange. Model-calculated bank storage is at a maximum when no hydraulic gradient is present and increases as the hydraulic conductivity increases. However, in the presence of a hydraulic gradient, the largest exchange flux or distance of penetration does not necessarily correspond to the highest hydraulic conductivity, as high hydraulic conductivity increases the components of exchange both into and out of an aquifer. Flood wave characteristics do not influence ambient groundwater discharge, and so in large floods, hydraulic gradients must be high to reduce the volume of bank storage. Practical measurement of bank storage metrics is problematic due to the limitations of available measurement technologies and the nested processes of exchange that occur at the river-aquifer interface. Proxies, such as time series concentration data in rivers and groundwater, require further development to be representative and quantitative. © 2014, National GroundWater Association.

Development and preliminary verification of the 3D core neutronic code: COCO

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lu, H.; Mo, K.; Li, W.

As the recent blooming economic growth and following environmental concerns (China)) is proactively pushing forward nuclear power development and encouraging the tapping of clean energy. Under this situation, CGNPC, as one of the largest energy enterprises in China, is planning to develop its own nuclear related technology in order to support more and more nuclear plants either under construction or being operation. This paper introduces the recent progress in software development for CGNPC. The focus is placed on the physical models and preliminary verification results during the recent development of the 3D Core Neutronic Code: COCO. In the COCO code,more » the non-linear Green's function method is employed to calculate the neutron flux. In order to use the discontinuity factor, the Neumann (second kind) boundary condition is utilized in the Green's function nodal method. Additionally, the COCO code also includes the necessary physical models, e.g. single-channel thermal-hydraulic module, burnup module, pin power reconstruction module and cross-section interpolation module. The preliminary verification result shows that the COCO code is sufficient for reactor core design and analysis for pressurized water reactor (PWR). (authors)« less

Actuator digital interface unit (AIU). [control units for space shuttle data system

NASA Technical Reports Server (NTRS)

1973-01-01

Alternate versions of the actuator interface unit are presented. One alternate is a dual-failure immune configuration which feeds a look-and-switch dual-failure immune hydraulic system. The other alternate is a single-failure immune configuration which feeds a majority voting hydraulic system. Both systems communicate with the data bus through data terminals dedicated to each user subsystem. Both operational control data and configuration control information are processed in and out of the subsystem via the data terminal which yields the actuator interface subsystem, self-managing within its failure immunity capability.

Flow and Temperature Distribution Evaluation on Sodium Heated Large-sized Straight Double-wall-tube Steam Generator

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kisohara, Naoyuki; Moribe, Takeshi; Sakai, Takaaki

2006-07-01

The sodium heated steam generator (SG) being designed in the feasibility study on commercialized fast reactor cycle systems is a straight double-wall-tube type. The SG is large sized to reduce its manufacturing cost by economics of scale. This paper addresses the temperature and flow multi-dimensional distributions at steady state to obtain the prospect of the SG. Large-sized heat exchanger components are prone to have non-uniform flow and temperature distributions. These phenomena might lead to tube buckling or tube to tube-sheet junction failure in straight tube type SGs, owing to tubes thermal expansion difference. The flow adjustment devices installed in themore » SG are optimized to prevent these issues, and the temperature distribution properties are uncovered by analysis methods. The analysis model of the SG consists of two parts, a sodium inlet distribution plenum (the plenum) and a heat transfer tubes bundle region (the bundle). The flow and temperature distributions in the plenum and the bundle are evaluated by the three-dimensional code 'FLUENT' and the two dimensional thermal-hydraulic code 'MSG', respectively. The MSG code is particularly developed for sodium heated SGs in JAEA. These codes have revealed that the sodium flow is distributed uniformly by the flow adjustment devices, and that the lateral tube temperature distributions remain within the allowable temperature range for the structural integrity of the tubes and the tube to tube-sheet junctions. (authors)« less

System Simulation of Nuclear Power Plant by Coupling RELAP5 and Matlab/Simulink

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meng Lin; Dong Hou; Zhihong Xu

2006-07-01

Since RELAP5 code has general and advanced features in thermal-hydraulic computation, it has been widely used in transient and accident safety analysis, experiment planning analysis, and system simulation, etc. So we wish to design, analyze, verify a new Instrumentation And Control (I and C) system of Nuclear Power Plant (NPP) based on the best-estimated code, and even develop our engineering simulator. But because of limited function of simulating control and protection system in RELAP5, it is necessary to expand the function for high efficient, accurate, flexible design and simulation of I and C system. Matlab/Simulink, a scientific computation software, justmore » can compensate the limitation, which is a powerful tool in research and simulation of plant process control. The software is selected as I and C part to be coupled with RELAP5 code to realize system simulation of NPPs. There are two key techniques to be solved. One is the dynamic data exchange, by which Matlab/Simulink receives plant parameters and returns control results. Database is used to communicate the two codes. Accordingly, Dynamic Link Library (DLL) is applied to link database in RELAP5, while DLL and S-Function is applied in Matlab/Simulink. The other problem is synchronization between the two codes for ensuring consistency in global simulation time. Because Matlab/Simulink always computes faster than RELAP5, the simulation time is sent by RELAP5 and received by Matlab/Simulink. A time control subroutine is added into the simulation procedure of Matlab/Simulink to control its simulation advancement. Through these ways, Matlab/Simulink is dynamically coupled with RELAP5. Thus, in Matlab/Simulink, we can freely design control and protection logic of NPPs and test it with best-estimated plant model feedback. A test will be shown to illuminate that results of coupling calculation are nearly the same with one of single RELAP5 with control logic. In practice, a real Pressurized Water Reactor (PWR) is modeled by RELAP5 code, and its main control and protection system is duplicated by Matlab/Simulink. Some steady states and transients are calculated under control of these I and C systems, and the results are compared with the plant test curves. The application showed that it can do exact system simulation of NPPs by coupling RELAP5 and Matlab/Simulink. This paper will mainly focus on the coupling method, plant thermal-hydraulic model, main control logics, test and application results. (authors)« less

BISON Modeling of Reactivity-Initiated Accident Experiments in a Static Environment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Folsom, Charles P.; Jensen, Colby B.; Williamson, Richard L.

2016-09-01

In conjunction with the restart of the TREAT reactor and the design of test vehicles, modeling and simulation efforts are being used to model the response of Accident Tolerant Fuel (ATF) concepts under reactivity insertion accident (RIA) conditions. The purpose of this work is to model a baseline case of a 10 cm long UO2-Zircaloy fuel rodlet using BISON and RELAP5 over a range of energy depositions and with varying reactor power pulse widths. The results show the effect of varying the pulse width and energy deposition on both thermal and mechanical parameters that are important for predicting failure ofmore » the fuel rodlet. The combined BISON/RELAP5 model captures coupled thermal and mechanical effects on the fuel-to-cladding gap conductance, cladding-to-coolant heat transfer coefficient and water temperature and pressure that would not be capable in each code individually. These combined effects allow for a more accurate modeling of the thermal and mechanical response in the fuel rodlet and thermal-hydraulics of the test vehicle.« less

Preliminary Thermal Modeling of HI-Storm 100S-218 Version B Storage Modules at Hope Creek Cuclear Power Station ISFSI

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cuta, Judith M.; Adkins, Harold E.

2013-08-30

As part of the Used Fuel Disposition Campaign of the U. S. Department of Energy, Office of Nuclear Energy (DOE-NE) Fuel Cycle Research and Development, a consortium of national laboratories and industry is performing visual inspections and temperature measurements of selected storage modules at various locations around the United States. This report documents thermal analyses in in support of the inspections at the Hope Creek Nuclear Generating Station ISFSI. This site utilizes the HI-STORM100 vertical storage system developed by Holtec International. This is a vertical storage module design, and the thermal models are being developed using COBRA-SFS (Michener, et al.,more » 1987), a code developed by PNNL for thermal-hydraulic analyses of multi assembly spent fuel storage and transportation systems. This report describes the COBRA-SFS model in detail, and presents pre-inspection predictions of component temperatures and temperature distributions. The final report will include evaluation of inspection results, and if required, additional post-test calculations, with appropriate discussion of results.« less

ENVIRONMENTAL HYDRAULICS

EPA Science Inventory

The thermal, chemical, and biological quality of water in rivers, lakes, reservoirs, and near coastal areas is inseparable from a consideration of hydraulic engineering principles: therefore, the term environmental hydraulics. In this chapter we discuss the basic principles of w...

Ocean thermal gradient hydraulic power plant.

PubMed

Beck, E J

1975-07-25

Solar energy stored in the oceans may be used to generate power by exploiting ploiting thermal gradients. A proposed open-cycle system uses low-pressure steam to elevate vate water, which is then run through a hydraulic turbine to generate power. The device is analogous to an air lift pump.

Thermal APU/hydraulics analysis program. User's guide and programmer's manual

NASA Technical Reports Server (NTRS)

Deluna, T. A.

1976-01-01

The User's Guide information plus program description necessary to run and have a general understanding of the Thermal APU/Hydraulics Analysis Program (TAHAP) is described. This information consists of general descriptions of the APU/hydraulic system and the TAHAP model, input and output data descriptions, and specific subroutine requirements. Deck setups and input data formats are included and other necessary and/or helpful information for using TAHAP is given. The math model descriptions for the driver program and each of its supporting subroutines are outlined.

IAC - INTEGRATED ANALYSIS CAPABILITY

NASA Technical Reports Server (NTRS)

Frisch, H. P.

1994-01-01

The objective of the Integrated Analysis Capability (IAC) system is to provide a highly effective, interactive analysis tool for the integrated design of large structures. With the goal of supporting the unique needs of engineering analysis groups concerned with interdisciplinary problems, IAC was developed to interface programs from the fields of structures, thermodynamics, controls, and system dynamics with an executive system and database to yield a highly efficient multi-disciplinary system. Special attention is given to user requirements such as data handling and on-line assistance with operational features, and the ability to add new modules of the user's choice at a future date. IAC contains an executive system, a data base, general utilities, interfaces to various engineering programs, and a framework for building interfaces to other programs. IAC has shown itself to be effective in automatic data transfer among analysis programs. IAC 2.5, designed to be compatible as far as possible with Level 1.5, contains a major upgrade in executive and database management system capabilities, and includes interfaces to enable thermal, structures, optics, and control interaction dynamics analysis. The IAC system architecture is modular in design. 1) The executive module contains an input command processor, an extensive data management system, and driver code to execute the application modules. 2) Technical modules provide standalone computational capability as well as support for various solution paths or coupled analyses. 3) Graphics and model generation interfaces are supplied for building and viewing models. Advanced graphics capabilities are provided within particular analysis modules such as INCA and NASTRAN. 4) Interface modules provide for the required data flow between IAC and other modules. 5) User modules can be arbitrary executable programs or JCL procedures with no pre-defined relationship to IAC. 6) Special purpose modules are included, such as MIMIC (Model Integration via Mesh Interpolation Coefficients), which transforms field values from one model to another; LINK, which simplifies incorporation of user specific modules into IAC modules; and DATAPAC, the National Bureau of Standards statistical analysis package. The IAC database contains structured files which provide a common basis for communication between modules and the executive system, and can contain unstructured files such as NASTRAN checkpoint files, DISCOS plot files, object code, etc. The user can define groups of data and relations between them. A full data manipulation and query system operates with the database. The current interface modules comprise five groups: 1) Structural analysis - IAC contains a NASTRAN interface for standalone analysis or certain structural/control/thermal combinations. IAC provides enhanced structural capabilities for normal modes and static deformation analysis via special DMAP sequences. IAC 2.5 contains several specialized interfaces from NASTRAN in support of multidisciplinary analysis. 2) Thermal analysis - IAC supports finite element and finite difference techniques for steady state or transient analysis. There are interfaces for the NASTRAN thermal analyzer, SINDA/SINFLO, and TRASYS II. FEMNET, which converts finite element structural analysis models to finite difference thermal analysis models, is also interfaced with the IAC database. 3) System dynamics - The DISCOS simulation program which allows for either nonlinear time domain analysis or linear frequency domain analysis, is fully interfaced to the IAC database management capability. 4) Control analysis - Interfaces for the ORACLS, SAMSAN, NBOD2, and INCA programs allow a wide range of control system analyses and synthesis techniques. Level 2.5 includes EIGEN, which provides tools for large order system eigenanalysis, and BOPACE, which allows for geometric capabilities and finite element analysis with nonlinear material. Also included in IAC level 2.5 is SAMSAN 3.1, an engineering analysis program which contains a general purpose library of over 600 subroutin

Discrete fracture modeling of multiphase flow and hydrocarbon production in fractured shale or low permeability reservoirs

NASA Astrophysics Data System (ADS)

Hao, Y.; Settgast, R. R.; Fu, P.; Tompson, A. F. B.; Morris, J.; Ryerson, F. J.

2016-12-01

It has long been recognized that multiphase flow and transport in fractured porous media is very important for various subsurface applications. Hydrocarbon fluid flow and production from hydraulically fractured shale reservoirs is an important and complicated example of multiphase flow in fractured formations. The combination of horizontal drilling and hydraulic fracturing is able to create extensive fracture networks in low permeability shale rocks, leading to increased formation permeability and enhanced hydrocarbon production. However, unconventional wells experience a much faster production decline than conventional hydrocarbon recovery. Maintaining sustainable and economically viable shale gas/oil production requires additional wells and re-fracturing. Excessive fracturing fluid loss during hydraulic fracturing operations may also drive up operation costs and raise potential environmental concerns. Understanding and modeling processes that contribute to decreasing productivity and fracturing fluid loss represent a critical component for unconventional hydrocarbon recovery analysis. Towards this effort we develop a discrete fracture model (DFM) in GEOS (LLNL multi-physics computational code) to simulate multiphase flow and transfer in hydraulically fractured reservoirs. The DFM model is able to explicitly account for both individual fractures and their surrounding rocks, therefore allowing for an accurate prediction of impacts of fracture-matrix interactions on hydrocarbon production. We apply the DFM model to simulate three-phase (water, oil, and gas) flow behaviors in fractured shale rocks as a result of different hydraulic stimulation scenarios. Numerical results show that multiphase flow behaviors at the fracture-matrix interface play a major role in controlling both hydrocarbon production and fracturing fluid recovery rates. The DFM model developed in this study will be coupled with the existing hydro-fracture model to provide a fully integrated geomechanical and reservoir simulation capability for an accurate prediction and assessment of hydrocarbon production and hydraulic fracturing performance. This work was performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ling Zou; Hongbin Zhang; Jess Gehin

A coupled TH/Neutronics/CRUD framework, which is able to simulate the CRUD deposits impact on CIPS phenomenon, was described in this paper. This framework includes the coupling among three essential physics, thermal-hydraulics, CRUD and neutronics. The overall framework was implemented by using the CFD software STAR-CCM+, developing CRUD codes, and using the neutronics code DeCART. The coupling was implemented by exchanging data between softwares using intermediate exchange files. A typical 3 by 3 PWR fuel pin problem was solved under this framework. The problem was solved in a 12 months length period of time. Time-dependent solutions were provided, including CRUD depositsmore » inventory and their distributions on fuels, boron hideout amount inside CRUD deposits, as well as power shape changing over time. The results clearly showed the power shape suppression in regions where CRUD deposits exist, which is a strong indication of CIPS phenomenon.« less

Development of a New 47-Group Library for the CASL Neutronics Simulators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Kang Seog; Williams, Mark L; Wiarda, Dorothea

The CASL core simulator MPACT is under development for the neutronics and thermal-hydraulics coupled simulation for the pressurized light water reactors. The key characteristics of the MPACT code include a subgroup method for resonance self-shielding, and a whole core solver with a 1D/2D synthesis method. The ORNL AMPX/SCALE code packages have been significantly improved to support various intermediate resonance self-shielding approximations such as the subgroup and embedded self-shielding methods. New 47-group AMPX and MPACT libraries based on ENDF/B-VII.0 have been generated for the CASL core simulator MPACT of which group structure comes from the HELIOS library. The new 47-group MPACTmore » library includes all nuclear data required for static and transient core simulations. This study discusses a detailed procedure to generate the 47-group AMPX and MPACT libraries and benchmark results for the VERA progression problems.« less

Coolant mixing in LMFBR rod bundles and outlet plenum mixing transients. Final report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Todreas, N.E.; Cheng, S.K.; Basehore, K.

1984-08-01

This project principally undertook the investigation of the thermal hydraulic performance of wire wrapped fuel bundles of LMFBR configuration. Results obtained included phenomenological models for friction factors, flow split and mixing characteristics; correlations for predicting these characteristics suitable for insertion in design codes; numerical codes for analyzing bundle behavior both of the lumped subchannel and distributed parameter categories and experimental techniques for pressure velocity, flow split, salt conductivity and temperature measurement in water cooled mockups of bundles and subchannels. Flow regimes investigated included laminar, transition and turbulent flow under forced convection and mixed convection conditions. Forced convections conditions were emphasized.more » Continuing efforts are underway at MIT to complete the investigation of the mixed convection regime initiated here. A number of investigations on outlet plenum behavior were also made. The reports of these investigations are identified.« less

NASA Lewis Steady-State Heat Pipe Code Architecture

NASA Technical Reports Server (NTRS)

Mi, Ye; Tower, Leonard K.

2013-01-01

NASA Glenn Research Center (GRC) has developed the LERCHP code. The PC-based LERCHP code can be used to predict the steady-state performance of heat pipes, including the determination of operating temperature and operating limits which might be encountered under specified conditions. The code contains a vapor flow algorithm which incorporates vapor compressibility and axially varying heat input. For the liquid flow in the wick, Darcy s formula is employed. Thermal boundary conditions and geometric structures can be defined through an interactive input interface. A variety of fluid and material options as well as user defined options can be chosen for the working fluid, wick, and pipe materials. This report documents the current effort at GRC to update the LERCHP code for operating in a Microsoft Windows (Microsoft Corporation) environment. A detailed analysis of the model is presented. The programming architecture for the numerical calculations is explained and flowcharts of the key subroutines are given

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

DOE Office of Scientific and Technical Information (OSTI.GOV)

Khodak, A.; Zhai, Y.; Wang, W.

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of themore » coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.« less

Parametric Thermal and Flow Analysis of ITER Diagnostic Shield Module

DOE PAGES

Khodak, A.; Zhai, Y.; Wang, W.; ...

2017-06-19

As part of the diagnostic port plug assembly, the ITER Diagnostic Shield Module (DSM) is designed to provide mechanical support and the plasma shielding while allowing access to plasma diagnostics. Thermal and hydraulic analysis of the DSM was performed using a conjugate heat transfer approach, in which heat transfer was resolved in both solid and liquid parts, and simultaneously, fluid dynamics analysis was performed only in the liquid part. ITER Diagnostic First Wall (DFW) and cooling tubing were also included in the analysis. This allowed direct modeling of the interface between DSM and DFW, and also direct assessment of themore » coolant flow distribution between the parts of DSM and DFW to ensure DSM design meets the DFW cooling requirements. Design of the DSM included voids filled with Boron Carbide pellets, allowing weight reduction while keeping shielding capability of the DSM. These voids were modeled as a continuous solid with smeared material properties using analytical relation for thermal conductivity. Results of the analysis lead to design modifications improving heat transfer efficiency of the DSM. Furthermore, the effect of design modifications on thermal performance as well as effect of Boron Carbide will be presented.« less

Experiment on large scale plume interaction with a stratified gas environment resembling the thermal activity of a autocatalytic recombiner

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mignot, G.; Kapulla, R.; Paladino, D.

Computational Fluid Dynamics codes (CFD) are increasingly being used to simulate containment conditions after various transient accident scenarios. Consequently, the reliability of such codes must be tested against experimental data. Such validation experiments related to gas mixing and hydrogen transport within containment compartments addressing the effect of heat source are presented in this paper. The experiments were conducted in the large-scale thermal-hydraulics PANDA facility located at the Paul-Scherrer-Inst. (PSI) in Switzerland, in the frame of the OECD/SETH-2 project. A 10 kW electric heater simulating the thermal activity of the autocatalytic recombiner was activated at full power in a containment vesselmore » at the top of which a thick helium layer is initially present. The hot plume interacts with the bottom of the helium layer which is slowly eroded until complete break up at 1350 s. After final erosion of the layer a strong temperature and concentration gradient is maintained in the vessel below the heater inlet as well as in the adjacent vessel below the interconnecting pipe. A detailed characterization of the operating heater suggests the presence of cold gas ingress at the outlet that affects the flow in the chimney. This can be of concern if present in a real PAR unit. (authors)« less

Modeling and Simulation of the ITER First Wall/Blanket Primary Heat Transfer System

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ying, Alice; Popov, Emilian L

2011-01-01

ITER inductive power operation is modeled and simulated using a thermal-hydraulics system code (RELAP5) integrated with a 3-D CFD (SC-Tetra) code. The Primary Heat Transfer System (PHTS) functions are predicted together with the main parameters operational ranges. The control algorithm strategy and derivation are summarized as well. The First Wall and Blanket modules are the primary components of PHTS, used to remove the major part of the thermal heat from the plasma. The modules represent a set of flow channels in solid metal structure that serve to absorb the radiation heat and nuclear heating from the fusion reactions and tomore » provide shield for the vacuum vessel. The blanket modules are water cooled. The cooling is forced convective with constant blanket inlet temperature and mass flow rate. Three independent water loops supply coolant to the three blanket sectors. The main equipment of each loop consists of a pump, a steam pressurizer and a heat exchanger. A major feature of ITER is the pulsed operation. The plasma does not burn continuously, but on intervals with large periods of no power between them. This specific feature causes design challenges to accommodate the thermal expansion of the coolant during the pulse period and requires active temperature control to maintain a constant blanket inlet temperature.« less

Pumping Test Determination of Unsaturated Aquifer Properties

NASA Astrophysics Data System (ADS)

Mishra, P. K.; Neuman, S. P.

2008-12-01

Tartakovsky and Neuman [2007] presented a new analytical solution for flow to a partially penetrating well pumping at a constant rate from a compressible unconfined aquifer considering the unsaturated zone. In their solution three-dimensional, axially symmetric unsaturated flow is described by a linearized version of Richards' equation in which both hydraulic conductivity and water content vary exponentially with incremental capillary pressure head relative to its air entry value, the latter defining the interface between the saturated and unsaturated zones. Both exponential functions are characterized by a common exponent k having the dimension of inverse length, or equivalently a dimensionless exponent kd=kb where b is initial saturated thickness. The authors used their solution to analyze drawdown data from a pumping test conducted by Moench et al. [2001] in a Glacial Outwash Deposit at Cape Cod, Massachusetts. Their analysis yielded estimates of horizontal and vertical saturated hydraulic conductivities, specific storage, specific yield and k . Recognizing that hydraulic conductivity and water content seldom vary identically with incremental capillary pressure head, as assumed by Tartakovsky and Neuman [2007], we note that k is at best an effective rather than a directly measurable soil parameter. We therefore ask to what extent does interpretation of a pumping test based on the Tartakovsky-Neuman solution allow estimating aquifer unsaturated parameters as described by more common constitutive water retention and relative hydraulic conductivity models such as those of Brooks and Corey [1964] or van Genuchten [1980] and Mualem [1976a]? We address this question by showing how may be used to estimate the capillary air entry pressure head k and the parameters of such constitutive models directly, without a need for inverse unsaturated numerical simulations of the kind described by Moench [2003]. To assess the validity of such direct estimates we use maximum likelihood- based model selection criteria to compare the abilities of numerical models based on the STOMP code to reproduce observed drawdowns during the test when saturated and unsaturated aquifer parameters are estimated either in the above manner or by means of the inverse code PEST.

78 FR 8202 - Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy...

Federal Register 2010, 2011, 2012, 2013, 2014

2013-02-05

... NUCLEAR REGULATORY COMMISSION Advisory Committee on Reactor Safeguards (ACRS) Meeting of the Joint ACRS Subcommittees on Thermal Hydraulic Phenomena and Materials, Metallurgy and Reactor Fuels; Notice... Reactor Fuels will hold a meeting on February 20, 2013, Room T-2B1, 11545 Rockville Pike, Rockville...

75 FR 80544 - NUREG-1953, Confirmatory Thermal-Hydraulic Analysis To Support Specific Success Criteria in the...

Federal Register 2010, 2011, 2012, 2013, 2014

2010-12-22

... To Support Specific Success Criteria in the Standardized Plant Analysis Risk Models--Surry and Peach... Specific Success Criteria in the Standardized Plant Analysis Risk Models--Surry and Peach Bottom, Draft..., ``Confirmatory Thermal-Hydraulic Analysis to Support Specific Success Criteria in the Standardized Plant Analysis...

VERA 3.6 Release Notes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Williamson, Richard L.; Kochunas, Brendan; Adams, Brian M.

The Virtual Environment for Reactor Applications components included in this distribution include selected computational tools and supporting infrastructure that solve neutronics, thermal-hydraulics, fuel performance, and coupled neutronics-thermal hydraulics problems. The infrastructure components provide a simplified common user input capability and provide for the physics integration with data transfer and coupled-physics iterative solution algorithms.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Epiney, A.; Canepa, S.; Zerkak, O.

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

Preliminary Analysis of a Fully Solid State Magnetocaloric Refrigeration

DOE Office of Scientific and Technical Information (OSTI.GOV)

Abdelaziz, Omar

Magnetocaloric refrigeration is an alternative refrigeration technology with significant potential energy savings compared to conventional vapor compression refrigeration technology. Most of the reported active magnetic regenerator (AMR) systems that operate based on the magnetocaloric effect use heat transfer fluid to exchange heat, which results in complicated mechanical subsystems and components such as rotating valves and hydraulic pumps. In this paper, we propose an alternative mechanism for heat transfer between the AMR and the heat source/sink. High-conductivity moving rods/sheets (e.g. copper, brass, iron, graphite, aluminum or composite structures from these) are utilized instead of heat transfer fluid significantly enhancing the heatmore » transfer rate hence cooling/heating capacity. A one-dimensional model is developed to study the solid state AMR. In this model, the heat exchange between the solid-solid interfaces is modeled via a contact conductance, which depends on the interface apparent pressure, material hardness, thermal conductivity, surface roughness, surface slope between the interfaces, and material filled in the gap between the interfaces. Due to the tremendous impact of the heat exchange on the AMR cycle performance, a sensitivity analysis is conducted employing a response surface method, in which the apparent pressure, effective surface roughness and grease thermal conductivity are the uncertainty factors. COP and refrigeration capacity are presented as the response in the sensitivity analysis to reveal the important factors influencing the fully solid state AMR and optimize the solid state AMR efficiency. The performances of fully solid state AMR and traditional AMR are also compared and discussed in present work. The results of this study will provide general guidelines for designing high performance solid state AMR systems.« less

Plasma kinetic effects on atomistic mix in one dimension and at structured interfaces (I)

NASA Astrophysics Data System (ADS)

Yin, L.; Albright, B. J.; Vold, E. L.; Taitano, W.; Chacon, L.; Simakov, A.

2017-10-01

Kinetic effects on interfacial mix are examined using VPIC simulations. In 1D, comparisons are made to the results of analytic theory in the small Knudsen number limit. While the bulk mixing properties of interfaces are in general agreement, differences arise near the low-concentration fronts during the early evolution of a sharp interface when the species' perpendicular scattering rate dominates over the slowing down rate. In kinetic simulations, the diffusion velocities can be larger or comparable to the ion thermal speeds, and the Knudsen number can be large. Super-diffusive growth in mix widths (Δx ta where a >=1/2) is seen before transition to the slow diffusive process predicted from theory (a =1/2). Mixing at interfaces leads to persistent, bulk, hydrodynamic features in the center of mass flow profiles as a result of diffusion and momentum conservation. These conclusions are drawn from VPIC results together with simulations from the RAGE hydrodynamics code with an implementation of diffusion and viscosity from theory and an implicit Vlasov-Fokker-Planck code iFP. In perturbed 2D and 3D interfaces, it is found that 1D ambipolarity is still valid and that initial perturbations flatten out on a-few-ps time scale, implying that finite diffusivity and viscosity can slow instability growth in ICF and HED settings. Work supported by the LANL ASC and Science programs.

CTF Theory Manual

DOE Office of Scientific and Technical Information (OSTI.GOV)

Avramova, Maria N.; Salko, Robert K.

Coolant-Boiling in Rod Arrays|Two Fluids (COBRA-TF) is a thermal/ hydraulic (T/H) simulation code designed for light water reactor (LWR) vessel analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of 9 conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and had been used and modified by several institutions over the last few decades. COBRA-TF also found use at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG) and has been improved, updated, andmore » subsequently re-branded as CTF. As part of the improvement process, it was necessary to generate sufficient documentation for the open-source code which had lacked such material upon being adopted by RDFMG. This document serves mainly as a theory manual for CTF, detailing the many two-phase heat transfer, drag, and important accident scenario models contained in the code as well as the numerical solution process utilized. Coding of the models is also discussed, all with consideration for updates that have been made when transitioning from COBRA-TF to CTF. Further documentation outside of this manual is also available at RDFMG which focus on code input deck generation and source code global variable and module listings.« less

Hydraulic characterization of aquifers by thermal response testing

NASA Astrophysics Data System (ADS)

Wagner, Valentin; Blum, Philipp; Bayer, Peter

2014-05-01

Temperature as a major physical quantity of the subsurface, and naturally occurring thermal anomalies are recognized as promising passive tracers to characterize the subsurface. Accelerated by the increasing popularity of geothermal energy, also active thermal field experiments have gained interest in hydrogeology. Such experiments involve artificial local ground heating or cooling. Among these, the thermal response test (TRT) is one of the most established field investigation techniques in shallow geothermal applications. It is a common method to investigate important subsurface heat transport parameters to design sustainable ground-source heat pump (GSHP) systems. During the test, the borehole heat exchanger (BHE) is heated up with a defined amount of energy by circulating a heat carrier fluid. By comparing temperature change between BHE inlet and outlet, the ability of the BHE to transfer heat or cold to the ambient ground is assessed. However, standard interpretation does not provide any insight into the governing processes of in-situ heat transfer. We utilize a groundwater advection sensitive TRT evaluation approach based on the analytical moving line source equation. It is shown that the TRT as a classical geothermal field test can also be used as a hydrogeological field test. Our approach benefits from the fact that thermal properties, such as thermal conductivity, of natural aquifers typically are much less variable than hydraulic properties, such as hydraulic conductivity. It is possible to determine a relatively small hydraulic conductivity range with our TRT evaluation approach, given realistic ranges for thermal conductivity, volumetric heat capacity, thermal dispersivity and thermal borehole resistance. The method is successfully tested on a large-scale geothermal laboratory experiment (9 m × 6 m × 4.5 m) and with a commercially performed TRT in the field scale. The laboratory experiment consists of a layered artificial aquifer, which is penetrated by a short BHE. This BHE is used to record a groundwater influenced TRT dataset. The performed field TRT is measured at a BHE located in the Upper Rhine Valley in South-West Germany, which penetrates a 68 m thick gravel aquifer with significant horizontal groundwater flow. At both sites, the derived hydraulic conductivity ranges obtained from TRT evaluation are shown to be within the ranges obtained from classical hydrogeological methods such as sieve analysis and pumping tests. This confirms that the temperature signal recorded during thermal response tests can be employed as a thermal tracer and that the evaluation of such a signal can be applied to estimate aquifer hydraulic conductivities.

Hydraulic manipulator design, analysis, and control at Oak Ridge National Laboratory

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kress, R.L.; Jansen, J.F.; Love, L.J.

1996-09-01

To meet the increased payload capacities demanded by present-day tasks, manipulator designers have turned to hydraulics as a means of actuation. Hydraulics have always been the actuator of choice when designing heavy-life construction and mining equipment such as bulldozers, backhoes, and tunneling devices. In order to successfully design, build, and deploy a new hydraulic manipulator (or subsystem) sophisticated modeling, analysis, and control experiments are usually needed. To support the development and deployment of new hydraulic manipulators Oak Ridge National Laboratory (ORNL) has outfitted a significant experimental laboratory and has developed the software capability for research into hydraulic manipulators, hydraulic actuators,more » hydraulic systems, modeling of hydraulic systems, and hydraulic controls. The hydraulics laboratory at ORNL has three different manipulators. First is a 6-Degree-of-Freedom (6-DoF), multi-planer, teleoperated, flexible controls test bed used for the development of waste tank clean-up manipulator controls, thermal studies, system characterization, and manipulator tracking. Finally, is a human amplifier test bed used for the development of an entire new class of teleoperated systems. To compliment the hardware in the hydraulics laboratory, ORNL has developed a hydraulics simulation capability including a custom package to model the hydraulic systems and manipulators for performance studies and control development. This paper outlines the history of hydraulic manipulator developments at ORNL, describes the hydraulics laboratory, discusses the use of the equipment within the laboratory, and presents some of the initial results from experiments and modeling associated with these hydraulic manipulators. Included are some of the results from the development of the human amplifier/de-amplifier concepts, the characterization of the thermal sensitivity of hydraulic systems, and end-point tracking accuracy studies. Experimental and analytical results are included.« less

Decay-ratio calculation in the frequency domain with the LAPUR code using 1D-kinetics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Munoz-Cobo, J. L.; Escriva, A.; Garcia, C.

This paper deals with the problem of computing the Decay Ratio in the frequency domain codes as the LAPUR code. First, it is explained how to calculate the feedback reactivity in the frequency domain using slab-geometry i.e. 1D kinetics, also we show how to perform the coupling of the 1D kinetics with the thermal-hydraulic part of the LAPUR code in order to obtain the reactivity feedback coefficients for the different channels. In addition, we show how to obtain the reactivity variation in the complex domain by solving the eigenvalue equation in the frequency domain and we compare this result withmore » the reactivity variation obtained in first order perturbation theory using the 1D neutron fluxes of the base case. Because LAPUR works in the linear regime, it is assumed that in general the perturbations are small. There is also a section devoted to the reactivity weighting factors used to couple the reactivity contribution from the different channels to the reactivity of the entire reactor core in point kinetics and 1D kinetics. Finally we analyze the effects of the different approaches on the DR value. (authors)« less

Post-test analysis of PIPER-ONE PO-IC-2 experiment by RELAP5/MOD3 codes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bovalini, R.; D`Auria, F.; Galassi, G.M.

1996-11-01

RELAP5/MOD3.1 was applied to the PO-IC-2 experiment performed in PIPER-ONE facility, which has been modified to reproduce typical isolation condenser thermal-hydraulic conditions. RELAP5 is a well known code widely used at the University of Pisa during the past seven years. RELAP5/MOD3.1 was the latest version of the code made available by the Idaho National Engineering Laboratory at the time of the reported study. PIPER-ONE is an experimental facility simulating a General Electric BWR-6 with volume and height scaling ratios of 1/2,200 and 1./1, respectively. In the frame of the present activity a once-through heat exchanger immersed in a pool ofmore » ambient temperature water, installed approximately 10 m above the core, was utilized to reproduce qualitatively the phenomenologies expected for the Isolation Condenser in the simplified BWR (SBWR). The PO-IC-2 experiment is the flood up of the PO-SD-8 and has been designed to solve some of the problems encountered in the analysis of the PO-SD-8 experiment. A very wide analysis is presented hereafter including the use of different code versions.« less

Simulation of Containment Atmosphere Mixing and Stratification Experiment in the ThAI Facility with a CFD Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babic, Miroslav; Kljenak, Ivo; Mavko, Borut

2006-07-01

The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. The main purpose of the proposed work was to assess the capabilities of the CFD code to reproduce the atmosphere structure with a three-dimensional model, coupled with condensation models proposed by the authors. A three-dimensional modelmore » of the ThAI vessel for the CFX4.4 code was developed. The flow in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also included. Calculated time-dependent variables together with temperature and volume fraction distributions at the end of different experiment phases are compared to experimental results. (authors)« less

Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

NASA Technical Reports Server (NTRS)

Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

2007-01-01

A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

Influence of spatial variability of hydraulic characteristics of soils on surface parameters obtained from remote sensing data in infrared and microwaves

NASA Technical Reports Server (NTRS)

Brunet, Y.; Vauclin, M.

1985-01-01

The correct interpretation of thermal and hydraulic soil parameters infrared from remotely sensed data (thermal infrared, microwaves) implies a good understanding of the causes of their temporal and spatial variability. Given this necessity, the sensitivity of the surface variables (temperature, moisture) to the spatial variability of hydraulic soil properties is tested with a numerical model of heat and mass transfer between bare soil and atmosphere. The spatial variability of hydraulic soil properties is taken into account in terms of the scaling factor. For a given soil, the knowledge of its frequency distribution allows a stochastic use of the model. The results are treated statistically, and the part of the variability of soil surface parameters due to that of soil hydraulic properties is evaluated quantitatively.

Parametric Analysis of a Turbine Trip Event in a BWR Using a 3D Nodal Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorzel, A.

2006-07-01

Two essential thermal hydraulics safety criteria concerning the reactor core are that even during operational transients there is no fuel melting and not-permissible cladding temperatures are avoided. A common concept for boiling water reactors is to establish a minimum critical power ratio (MCPR) for steady state operation. For this MCPR it is shown that only a very small number of fuel rods suffers a short-term dryout during the transient. It is known from experience that the limiting transient for the determination of the MCPR is the turbine trip with blocked bypass system. This fast transient was simulated for a Germanmore » BWR by use of the three-dimensional reactor analysis transient code SIMULATE-3K. The transient behaviour of the hot channels was used as input for the dryout calculation with the transient thermal hydraulics code FRANCESCA. By this way the maximum reduction of the CPR during the transient could be calculated. The fast increase in reactor power due to the pressure increase and to an increased core inlet flow is limited mainly by the Doppler effect, but automatically triggered operational measures also can contribute to the mitigation of the turbine trip. One very important method is the short-term fast reduction of the recirculation pump speed which is initiated e. g. by a pressure increase in front of the turbine. The large impacts of the starting time and of the rate of the pump speed reduction on the power progression and hence on the deterioration of CPR is presented. Another important procedure to limit the effects of the transient is the fast shutdown of the reactor that is caused when the reactor power reaches the limit value. It is shown that the SCRAM is not fast enough to reduce the first power maximum, but is able to prevent the appearance of a second - much smaller - maximum that would occur around one second after the first one in the absence of a SCRAM. (author)« less

The Design of PSB-VVER Experiments Relevant to Accident Management

NASA Astrophysics Data System (ADS)

Nevo, Alessandro Del; D'Auria, Francesco; Mazzini, Marino; Bykov, Michael; Elkin, Ilya V.; Suslov, Alexander

Experimental programs carried-out in integral test facilities are relevant for validating the best estimate thermal-hydraulic codes(1), which are used for accident analyses, design of accident management procedures, licensing of nuclear power plants, etc. The validation process, in fact, is based on well designed experiments. It consists in the comparison of the measured and calculated parameters and the determination whether a computer code has an adequate capability in predicting the major phenomena expected to occur in the course of transient and/or accidents. University of Pisa was responsible of the numerical design of the 12 experiments executed in PSB-VVER facility (2), operated at Electrogorsk Research and Engineering Center (Russia), in the framework of the TACIS 2.03/97 Contract 3.03.03 Part A, EC financed (3). The paper describes the methodology adopted at University of Pisa, starting form the scenarios foreseen in the final test matrix until the execution of the experiments. This process considers three key topics: a) the scaling issue and the simulation, with unavoidable distortions, of the expected performance of the reference nuclear power plants; b) the code assessment process involving the identification of phenomena challenging the code models; c) the features of the concerned integral test facility (scaling limitations, control logics, data acquisition system, instrumentation, etc.). The activities performed in this respect are discussed, and emphasis is also given to the relevance of the thermal losses to the environment. This issue affects particularly the small scaled facilities and has relevance on the scaling approach related to the power and volume of the facility.

Warthog: Coupling Status Update

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hart, Shane W. D.; Reardon, Bradley T.

The Warthog code was developed to couple codes that are developed in both the Multi-Physics Object-Oriented Simulation Environment (MOOSE) from Idaho National Laboratory (INL) and SHARP from Argonne National Laboratory (ANL). The initial phase of this work, focused on coupling the neutronics code PROTEUS with the fuel performance code BISON. The main technical challenge involves mapping the power density solution determined by PROTEUS to the fuel in BISON. This presents a challenge since PROTEUS uses the MOAB mesh format, but BISON, like all other MOOSE codes, uses the libMesh format. When coupling the different codes, one must consider that Warthogmore » is a light-weight MOOSE-based program that uses the Data Transfer Kit (DTK) to transfer data between the various mesh types. Users set up inputs for the codes they want to run, and then Warthog transfers the data between them. Currently Warthog supports XSProc from SCALE or the Sub-Group Application Programming Interface (SGAPI) in PROTEUS for generating cross sections. It supports arbitrary geometries using PROTEUS and BISON. DTK will transfer power densities and temperatures between the codes where the domains overlap. In the past fiscal year (FY), much work has gone into demonstrating two-way coupling for simple pin cells of various materials. XSProc was used to calculate the cross sections, which were then passed to PROTEUS in an external file. PROTEUS calculates the fission/power density, and Warthog uses DTK to pass this information to BISON, where it is used as the heat source. BISON then calculates the temperature profile of the pin cell and sends it back to XSProc to obtain the temperature corrected cross sections. This process is repeated until the convergence criteria (tolerance on BISON solve, or number of time steps) is reached. Models have been constructed and run for both uranium oxide and uranium silicide fuels. These models demonstrate a clear difference in power shape that is not accounted for in a stand-alone BISON run. Future work involves improving the user interface (UI), likely through integration with the Nuclear Energy Advanced Modeling and Simulation (NEAMS) Workbench. Furthermore, automating the input creation would ease the user experience. The next priority is to continue coupling the work with other codes in the SHARP package. Efforts on other projects include work to couple the Nek5000 thermo-hydraulics code to MOOSE, but this is in the preliminary stages.« less

Thermal-hydraulic simulation of mercury target concepts for a pulsed spallation neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Siman-Tov, M.; Wendel, M.; Haines, J.

1996-06-01

The Oak Ridge Spallation Neutron Source (ORSNS) is a high-power, accelerator-based pulsed spallation neutron source being designed by a multi-laboratory team led by Oak Ridge National Laboratory to achieve very high fluxes of neutrons for scientific experiments. The ORSNS is projected to have a 1 MW proton beam upgradable to 5 MW. About 60% of the beam power (1-5 MW, 17-83 kJ/pulse in 0.5 microsec at 60 cps) is deposited in the liquid metal (mercury) target having the dimensions of 65x30x10 cm (about 19.5 liter). Peak steady state power density is about 150 and 785 MW/m{sup 3} for 1 MWmore » and 5 MW beam respectively, whereas peak pulsed power density is as high as 5.2 and 26.1 GW/m{sup 3}, respectively. The peak pulse temperature rise rate is 14 million C/s (for 5 MW beam) whereas the total pulse temperature rise is only 7 C. In addition to thermal shock and materials compatibility, key feasibility issues for the target are related to its thermal-hydraulic performance. This includes proper flow distribution, flow reversals, possible {open_quotes}hot spots{close_quotes} and the challenge of mitigating the effects of thermal shock through possible injection of helium bubbles throughout the mercury volume or other concepts. The general computational fluid dynamics (CFD) code CFDS-FLOW3D was used to simulate the thermal and flow distribution in three preliminary concepts of the mercury target. Very initial CFD simulation of He bubbles injection demonstrates some potential for simulating behavior of He bubbles in flowing mercury. Much study and development will be required to be able to `predict`, even in a crude way, such a complex phenomena. Future direction in both design and R&D is outlined.« less

A hydrogen-oxygen rocket engine coolant passage design program (RECOP) for fluid-cooled thrust chambers and nozzles

NASA Technical Reports Server (NTRS)

Tomsik, Thomas M.

1994-01-01

The design of coolant passages in regeneratively cooled thrust chambers is critical to the operation and safety of a rocket engine system. Designing a coolant passage is a complex thermal and hydraulic problem requiring an accurate understanding of the heat transfer between the combustion gas and the coolant. Every major rocket engine company has invested in the development of thrust chamber computer design and analysis tools; two examples are Rocketdyne's REGEN code and Aerojet's ELES program. In an effort to augment current design capabilities for government and industry, the NASA Lewis Research Center is developing a computer model to design coolant passages for advanced regeneratively cooled thrust chambers. The RECOP code incorporates state-of-the-art correlations, numerical techniques and design methods, certainly minimum requirements for generating optimum designs of future space chemical engines. A preliminary version of the RECOP model was recently completed and code validation work is in progress. This paper introduces major features of RECOP and compares the analysis to design points for the first test case engine; the Pratt & Whitney RL10A-3-3A thrust chamber.

Assessment of a hybrid finite element and finite volume code for turbulent incompressible flows

DOE PAGES

Xia, Yidong; Wang, Chuanjin; Luo, Hong; ...

2015-12-15

Hydra-TH is a hybrid finite-element/finite-volume incompressible/low-Mach flow simulation code based on the Hydra multiphysics toolkit being developed and used for thermal-hydraulics applications. In the present work, a suite of verification and validation (V&V) test problems for Hydra-TH was defined to meet the design requirements of the Consortium for Advanced Simulation of Light Water Reactors (CASL). The intent for this test problem suite is to provide baseline comparison data that demonstrates the performance of the Hydra-TH solution methods. The simulation problems vary in complexity from laminar to turbulent flows. A set of RANS and LES turbulence models were used in themore » simulation of four classical test problems. Numerical results obtained by Hydra-TH agreed well with either the available analytical solution or experimental data, indicating the verified and validated implementation of these turbulence models in Hydra-TH. Where possible, we have attempted some form of solution verification to identify sensitivities in the solution methods, and to suggest best practices when using the Hydra-TH code.« less

Assessment of a hybrid finite element and finite volume code for turbulent incompressible flows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xia, Yidong; Wang, Chuanjin; Luo, Hong

Hydra-TH is a hybrid finite-element/finite-volume incompressible/low-Mach flow simulation code based on the Hydra multiphysics toolkit being developed and used for thermal-hydraulics applications. In the present work, a suite of verification and validation (V&V) test problems for Hydra-TH was defined to meet the design requirements of the Consortium for Advanced Simulation of Light Water Reactors (CASL). The intent for this test problem suite is to provide baseline comparison data that demonstrates the performance of the Hydra-TH solution methods. The simulation problems vary in complexity from laminar to turbulent flows. A set of RANS and LES turbulence models were used in themore » simulation of four classical test problems. Numerical results obtained by Hydra-TH agreed well with either the available analytical solution or experimental data, indicating the verified and validated implementation of these turbulence models in Hydra-TH. Where possible, we have attempted some form of solution verification to identify sensitivities in the solution methods, and to suggest best practices when using the Hydra-TH code.« less

Modeling of Non-Homogeneous Containment Atmosphere in the ThAI Experimental Facility Using a CFD Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Babic, Miroslav; Kljenak, Ivo; Mavko, Borut

2006-07-01

The CFD code CFX4.4 was used to simulate an experiment in the ThAI facility, which was designed for investigation of thermal-hydraulic processes during a severe accident inside a Light Water Reactor containment. In the considered experiment, air was initially present in the vessel, and helium and steam were injected during different phases of the experiment at various mass flow rates and at different locations. The main purpose of the simulation was to reproduce the non-homogeneous temperature and species concentration distributions in the ThAI experimental facility. A three-dimensional model of the ThAI vessel for the CFX4.4 code was developed. The flowmore » in the simulation domain was modeled as single-phase. Steam condensation on vessel walls was modeled as a sink of mass and energy using a correlation that was originally developed for an integral approach. A simple model of bulk phase change was also introduced. The calculated time-dependent variables together with temperature and concentration distributions at the end of experiment phases are compared to experimental results. (authors)« less

Experimental Investigation of Natural-Circulation Flow Behavior Under Low-Power/Low-Pressure Conditions in the Large-Scale PANDA Facility

DOE Office of Scientific and Technical Information (OSTI.GOV)

Auban, Olivier; Paladino, Domenico; Zboray, Robert

2004-12-15

Twenty-five tests have been carried out in the large-scale thermal-hydraulic facility PANDA to investigate natural-circulation and stability behavior under low-pressure/low-power conditions, when void flashing might play an important role. This work, which extends the current experimental database to a large geometric scale, is of interest notably with regard to the start-up procedures in natural-circulation-cooled boiling water reactors. It should help the understanding of the physical phenomena that may cause flow instability in such conditions and can be used for validation of thermal-hydraulics system codes. The tests were performed at a constant power, balanced by a specific condenser heat removal capacity.more » The test matrix allowed the reactor pressure vessel power and pressure to be varied, as well as other parameters influencing the natural-circulation flow. The power spectra of flow oscillations showed in a few tests a major and unique resonance peak, and decay ratios between 0.5 and 0.9 have been found. The remainder of the tests showed an even more pronounced stable behavior. A classification of the tests is presented according to the circulation modes (from single-phase to two-phase flow) that could be assumed and particularly to the importance and the localization of the flashing phenomenon.« less

Uncertainty in dual permeability model parameters for structured soils.

PubMed

Arora, B; Mohanty, B P; McGuire, J T

2012-01-01

Successful application of dual permeability models (DPM) to predict contaminant transport is contingent upon measured or inversely estimated soil hydraulic and solute transport parameters. The difficulty in unique identification of parameters for the additional macropore- and matrix-macropore interface regions, and knowledge about requisite experimental data for DPM has not been resolved to date. Therefore, this study quantifies uncertainty in dual permeability model parameters of experimental soil columns with different macropore distributions (single macropore, and low- and high-density multiple macropores). Uncertainty evaluation is conducted using adaptive Markov chain Monte Carlo (AMCMC) and conventional Metropolis-Hastings (MH) algorithms while assuming 10 out of 17 parameters to be uncertain or random. Results indicate that AMCMC resolves parameter correlations and exhibits fast convergence for all DPM parameters while MH displays large posterior correlations for various parameters. This study demonstrates that the choice of parameter sampling algorithms is paramount in obtaining unique DPM parameters when information on covariance structure is lacking, or else additional information on parameter correlations must be supplied to resolve the problem of equifinality of DPM parameters. This study also highlights the placement and significance of matrix-macropore interface in flow experiments of soil columns with different macropore densities. Histograms for certain soil hydraulic parameters display tri-modal characteristics implying that macropores are drained first followed by the interface region and then by pores of the matrix domain in drainage experiments. Results indicate that hydraulic properties and behavior of the matrix-macropore interface is not only a function of saturated hydraulic conductivity of the macroporematrix interface ( K sa ) and macropore tortuosity ( l f ) but also of other parameters of the matrix and macropore domains.

Uncertainty in dual permeability model parameters for structured soils

NASA Astrophysics Data System (ADS)

Arora, B.; Mohanty, B. P.; McGuire, J. T.

2012-01-01

Successful application of dual permeability models (DPM) to predict contaminant transport is contingent upon measured or inversely estimated soil hydraulic and solute transport parameters. The difficulty in unique identification of parameters for the additional macropore- and matrix-macropore interface regions, and knowledge about requisite experimental data for DPM has not been resolved to date. Therefore, this study quantifies uncertainty in dual permeability model parameters of experimental soil columns with different macropore distributions (single macropore, and low- and high-density multiple macropores). Uncertainty evaluation is conducted using adaptive Markov chain Monte Carlo (AMCMC) and conventional Metropolis-Hastings (MH) algorithms while assuming 10 out of 17 parameters to be uncertain or random. Results indicate that AMCMC resolves parameter correlations and exhibits fast convergence for all DPM parameters while MH displays large posterior correlations for various parameters. This study demonstrates that the choice of parameter sampling algorithms is paramount in obtaining unique DPM parameters when information on covariance structure is lacking, or else additional information on parameter correlations must be supplied to resolve the problem of equifinality of DPM parameters. This study also highlights the placement and significance of matrix-macropore interface in flow experiments of soil columns with different macropore densities. Histograms for certain soil hydraulic parameters display tri-modal characteristics implying that macropores are drained first followed by the interface region and then by pores of the matrix domain in drainage experiments. Results indicate that hydraulic properties and behavior of the matrix-macropore interface is not only a function of saturated hydraulic conductivity of the macroporematrix interface (Ksa) and macropore tortuosity (lf) but also of other parameters of the matrix and macropore domains.

Numerical Analysis of Temperature Gradients and Interface Shape During Directional Solidification of Al and Al-Cu Alloy Under Microgravity Conditions

NASA Technical Reports Server (NTRS)

Bune, Andris V.; Sen, Subhayu; Mukherjee, Sundeep; Catalina, Adrian; Stefanescu, Doru M.

1999-01-01

Numerical modeling was undertaken to analyze the influence of radial thermal gradient on solid/liquid (s/1) interface shape and convection patterns during solidification of pure Al and Al-4 wt% Cu alloy. The objective of the numerical task was to predict the influence of convective velocity on an insoluble particle near a s/l interface. These predictions would then be used to define the minimum gravity level (g) required to investigate the fundamental physics of interaction between a particle and a s/I interface. To satisfy this objective, steady state calculations were performed for different gravity levels and orientations with the gravity vector. ne furnace configuration used in this analysis is the proposed International Space Station Furnace, Quench Module Insert (QMI) 1. Results from a thermal model of the furnace core were used as initial boundary conditions for solidification modeling. General model of binary alloy solidification was based on the finite element code FIDAP. It was found that for the worst case orientation of 90 degrees with the gravity vector and a g level of 10(exp -4)g(sub o) (g(sub o) = 9.8 m/s(exp 2)) the dominant forces acting on the particle would be the fundamental drag and interfacial forces.

Deciphering neuronal population codes for acute thermal pain

NASA Astrophysics Data System (ADS)

Chen, Zhe; Zhang, Qiaosheng; Phuong Sieu Tong, Ai; Manders, Toby R.; Wang, Jing

2017-06-01

Objective. Pain is defined as an unpleasant sensory and emotional experience associated with actual or potential tissue damage, or described in terms of such damage. Current pain research mostly focuses on molecular and synaptic changes at the spinal and peripheral levels. However, a complete understanding of pain mechanisms requires the physiological study of the neocortex. Our goal is to apply a neural decoding approach to read out the onset of acute thermal pain signals, which can be used for brain-machine interface. Approach. We used micro wire arrays to record ensemble neuronal activities from the primary somatosensory cortex (S1) and anterior cingulate cortex (ACC) in freely behaving rats. We further investigated neural codes for acute thermal pain at both single-cell and population levels. To detect the onset of acute thermal pain signals, we developed a novel latent state-space framework to decipher the sorted or unsorted S1 and ACC ensemble spike activities, which reveal information about the onset of pain signals. Main results. The state space analysis allows us to uncover a latent state process that drives the observed ensemble spike activity, and to further detect the ‘neuronal threshold’ for acute thermal pain on a single-trial basis. Our method achieved good detection performance in sensitivity and specificity. In addition, our results suggested that an optimal strategy for detecting the onset of acute thermal pain signals may be based on combined evidence from S1 and ACC population codes. Significance. Our study is the first to detect the onset of acute pain signals based on neuronal ensemble spike activity. It is important from a mechanistic viewpoint as it relates to the significance of S1 and ACC activities in the regulation of the acute pain onset.

Thermal hydraulic behavior and efficiency analysis of an all-vanadium redox flow battery

NASA Astrophysics Data System (ADS)

Xiong, Binyu; Zhao, Jiyun; Tseng, K. J.; Skyllas-Kazacos, Maria; Lim, Tuti Mariana; Zhang, Yu

2013-11-01

Vanadium redox flow batteries (VRBs) are very competitive for large-capacity energy storage in power grids and in smart buildings due to low maintenance costs, high design flexibility, and long cycle life. Thermal hydraulic modeling of VRB energy storage systems is an important issue and temperature has remarkable impacts on the battery efficiency, the lifetime of material and the stability of the electrolytes. In this paper, a lumped model including auxiliary pump effect is developed to investigate the VRB temperature responses under different operating and surrounding environmental conditions. The impact of electrolyte flow rate and temperature on the battery electrical characteristics and efficiencies are also investigated. A one kilowatt VRB system is selected to conduct numerical simulations. The thermal hydraulic model is benchmarked with experimental data and good agreement is found. Simulation results show that pump power is sensitive to hydraulic design and flow rates. The temperature in the stack and tanks rises up about 10 °C under normal operating conditions for the stack design and electrolyte volume selected. An optimal flow rate of around 90 cm3 s-1 is obtained for the proposed battery configuration to maximize battery efficiency. The models developed in this paper can also be used for the development of a battery control strategy to achieve satisfactory thermal hydraulic performance and maximize energy efficiency.

The 25 kWe solar thermal Stirling hydraulic engine system: Conceptual design

NASA Technical Reports Server (NTRS)

White, Maurice; Emigh, Grant; Noble, Jack; Riggle, Peter; Sorenson, Torvald

1988-01-01

The conceptual design and analysis of a solar thermal free-piston Stirling hydraulic engine system designed to deliver 25 kWe when coupled to a 11 meter test bed concentrator is documented. A manufacturing cost assessment for 10,000 units per year was made. The design meets all program objectives including a 60,000 hr design life, dynamic balancing, fully automated control, more than 33.3 percent overall system efficiency, properly conditioned power, maximum utilization of annualized insolation, and projected production costs. The system incorporates a simple, rugged, reliable pool boiler reflux heat pipe to transfer heat from the solar receiver to the Stirling engine. The free-piston engine produces high pressure hydraulic flow which powers a commercial hydraulic motor that, in turn, drives a commercial rotary induction generator. The Stirling hydraulic engine uses hermetic bellows seals to separate helium working gas from hydraulic fluid which provides hydrodynamic lubrication to all moving parts. Maximum utilization of highly refined, field proven commercial components for electric power generation minimizes development cost and risk.

The InterFrost benchmark of Thermo-Hydraulic codes for cold regions hydrology - first inter-comparison results

NASA Astrophysics Data System (ADS)

Grenier, Christophe; Roux, Nicolas; Anbergen, Hauke; Collier, Nathaniel; Costard, Francois; Ferrry, Michel; Frampton, Andrew; Frederick, Jennifer; Holmen, Johan; Jost, Anne; Kokh, Samuel; Kurylyk, Barret; McKenzie, Jeffrey; Molson, John; Orgogozo, Laurent; Rivière, Agnès; Rühaak, Wolfram; Selroos, Jan-Olof; Therrien, René; Vidstrand, Patrik

2015-04-01

The impacts of climate change in boreal regions has received considerable attention recently due to the warming trends that have been experienced in recent decades and are expected to intensify in the future. Large portions of these regions, corresponding to permafrost areas, are covered by water bodies (lakes, rivers) that interact with the surrounding permafrost. For example, the thermal state of the surrounding soil influences the energy and water budget of the surface water bodies. Also, these water bodies generate taliks (unfrozen zones below) that disturb the thermal regimes of permafrost and may play a key role in the context of climate change. Recent field studies and modeling exercises indicate that a fully coupled 2D or 3D Thermo-Hydraulic (TH) approach is required to understand and model the past and future evolution of landscapes, rivers, lakes and associated groundwater systems in a changing climate. However, there is presently a paucity of 3D numerical studies of permafrost thaw and associated hydrological changes, and the lack of study can be partly attributed to the difficulty in verifying multi-dimensional results produced by numerical models. Numerical approaches can only be validated against analytical solutions for a purely thermic 1D equation with phase change (e.g. Neumann, Lunardini). When it comes to the coupled TH system (coupling two highly non-linear equations), the only possible approach is to compare the results from different codes to provided test cases and/or to have controlled experiments for validation. Such inter-code comparisons can propel discussions to try to improve code performances. A benchmark exercise was initialized in 2014 with a kick-off meeting in Paris in November. Participants from USA, Canada, Germany, Sweden and France convened, representing altogether 13 simulation codes. The benchmark exercises consist of several test cases inspired by existing literature (e.g. McKenzie et al., 2007) as well as new ones. They range from simpler, purely thermal cases (benchmark T1) to more complex, coupled 2D TH cases (benchmarks TH1, TH2, and TH3). Some experimental cases conducted in cold room complement the validation approach. A web site hosted by LSCE (Laboratoire des Sciences du Climat et de l'Environnement) is an interaction platform for the participants and hosts the test cases database at the following address: https://wiki.lsce.ipsl.fr/interfrost. The results of the first stage of the benchmark exercise will be presented. We will mainly focus on the inter-comparison of participant results for the coupled cases (TH1, TH2 & TH3). Further perspectives of the exercise will also be presented. Extensions to more complex physical conditions (e.g. unsaturated conditions and geometrical deformations) are contemplated. In addition, 1D vertical cases of interest to the Climate Modeling community will be proposed. Keywords: Permafrost; Numerical modeling; River-soil interaction; Arctic systems; soil freeze-thaw

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bakosi, Jozsef; Christon, Mark A.; Francois, Marianne M.

Progress is reported on computational capabilities for the grid-to-rod-fretting (GTRF) problem of pressurized water reactors. Numeca's Hexpress/Hybrid mesh generator is demonstrated as an excellent alternative to generating computational meshes for complex flow geometries, such as in GTRF. Mesh assessment is carried out using standard industrial computational fluid dynamics practices. Hydra-TH, a simulation code developed at LANL for reactor thermal-hydraulics, is demonstrated on hybrid meshes, containing different element types. A series of new Hydra-TH calculations has been carried out collecting turbulence statistics. Preliminary results on the newly generated meshes are discussed; full analysis will be documented in the L3 milestone, THM.CFD.P5.05,more » Sept. 2012.« less

Comparison of the PHISICS/RELAP5-3D ring and block model results for phase I of the OECD/NEA MHTGR-350 benchmark

DOE PAGES

Strydom, G.; Epiney, A. S.; Alfonsi, Andrea; ...

2015-12-02

The PHISICS code system has been under development at INL since 2010. It consists of several modules providing improved coupled core simulation capability: INSTANT (3D nodal transport core calculations), MRTAU (depletion and decay heat generation) and modules performing criticality searches, fuel shuffling and generalized perturbation. Coupling of the PHISICS code suite to the thermal hydraulics system code RELAP5-3D was finalized in 2013, and as part of the verification and validation effort the first phase of the OECD/NEA MHTGR-350 Benchmark has now been completed. The theoretical basis and latest development status of the coupled PHISICS/RELAP5-3D tool are described in more detailmore » in a concurrent paper. This paper provides an overview of the OECD/NEA MHTGR-350 Benchmark and presents the results of Exercises 2 and 3 defined for Phase I. Exercise 2 required the modelling of a stand-alone thermal fluids solution at End of Equilibrium Cycle for the Modular High Temperature Reactor (MHTGR). The RELAP5-3D results of four sub-cases are discussed, consisting of various combinations of coolant bypass flows and material thermophysical properties. Exercise 3 required a coupled neutronics and thermal fluids solution, and the PHISICS/RELAP5-3D code suite was used to calculate the results of two sub-cases. The main focus of the paper is a comparison of results obtained with the traditional RELAP5-3D “ring” model approach against a much more detailed model that include kinetics feedback on individual block level and thermal feedbacks on a triangular sub-mesh. The higher fidelity that can be obtained by this “block” model is illustrated with comparison results on the temperature, power density and flux distributions. Furthermore, it is shown that the ring model leads to significantly lower fuel temperatures (up to 10%) when compared with the higher fidelity block model, and that the additional model development and run-time efforts are worth the gains obtained in the improved spatial temperature and flux distributions.« less

TRANSFORM - TRANsient Simulation Framework of Reconfigurable Models

DOE Office of Scientific and Technical Information (OSTI.GOV)

Greenwood, Michael S; Cetiner, Mustafa S; Fugate, David L

Existing development tools for early stage design and scoping of energy systems are often time consuming to use, proprietary, and do not contain the necessary function to model complete systems (i.e., controls, primary, and secondary systems) in a common platform. The Modelica programming language based TRANSFORM tool (1) provides a standardized, common simulation environment for early design of energy systems (i.e., power plants), (2) provides a library of baseline component modules to be assembled into full plant models using available geometry, design, and thermal-hydraulic data, (3) defines modeling conventions for interconnecting component models, and (4) establishes user interfaces and supportmore » tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.« less

An Experimental and Modeling Study of Evaporation from Bare Soils Subjected to Natural Boundary Conditions at the Land-Atmospheric Interface

NASA Astrophysics Data System (ADS)

Smits, K. M.; Ngo, V. V.; Cihan, A.; Sakaki, T.; Illangasekare, T. H.; kathleen m smits

2011-12-01

Bare soil evaporation is a key process for water exchange between the land and the atmosphere and an important component of the water balance in semiarid and arid regions. However, there is no agreement on the best methodology to determine evaporation under different boundary conditions. Because it is difficult to measure evaporation from soil,with the exception of using lysimeters, numerous formulations have been proposed to establish a relationship between the rate of evaporation and soil moisture and/or soil temperature and thermal properties. Different formulations vary in how they partition available energy and include, among others, a classical bulk aerodynamic formulation which requires knowledge of the relative humidity at the soil surface and a more non-traditional heat balance method which requires knowledge of soil temperature and soil thermal properties. A need exists to systematically compare existing methods to experimental data under highly controlled conditions not achievable in the field. The goal of this work is to perform controlled experiments under transient conditions of soil moisture, temperature and wind at the land/atmospheric interface to test different conceptual and mathematical formulations for evaporation rate estimates and to develop appropriate numerical models to be used in simulations. In this study, to better understand the coupled water-vapor-heat flow processes in the shallow subsurface near the land surface, we modified a previously developed theory that allows non-equilibrium liquid/gas phase change with gas phase vapor diffusion to better account for evaporation under dry soil conditions. This theory was used to compare estimates of evaporation based on different formulations of the bulk aerodynamic and heat balance methods. In order to experimentally validate the numerical formulations/code, we performed a series of two-dimensional physical model experiments under varying boundary conditions using test sand for which the hydraulic and thermal properties were well characterized. We developed a unique two dimensional cell apparatus equipped with a network of sensors for automated and continuous monitoring of soil moisture, soil and air temperature and relative humidity, and wind velocity. Precision data under well-controlled transient heat and wind boundary conditions was generated. Results from numerical simulations were compared with experimental data. Results demonstrate the importance of properly characterizing soil thermal properties and accounting for dry soil conditions to properly estimate evaporation. Initial comparisons of various formulations of evaporation demonstrate the need for joint evaluation of heat and mass transfer for better modeling accuracy. Detailed comparisons are still underway. This knowledge is applicable to many current hydrologic and environmental problems to include climate modeling and the simulation of contaminant transport and volatilization in the shallow subsurface.

Altered Landscapes and Groundwater Sustainability — Exploring Impacts with Induced Polarization, DC Resistivity, and Thermal Tracing

NASA Astrophysics Data System (ADS)

Eddy-Miller, C.; Caldwell, R.; Wheeler, J.; McCarthy, P.; Binley, A. M.; Constantz, J. E.; Stonestrom, D. A.

2009-12-01

Anthropogenically impacted landscapes constitute rising proportions of the Earth’s surface that are characterized by generally elevated nutrient and sediment loadings concurrent with increased consumptive water withdrawals. In recent years a growing number of hydraulically engineered riparian habitat restoration projects have attempted to ameliorate negative impacts of land use on groundwater-surface water systems resulting, e.g., from agricultural practices and urban development. Often the nature of groundwater-surface water interactions in pre- and minimally altered systems is poorly known, making it difficult to assess the impacts of land use and restoration projects on groundwater sustainability. Traditional assessments of surface water parameters (flow, temperature, dissolved oxygen, biotic composition, etc.) can be complemented by hydraulic and thermal measurements to better understand the important role played by groundwater-surface water interactions. Hydraulic and thermal measurements are usually limited to point samples, however, making non-invasive and spatially extensive geophysical characterizations an attractive additional tool. Groundwater-surface water interactions along the Smith River, a tributary to the Missouri River in Montana, and Fish Creek and Flat Creek, tributaries to the Snake River in Wyoming, are being examined using a combination of hydraulic measurements, thermal tracing, and electrical-property imaging. Ninety-two direct-current (DC) resistivity and induced polarization cross sections were obtained at stream transects covering a wide variety of hydrogeologic settings ranging from shallow bedrock to thick alluvial sequences, nature of groundwater-surface water interactions (always gaining, always losing, or seasonally varying) and anthropogenic impacts (minimal low-intensity agriculture to major landscape engineering, including channel reconstruction). DC resistivity and induced polarization delineated mutually distinct features related to hydraulic architecture. For example, induced polarization imaging resolved channel-edge muck deposits that are presumed to be sites of low hydraulic conductivity, chemical reduction, and metal accumulation. DC resistivity delineated bedrock-alluvium contacts and showed potential for tracking changes in salinization. While electrical properties cannot substitute for hydraulic and thermal data, the addition of relatively rapidly acquired, spatially extensive resistivity and induced polarization imaging offers synergistic opportunities for interpretive hydrologic investigations.

Integration of Flex Nozzle System and Electro Hydraulic Actuators to Solid Rocket Motors

NASA Astrophysics Data System (ADS)

Nayani, Kishore Nath; Bajaj, Dinesh Kumar

2017-10-01

A rocket motor assembly comprised of solid rocket motor and flex nozzle system. Integration of flex nozzle system and hydraulic actuators to the solid rocket motors are done after transportation to the required place where integration occurred. The flex nozzle system is integrated to the rocket motor in horizontal condition and the electro hydraulic actuators are assembled to the flex nozzle systems. The electro hydraulic actuators are connected to the hydraulic power pack to operate the actuators. The nozzle-motor critical interface are insulation diametrical compression, inhibition resin-28, insulation facial compression, shaft seal `O' ring compression and face seal `O' ring compression.

Nuclear Engineering Computer Modules, Thermal-Hydraulics, TH-3: High Temperature Gas Cooled Reactor Thermal-Hydraulics.

ERIC Educational Resources Information Center

Reihman, Thomas C.

This learning module is concerned with the temperature field, the heat transfer rates, and the coolant pressure drop in typical high temperature gas-cooled reactor (HTGR) fuel assemblies. As in all of the modules of this series, emphasis is placed on developing the theory and demonstrating its use with a simplified model. The heart of the module…

The kinetics of aerosol particle formation and removal in NPP severe accidents

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zatevakhin, Mikhail A.; Arefiev, Valentin K.; Semashko, Sergey E.

2016-06-08

Severe Nuclear Power Plant (NPP) accidents are accompanied by release of a massive amount of energy, radioactive products and hydrogen into the atmosphere of the NPP containment. A valid estimation of consequences of such accidents can only be carried out through the use of the integrated codes comprising a description of the basic processes which determine the consequences. A brief description of a coupled aerosol and thermal–hydraulic code to be used for the calculation of the aerosol kinetics within the NPP containment in case of a severe accident is given. The code comprises a KIN aerosol unit integrated into themore » KUPOL-M thermal–hydraulic code. Some features of aerosol behavior in severe NPP accidents are briefly described.« less

A Graphical-User Interface for the U. S. Geological Survey's SUTRA Code using Argus ONE (for simulation of variable-density saturated-unsaturated ground-water flow with solute or energy transport)

USGS Publications Warehouse

Voss, Clifford I.; Boldt, David; Shapiro, Allen M.

1997-01-01

This report describes a Graphical-User Interface (GUI) for SUTRA, the U.S. Geological Survey (USGS) model for saturated-unsaturated variable-fluid-density ground-water flow with solute or energy transport,which combines a USGS-developed code that interfaces SUTRA with Argus ONE, a commercial software product developed by Argus Interware. This product, known as Argus Open Numerical Environments (Argus ONETM), is a programmable system with geographic-information-system-like (GIS-like) functionality that includes automated gridding and meshing capabilities for linking geospatial information with finite-difference and finite-element numerical model discretizations. The GUI for SUTRA is based on a public-domain Plug-In Extension (PIE) to Argus ONE that automates the use of ArgusONE to: automatically create the appropriate geospatial information coverages (information layers) for SUTRA, provide menus and dialogs for inputting geospatial information and simulation control parameters for SUTRA, and allow visualization of SUTRA simulation results. Following simulation control data and geospatial data input bythe user through the GUI, ArgusONE creates text files in a format required for normal input to SUTRA,and SUTRA can be executed within the Argus ONE environment. Then, hydraulic head, pressure, solute concentration, temperature, saturation and velocity results from the SUTRA simulation may be visualized. Although the GUI for SUTRA discussed in this report provides all of the graphical pre- and post-processor functions required for running SUTRA, it is also possible for advanced users to apply programmable features within Argus ONE to modify the GUI to meet the unique demands of particular ground-water modeling projects.

Servo-hydraulic actuator in controllable canonical form: Identification and experimental validation

NASA Astrophysics Data System (ADS)

Maghareh, Amin; Silva, Christian E.; Dyke, Shirley J.

2018-02-01

Hydraulic actuators have been widely used to experimentally examine structural behavior at multiple scales. Real-time hybrid simulation (RTHS) is one innovative testing method that largely relies on such servo-hydraulic actuators. In RTHS, interface conditions must be enforced in real time, and controllers are often used to achieve tracking of the desired displacements. Thus, neglecting the dynamics of hydraulic transfer system may result either in system instability or sub-optimal performance. Herein, we propose a nonlinear dynamical model for a servo-hydraulic actuator (a.k.a. hydraulic transfer system) coupled with a nonlinear physical specimen. The nonlinear dynamical model is transformed into controllable canonical form for further tracking control design purposes. Through a number of experiments, the controllable canonical model is validated.

THERMAL-ENERGY STORAGE IN A DEEP SANDSTONE AQUIFER IN MINNESOTA: FIELD OBSERVATIONS AND THERMAL ENERGY-TRANSPORT MODELING.

USGS Publications Warehouse

Miller, R.T.

1986-01-01

A study of the feasibility of storing heated water in a deep sandstone aquifer in Minnesota is described. The aquifer consists of four hydraulic zones that are areally anisotropic and have average hydraulic conductivities that range from 0. 03 to 1. 2 meters per day. A preliminary axially symmetric, nonisothermal, isotropic, single-phase, radial-flow, thermal-energy-transport model was constructed to investigate the sensitivity of model simulation to various hydraulic and thermal properties of the aquifer. A three-dimensional flow and thermal-energy transport model was constructed to incorporate the areal anisotropy of the aquifer. Analytical solutions of equations describing areally anisotropic groundwater flow around a doublet-well system were used to specify model boundary conditions for simulation of heat injection. The entire heat-injection-testing period of approximately 400 days was simulated. Model-computed temperatures compared favorably with field-recorded temperatures, with differences of no more than plus or minus 8 degree C. For each test cycle, model-computed aquifer thermal efficiency, defined as total heat withdrawn divided by total heat injected, was within plus or minus 2% of the field-calculated values.

Nuclear modules for space electric propulsion

NASA Technical Reports Server (NTRS)

Difilippo, F. C.

1998-01-01

Analysis of interplanetary cargo and piloted missions requires calculations of the performances and masses of subsystems to be integrated in a final design. In a preliminary and scoping stage the designer needs to evaluate options iteratively by using fast computer simulations. The Oak Ridge National Laboratory (ORNL) has been involved in the development of models and calculational procedures for the analysis (neutronic and thermal hydraulic) of power sources for nuclear electric propulsion. The nuclear modules will be integrated into the whole simulation of the nuclear electric propulsion system. The vehicles use either a Brayton direct-conversion cycle, using the heated helium from a NERVA-type reactor, or a potassium Rankine cycle, with the working fluid heated on the secondary side of a heat exchanger and lithium on the primary side coming from a fast reactor. Given a set of input conditions, the codes calculate composition. dimensions, volumes, and masses of the core, reflector, control system, pressure vessel, neutron and gamma shields, as well as the thermal hydraulic conditions of the coolant, clad and fuel. Input conditions are power, core life, pressure and temperature of the coolant at the inlet of the core, either the temperature of the coolant at the outlet of the core or the coolant mass flow and the fluences and integrated doses at the cargo area. Using state-of-the-art neutron cross sections and transport codes, a database was created for the neutronic performance of both reactor designs. The free parameters of the models are the moderator/fuel mass ratio for the NERVA reactor and the enrichment and the pitch of the lattice for the fast reactor. Reactivity and energy balance equations are simultaneously solved to find the reactor design. Thermalhydraulic conditions are calculated by solving the one-dimensional versions of the equations of conservation of mass, energy, and momentum with compressible flow.

Validation and Calibration of Nuclear Thermal Hydraulics Multiscale Multiphysics Models - Subcooled Flow Boiling Study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anh Bui; Nam Dinh; Brian Williams

In addition to validation data plan, development of advanced techniques for calibration and validation of complex multiscale, multiphysics nuclear reactor simulation codes are a main objective of the CASL VUQ plan. Advanced modeling of LWR systems normally involves a range of physico-chemical models describing multiple interacting phenomena, such as thermal hydraulics, reactor physics, coolant chemistry, etc., which occur over a wide range of spatial and temporal scales. To a large extent, the accuracy of (and uncertainty in) overall model predictions is determined by the correctness of various sub-models, which are not conservation-laws based, but empirically derived from measurement data. Suchmore » sub-models normally require extensive calibration before the models can be applied to analysis of real reactor problems. This work demonstrates a case study of calibration of a common model of subcooled flow boiling, which is an important multiscale, multiphysics phenomenon in LWR thermal hydraulics. The calibration process is based on a new strategy of model-data integration, in which, all sub-models are simultaneously analyzed and calibrated using multiple sets of data of different types. Specifically, both data on large-scale distributions of void fraction and fluid temperature and data on small-scale physics of wall evaporation were simultaneously used in this work’s calibration. In a departure from traditional (or common-sense) practice of tuning/calibrating complex models, a modern calibration technique based on statistical modeling and Bayesian inference was employed, which allowed simultaneous calibration of multiple sub-models (and related parameters) using different datasets. Quality of data (relevancy, scalability, and uncertainty) could be taken into consideration in the calibration process. This work presents a step forward in the development and realization of the “CIPS Validation Data Plan” at the Consortium for Advanced Simulation of LWRs to enable quantitative assessment of the CASL modeling of Crud-Induced Power Shift (CIPS) phenomenon, in particular, and the CASL advanced predictive capabilities, in general. This report is prepared for the Department of Energy’s Consortium for Advanced Simulation of LWRs program’s VUQ Focus Area.« less

McStas event logger: Definition and applications

NASA Astrophysics Data System (ADS)

Bergbäck Knudsen, Erik; Bryndt Klinkby, Esben; Kjær Willendrup, Peter

2014-02-01

Functionality is added to the McStas neutron ray-tracing code, which allows individual neutron states before and after a scattering to be temporarily stored, and analysed. This logging mechanism has multiple uses, including studies of longitudinal intensity loss in neutron guides and guide coating design optimisations. Furthermore, the logging method enables the cold/thermal neutron induced gamma background along the guide to be calculated from the un-reflected neutron, using a recently developed MCNPX-McStas interface.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jensen, Colby B.; Folsom, Charles P.; Davis, Cliff B.

Experimental testing in the Multi-Static Environment Rodlet Transient Test Apparatus (SERTTA) will lead the rebirth of transient fuel testing in the United States as part of the Accident Tolerant Fuels (ATF) progam. The Multi-SERTTA is comprised of four isolated pressurized environments capable of a wide variety of working fluids and thermal conditions. Ultimately, the TREAT reactor as well as the Multi-SERTTA test vehicle serve the purpose of providing desired thermal-hydraulic boundary conditions to the test specimen. The initial ATF testing in TREAT will focus on reactivity insertion accident (RIA) events using both gas and water environments including typical PWR operatingmore » pressures and temperatures. For the water test environment, a test configuration is envisioned using the expansion tank as part of the gas-filled expansion volume seen by the test to provide additional pressure relief. The heat transfer conditions during the high energy power pulses of RIA events remains a subject of large uncertainty and great importance for fuel performance predictions. To support transient experiments, the Multi-SERTTA vehicle has been modeled using RELAP5 with a baseline test specimen composed of UO2 fuel in zircaloy cladding. The modeling results show the influence of the designs of the specimen, vehicle, and transient power pulses. The primary purpose of this work is to provide input and boundary conditions to fuel performance code BISON. Therefore, studies of parameters having influence on specimen performance during RIA transients are presented including cladding oxidation, power pulse magnitude and width, cladding-to-coolant heat fluxes, fuel-to-cladding gap, transient boiling effects (modified CHF values), etc. The results show the great flexibility and capacity of the TREAT Multi-SERTTA test vehicle to provide testing under a wide range of prototypic thermal-hydraulic conditions as never done before.« less

Effect of thermal interface on heat flow in carbon nanofiber composites.

PubMed

Gardea, F; Naraghi, M; Lagoudas, D

2014-01-22

The thermal transport process in carbon nanofiber (CNF)/epoxy composites is addressed through combined micromechanics and finite element modeling, guided by experiments. The heat exchange between CNF constituents and matrix is studied by explicitly accounting for interface thermal resistance between the CNFs and the epoxy matrix. The effects of nanofiber orientation and discontinuity on heat flow and thermal conductivity of nanocomposites are investigated through simulation of the laser flash experiment technique and Fourier's model of heat conduction. Our results indicate that when continuous CNFs are misoriented with respect to the average temperature gradient, the presence of interfacial resistance does not affect the thermal conductivity of the nanocomposites, as most of the heat flow will be through CNFs; however, interface thermal resistance can significantly alter the patterns of heat flow within the nanocomposite. It was found that very high interface resistance leads to heat entrapment at the interface near to the heat source, which can promote interface thermal degradation. The magnitude of heat entrapment, quantified via the peak transient temperature rise at the interface, in the case of high thermal resistance interfaces becomes an order of magnitude more intense as compared to the case of low thermal resistance interfaces. Moreover, high interface thermal resistance in the case of discontinuous fibers leads to a nearly complete thermal isolation of the fibers from the matrix, which will marginalize the contribution of the CNF thermal conductivity to the heat transfer in the composite.

Advanced nodal neutron diffusion method with space-dependent cross sections: ILLICO-VX

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rajic, H.L.; Ougouag, A.M.

1987-01-01

Advanced transverse integrated nodal methods for neutron diffusion developed since the 1970s require that node- or assembly-homogenized cross sections be known. The underlying structural heterogeneity can be accurately accounted for in homogenization procedures by the use of heterogeneity or discontinuity factors. Other (milder) types of heterogeneity, burnup-induced or due to thermal-hydraulic feedback, can be resolved by explicitly accounting for the spatial variations of material properties. This can be done during the nodal computations via nonlinear iterations. The new method has been implemented in the code ILLICO-VX (ILLICO variable cross-section method). Numerous numerical tests were performed. As expected, the convergence ratemore » of ILLICO-VX is lower than that of ILLICO, requiring approx. 30% more outer iterations per k/sub eff/ computation. The methodology has also been implemented as the NOMAD-VX option of the NOMAD, multicycle, multigroup, two- and three-dimensional nodal diffusion depletion code. The burnup-induced heterogeneities (space dependence of cross sections) are calculated during the burnup steps.« less

Numerical Simulation of the Emergency Condenser of the SWR-1000

DOE Office of Scientific and Technical Information (OSTI.GOV)

Krepper, Eckhard; Schaffrath, Andreas; Aszodi, Attila

The SWR-1000 is a new innovative boiling water reactor (BWR) concept, which was developed by Siemens AG. This concept is characterized in particular by passive safety systems (e.g., four emergency condensers, four building condensers, eight passive pressure pulse transmitters, and six gravity-driven core-flooding lines). In the framework of the BWR Physics and Thermohydraulic Complementary Action to the European Union BWR Research and Development Cluster, emergency condenser tests were performed by Forschungszentrum Juelich at the NOKO test facility. Posttest calculations with ATHLET are presented, which aim at the determination of the removable power of the emergency condenser and its operation mode.more » The one-dimensional thermal-hydraulic code ATHLET was extended by the module KONWAR for the calculation of the heat transfer coefficient during condensation in horizontal tubes. In addition, results of conventional finite difference calculations using the code CFX-4 are presented, which investigate the natural convection during the heatup process at the secondary side of the NOKO test facility.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stimpson, Shane G.

Activities to incorporate fuel performance capabilities into the Virtual Environment for Reactor Applications (VERA) are receiving increasing attention. The multiphysics emphasis is expanding as the neutronics (MPACT) and thermal-hydraulics (CTF) packages are becoming more mature. Capturing the finer details of fuel phenomena (swelling, densification, relocation, gap closure, etc.) is the natural next step in the VERA Core Simulator (VERA-CS) development process since these phenomena are currently not directly taken into account. While several codes could be used to accomplish this, the BISON fuel performance code being developed by the Idaho National Laboratory (INL) is the focus of ongoing work inmore » the Consortium for Advanced Simulation of Light Water Reactors (CASL). Built on INL’s MOOSE framework, BISON uses the finite element method for geometric representation and a Jacobian-free Newton-Krylov (JFNK) scheme to solve systems of partial differential equations for various fuel characteristic relationships. There are several modes of operation in BISON, but, for this work, it uses a 2D azimuthally symmetric (R-Z) smeared-pellet model.« less

Development of concepts for the management of thermal resources in urban areas - Assessment of transferability from the Basel (Switzerland) and Zaragoza (Spain) case studies

NASA Astrophysics Data System (ADS)

Epting, Jannis; García-Gil, Alejandro; Huggenberger, Peter; Vázquez-Suñe, Enric; Mueller, Matthias H.

2017-05-01

The shallow subsurface in urban areas is increasingly used by shallow geothermal energy systems as a renewable energy resource and as a cheap cooling medium, e.g. for building air conditioning. In combination with further anthropogenic activities, this results in altered thermal regimes in the subsurface and the so-called subsurface urban heat island effect. Successful thermal management of urban groundwater resources requires understanding the relative contributions of the different thermal parameters and boundary conditions that result in the "present thermal state" of individual urban groundwater bodies. To evaluate the "present thermal state" of urban groundwater bodies, good quality data are required to characterize the hydraulic and thermal aquifer parameters. This process also involved adequate monitoring systems which provide consistent subsurface temperature measurements and are the basis for parameterizing numerical heat-transport models. This study is based on previous work already published for two urban groundwater bodies in Basel (CH) and Zaragoza (ES), where comprehensive monitoring networks (hydraulics and temperature) as well as calibrated high-resolution numerical flow- and heat-transport models have been analyzed. The "present thermal state" and how it developed according to the different hydraulic and thermal boundary conditions is compared to a "potential natural state" in order to assess the anthropogenic thermal changes that have already occurred in the urban groundwater bodies we investigated. This comparison allows us to describe the various processes concerning groundwater flow and thermal regimes for the different urban settings. Furthermore, the results facilitate defining goals for specific aquifer regions, including future aquifer use and urbanization, as well as evaluating the thermal use potential for these regions. As one example for a more sustainable thermal use of subsurface water resources, we introduce the thermal management concept of the "relaxation factor", which is a first approach to overcome the present policy of "first come, first served". Remediation measures to regenerate overheated urban aquifers are also introduced. The transferability of the applied methods to other urban areas is discussed. It is shown that an appropriate selection of locations for monitoring hydraulic and thermal boundary conditions make it possible to implement representative interpretations of groundwater flow and thermal regimes as well as to set up high-resolution numerical flow- and heat-transport models. Those models are the basis for the sustainable management of thermal resources.

Flexible parallel implicit modelling of coupled thermal-hydraulic-mechanical processes in fractured rocks

NASA Astrophysics Data System (ADS)

Cacace, Mauro; Jacquey, Antoine B.

2017-09-01

Theory and numerical implementation describing groundwater flow and the transport of heat and solute mass in fully saturated fractured rocks with elasto-plastic mechanical feedbacks are developed. In our formulation, fractures are considered as being of lower dimension than the hosting deformable porous rock and we consider their hydraulic and mechanical apertures as scaling parameters to ensure continuous exchange of fluid mass and energy within the fracture-solid matrix system. The coupled system of equations is implemented in a new simulator code that makes use of a Galerkin finite-element technique. The code builds on a flexible, object-oriented numerical framework (MOOSE, Multiphysics Object Oriented Simulation Environment) which provides an extensive scalable parallel and implicit coupling to solve for the multiphysics problem. The governing equations of groundwater flow, heat and mass transport, and rock deformation are solved in a weak sense (either by classical Newton-Raphson or by free Jacobian inexact Newton-Krylow schemes) on an underlying unstructured mesh. Nonlinear feedbacks among the active processes are enforced by considering evolving fluid and rock properties depending on the thermo-hydro-mechanical state of the system and the local structure, i.e. degree of connectivity, of the fracture system. A suite of applications is presented to illustrate the flexibility and capability of the new simulator to address problems of increasing complexity and occurring at different spatial (from centimetres to tens of kilometres) and temporal scales (from minutes to hundreds of years).

TOUGH-RBSN simulator for hydraulic fracture propagation within fractured media: Model validations against laboratory experiments

NASA Astrophysics Data System (ADS)

Kim, Kunhwi; Rutqvist, Jonny; Nakagawa, Seiji; Birkholzer, Jens

2017-11-01

This paper presents coupled hydro-mechanical modeling of hydraulic fracturing processes in complex fractured media using a discrete fracture network (DFN) approach. The individual physical processes in the fracture propagation are represented by separate program modules: the TOUGH2 code for multiphase flow and mass transport based on the finite volume approach; and the rigid-body-spring network (RBSN) model for mechanical and fracture-damage behavior, which are coupled with each other. Fractures are modeled as discrete features, of which the hydrological properties are evaluated from the fracture deformation and aperture change. The verification of the TOUGH-RBSN code is performed against a 2D analytical model for single hydraulic fracture propagation. Subsequently, modeling capabilities for hydraulic fracturing are demonstrated through simulations of laboratory experiments conducted on rock-analogue (soda-lime glass) samples containing a designed network of pre-existing fractures. Sensitivity analyses are also conducted by changing the modeling parameters, such as viscosity of injected fluid, strength of pre-existing fractures, and confining stress conditions. The hydraulic fracturing characteristics attributed to the modeling parameters are investigated through comparisons of the simulation results.

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

NASA Astrophysics Data System (ADS)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, S. M.; Kim, K. Y.

Printed circuit heat exchanger (PCHE) is recently considered as a recuperator for the high temperature gas cooled reactor. In this work, the zigzag-channels of a PCHE have been optimized by using three-dimensional Reynolds-Averaged Navier-Stokes (RANS) analysis and response surface approximation (RSA) modeling technique to enhance thermal-hydraulic performance. Shear stress transport turbulence model is used as a turbulence closure. The objective function is defined as a linear combination of the functions related to heat transfer and friction loss of the PCHE, respectively. Three geometric design variables viz., the ratio of the radius of the fillet to hydraulic diameter of the channels,more » the ratio of wavelength to hydraulic diameter of the channels, and the ratio of wave height to hydraulic diameter of the channels, are used for the optimization. Design points are selected through Latin-hypercube sampling. The optimal design is determined through the RSA model which uses RANS derived calculations at the design points. The results show that the optimum shape enhances considerably the thermal-hydraulic performance than a reference shape. (authors)« less

3-D Modeling of Double-Diffusive Convection During Directional Solidification of a Non-Dilute Alloy with Application to the HgCdTe Growth Under Microgravity Conditions

NASA Technical Reports Server (NTRS)

Bune, Andris V.; Gillies, Donald C.; Lehoczky, Sandor L.

1998-01-01

A numerical calculation for a non-dilute alloy solidification was performed using the FIDAP finite element code. For low growth velocities plane front solidification occurs. The location and the shape of the interface was determined using melting temperatures from the HgCdTe liquidus curve. The low thermal conductivity of the solid HgCdTe causes thermal short circuit through the ampoule walls, resulting in curved isotherms in the vicinity of the interface. Double-diffusive convection in the melt is caused by radial temperature gradients and by material density inversion with temperature. Cooling from below and the rejection at the solid-melt interface of the heavier HgTe-rich solute each tend to reduce convection. Because of these complicating factors dimensional rather then non-dimensional modeling was performed. Estimates of convection contributions for various gravity conditions was performed parametrically. For gravity levels higher then 1 0 -7 of earth's gravity it was found that the maximum convection velocity is extremely sensitive to gravity vector orientation and can be reduced at least by factor of 50% for precise orientation of the ampoule in the microgravity environment. The predicted interface shape is in agreement with one obtained experimentally by quenching. The results of 3-D modeling are compared with previous 2-D finding. A video film featuring melt convection will be presented.

IAC-1.5 - INTEGRATED ANALYSIS CAPABILITY

NASA Technical Reports Server (NTRS)

Vos, R. G.

1994-01-01

The objective of the Integrated Analysis Capability (IAC) system is to provide a highly effective, interactive analysis tool for the integrated design of large structures. IAC was developed to interface programs from the fields of structures, thermodynamics, controls, and system dynamics with an executive system and a database to yield a highly efficient multi-disciplinary system. Special attention is given to user requirements such as data handling and on-line assistance with operational features, and the ability to add new modules of the user's choice at a future date. IAC contains an executive system, a database, general utilities, interfaces to various engineering programs, and a framework for building interfaces to other programs. IAC has shown itself to be effective in automating data transfer among analysis programs. The IAC system architecture is modular in design. 1) The executive module contains an input command processor, an extensive data management system, and driver code to execute the application modules. 2) Technical modules provide standalone computational capability as well as support for various solution paths or coupled analyses. 3) Graphics and model generation modules are supplied for building and viewing models. 4) Interface modules provide for the required data flow between IAC and other modules. 5) User modules can be arbitrary executable programs or JCL procedures with no pre-defined relationship to IAC. 6) Special purpose modules are included, such as MIMIC (Model Integration via Mesh Interpolation Coefficients), which transforms field values from one model to another; LINK, which simplifies incorporation of user specific modules into IAC modules; and DATAPAC, the National Bureau of Standards statistical analysis package. The IAC database contains structured files which provide a common basis for communication between modules and the executive system, and can contain unstructured files such as NASTRAN checkpoint files, DISCOS plot files, object code, etc. The user can define groups of data and relations between them. A full data manipulation and query system operates with the database. The current interface modules comprise five groups: 1) Structural analysis - IAC contains a NASTRAN interface for standalone analysis or certain structural/control/thermal combinations. IAC provides enhanced structural capabilities for normal modes and static deformation analysis via special DMAP sequences. 2) Thermal analysis - IAC supports finite element and finite difference techniques for steady state or transient analysis. There are interfaces for the NASTRAN thermal analyzer, SINDA/SINFLO, and TRASYS II. 3) System dynamics - A DISCOS interface allows full use of this simulation program for either nonlinear time domain analysis or linear frequency domain analysis. 4) Control analysis - Interfaces for the ORACLS, SAMSAN, NBOD2, and INCA programs allow a wide range of control system analyses and synthesis techniques. 5) Graphics - The graphics packages PLOT and MOSAIC are included in IAC. PLOT generates vector displays of tabular data in the form of curves, charts, correlation tables, etc., while MOSAIC generates color raster displays of either tabular of array type data. Either DI3000 or PLOT-10 graphics software is required for full graphics capability. IAC is available by license for a period of 10 years to approved licensees. The licensed program product includes one complete set of supporting documentation. Additional copies of the documentation may be purchased separately. IAC is written in FORTRAN 77 and has been implemented on a DEC VAX series computer operating under VMS. IAC can be executed by multiple concurrent users in batch or interactive mode. The basic central memory requirement is approximately 750KB. IAC includes the executive system, graphics modules, a database, general utilities, and the interfaces to all analysis and controls programs described above. Source code is provided for the control programs ORACLS, SAMSAN, NBOD2, and DISCOS. The following programs are also available from COSMIC a

IAC-1.5 - INTEGRATED ANALYSIS CAPABILITY

NASA Technical Reports Server (NTRS)

Vos, R. G.

1994-01-01

The objective of the Integrated Analysis Capability (IAC) system is to provide a highly effective, interactive analysis tool for the integrated design of large structures. IAC was developed to interface programs from the fields of structures, thermodynamics, controls, and system dynamics with an executive system and a database to yield a highly efficient multi-disciplinary system. Special attention is given to user requirements such as data handling and on-line assistance with operational features, and the ability to add new modules of the user's choice at a future date. IAC contains an executive system, a database, general utilities, interfaces to various engineering programs, and a framework for building interfaces to other programs. IAC has shown itself to be effective in automating data transfer among analysis programs. The IAC system architecture is modular in design. 1) The executive module contains an input command processor, an extensive data management system, and driver code to execute the application modules. 2) Technical modules provide standalone computational capability as well as support for various solution paths or coupled analyses. 3) Graphics and model generation modules are supplied for building and viewing models. 4) Interface modules provide for the required data flow between IAC and other modules. 5) User modules can be arbitrary executable programs or JCL procedures with no pre-defined relationship to IAC. 6) Special purpose modules are included, such as MIMIC (Model Integration via Mesh Interpolation Coefficients), which transforms field values from one model to another; LINK, which simplifies incorporation of user specific modules into IAC modules; and DATAPAC, the National Bureau of Standards statistical analysis package. The IAC database contains structured files which provide a common basis for communication between modules and the executive system, and can contain unstructured files such as NASTRAN checkpoint files, DISCOS plot files, object code, etc. The user can define groups of data and relations between them. A full data manipulation and query system operates with the database. The current interface modules comprise five groups: 1) Structural analysis - IAC contains a NASTRAN interface for standalone analysis or certain structural/control/thermal combinations. IAC provides enhanced structural capabilities for normal modes and static deformation analysis via special DMAP sequences. 2) Thermal analysis - IAC supports finite element and finite difference techniques for steady state or transient analysis. There are interfaces for the NASTRAN thermal analyzer, SINDA/SINFLO, and TRASYS II. 3) System dynamics - A DISCOS interface allows full use of this simulation program for either nonlinear time domain analysis or linear frequency domain analysis. 4) Control analysis - Interfaces for the ORACLS, SAMSAN, NBOD2, and INCA programs allow a wide range of control system analyses and synthesis techniques. 5) Graphics - The graphics packages PLOT and MOSAIC are included in IAC. PLOT generates vector displays of tabular data in the form of curves, charts, correlation tables, etc., while MOSAIC generates color raster displays of either tabular of array type data. Either DI3000 or PLOT-10 graphics software is required for full graphics capability. IAC is available by license for a period of 10 years to approved licensees. The licensed program product includes one complete set of supporting documentation. Additional copies of the documentation may be purchased separately. IAC is written in FORTRAN 77 and has been implemented on a DEC VAX series computer operating under VMS. IAC can be executed by multiple concurrent users in batch or interactive mode. The basic central memory requirement is approximately 750KB. IAC includes the executive system, graphics modules, a database, general utilities, and the interfaces to all analysis and controls programs described above. Source code is provided for the control programs ORACLS, SAMSAN, NBOD2, and DISCOS. The following programs are also available from COSMIC as separate packages: NASTRAN, SINDA/SINFLO, TRASYS II, DISCOS, ORACLS, SAMSAN, NBOD2, and INCA. IAC was developed in 1985.

Thermal-hydraulic analysis of N Reactor graphite and shield cooling system performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Low, J.O.; Schmitt, B.E.

1988-02-01

A series of bounding (worst-case) calculations were performed using a detailed hydrodynamic RELAP5 model of the N Reactor graphite and shield cooling system (GSCS). These calculations were specifically aimed to answer issues raised by the Westinghouse Independent Safety Review (WISR) committee. These questions address the operability of the GSCS during a worst-case degraded-core accident that requires the GDCS to mitigate the consequences of the accident. An accident scenario previously developed was designed as the hydrogen-mitigation design-basis accident (HMDBA). Previous HMDBA heat transfer analysis,, using the TRUMP-BD code, was used to define the thermal boundary conditions that the GSDS may bemore » exposed to. These TRUMP/HMDBA analysis results were used to define the bounding operating conditions of the GSCS during the course of an HMDBA transient. Nominal and degraded GSCS scenarios were investigated using RELAP5 within or at the bounds of the HMDBA transient. 10 refs., 42 figs., 10 tabs.« less

Assessment of saltwater intrusion in southern coastal Broward County, Florida

USGS Publications Warehouse

Merritt, M.L.

1996-01-01

Of the counties in southeastern Florida, Broward County has experienced some of the most severe effects of saltwater intrusion into the surficial Biscayne aquifer because, before 1950, most public water-supply well fields in the county were constructed near the principal early population centers located less than 5 miles from the Atlantic Ocean. The construction of major regional drainage canals in the early 20th century caused a lowering of the water table and a gradual inland movement of the saltwater front toward the well fields. The U.S. Geological Survey began field investigations of saltwater intrusion in the Biscayne aquifer of southeastern Broward County in 1939. As part of the present study, the positions of the saltwater front in 1945, 1969, and 1993 were estimated using chloride concentrations of water samples collected between 1939 and 1994 from various monitoring and exploratory wells. The data indicate that, between 1945 and 1993, the saltwater front has moved as much as 0.5 mile inland in parts of the study area. The position and movement of the saltwater front were simulated numerically to help determine which of the various hydrologic factors and water-management features characterizing the coastal subsurface environment and its alteration by man are of significance in increasing or decreasing the degree of saltwater intrusion. Two representational methods were applied by the selection and use of appropriate model codes. The SHARP code simulates the position of the saltwater front as a sharp interface, which implies that no transition zone (a zone in which a gradational change between freshwater and saltwater occurs) separates freshwater and saltwater. The Subsurface Waste Injection Program (SWIP) code simulates a two-fluid, variable-density system using a convective-diffusion approach that includes a representation of the transition zone that occurs between the freshwater and saltwater bodies. The models were applied to: (1) approximately replicate predevelopment and current positions of the interface in the study area; and (2) study the relative importance of various factors affecting the interface position. The model analyses assumed a conceptual model of uniform easterly flow in the aquifer toward points of offshore discharge to tidewater. Measurements of water-table altitude and the depth to the interface in the study area exhibit an interrelation that differes substantially from the classical Ghyben-Herzberg relation. However, both model codes simulated water-table altitudes and interface positions that were generally consistent with the Ghyben-Herzberg relation but differed substantially from observed data. The simulate interface positions were inland of the known positions, and simulate water-table altitudes were higher than measured ones. The SHARP and SWIP simulations were in general agreement with each other when a low value of longitudinal dispersivity was specified in the SWIP simulation and also for higher values of longitudinal dispersivity when modified dispersion algorithms were used in SWIP that greatly reduced the simulated degree of vertical dispersion. Sensitivity analyses performed using the SHARP code indicated simulation results to be relatively insensitive to a substantial change in the specified slope of the base of the aquifer and moderately sensitive to a 150-percent change in net atmospheric recharge to the aquifer (rainfall minus evapotranspiration). Representing well-field pumping by the City of hallandale had only a minor, localized influence on the simulated regional interface position. Using various cross-sectional grid designs in applications of the SWIP code, near convergence of all lines of equal concentrations in the transition zone was achieved within a simulation time of 10 years. The simulated equilibrium interface location was sensitive to substantial spatial variations in the specified hydraulic conductivity values, but was relatively insensitive to seasonal varying

The Euratom-Rosatom ERCOSAM-SAMARA projects on containment thermal-hydraulics of current and future LWRs for severe accident management

DOE Office of Scientific and Technical Information (OSTI.GOV)

Paladino, D.; Guentay, S.; Andreani, M.

2012-07-01

During a postulated severe accident with core degradation, hydrogen would form in the reactor pressure vessel mainly due to high temperatures zirconium-steam reaction and flow together with steam into the containment where it will mix with the containment atmosphere (steam-air). The hydrogen transport into the containment is a safety concern because it can lead to explosive mixtures through the associated phenomena of condensation, mixing and stratification. The ERCOSAM and SAMARA projects, co-financed by the European Union and the Russia, include various experiments addressing accident scenarios scaled down from existing plant calculations to different thermal-hydraulics facilities (TOSQAN, MISTRA, PANDA, SPOT). Themore » tests sequences aim to investigate hydrogen concentration build-up and stratification during a postulated accident and the effect of the activation of Severe Accident Management systems (SAMs), e.g. sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Analytical activities, performed by the project participants, are an essential component of the projects, as they aim to improve and validate various computational methods. They accompany the projects in the various phases; plant calculations, scaling to generic containment and to the different facilities, planning pre-test and post-test simulations are performed. Code benchmark activities on the basis of conceptual near full scale HYMIX facility will finally provide a further opportunity to evaluate the applicability of the various methods to the study of scaling issues. (authors)« less

Thermal-hydraulic simulation of natural convection decay heat removal in the High Flux Isotope Reactor (HFIR) using RELAP5 and TEMPEST: Part 2, Interpretation and validation of results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ruggles, A.E.; Morris, D.G.

The RELAP5/MOD2 code was used to predict the thermal-hydraulic behavior of the HFIR core during decay heat removal through boiling natural circulation. The low system pressure and low mass flux values associated with boiling natural circulation are far from conditions for which RELAP5 is well exercised. Therefore, some simple hand calculations are used herein to establish the physics of the results. The interpretation and validation effort is divided between the time average flow conditions and the time varying flow conditions. The time average flow conditions are evaluated using a lumped parameter model and heat balance. The Martinelli-Nelson correlations are usedmore » to model the two-phase pressure drop and void fraction vs flow quality relationship within the core region. Systems of parallel channels are susceptible to both density wave oscillations and pressure drop oscillations. Periodic variations in the mass flux and exit flow quality of individual core channels are predicted by RELAP5. These oscillations are consistent with those observed experimentally and are of the density wave type. The impact of the time varying flow properties on local wall superheat is bounded herein. The conditions necessary for Ledinegg flow excursions are identified. These conditions do not fall within the envelope of decay heat levels relevant to HFIR in boiling natural circulation. 14 refs., 5 figs., 1 tab.« less

Theoretical investigation of the thermal hydraulic behaviour of a slab-type liquid metal target

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dury, T.V.; Smith, B.L.

1996-06-01

The thermal hydraulics codes CFDS-FLOW3D and ASTEC have been used to simulate a slabtype design of ESS spallation target. This design is single-skinned, and of tapering form (in the beam direction), with rounded sides in a cross-section through a plane normal to the beam. The coolant fluid used is mercury, under forced circulation, with an inlet temperature of 180{degrees}C. The goal of these computer studies was to understand the behaviour of the coolant flow, and hence to arrive at a design which optimises the heat extraction for a given beam power - in the sense of: (1) minimising the peakmore » local fluid temperature within the target, (2) maintaining an acceptable temperature level and distribution over and through the target outer wall, (3) keeping the overall fluid pressure loss through the complete target to a minimum, (4) staying within the physical limits of overall size required, particularly in the region of primary spallation. Two- and three-dimensional models have been used, with different arrangements and design of internal baffles, and different coolant flow distributions at the target inlet. Nominal total inlet mass flow was 245 kg/s, and a heat deposition profile used which was based on the proton beam energy distribution. This gave a nominal total heat load of 3.23 MW - of which 8.2kW were deposited in the window steel.« less

Analysis of steam generator tube rupture transients with single failure

DOE Office of Scientific and Technical Information (OSTI.GOV)

Trambauer, K.

The Gesellschaft fuer Reaktorsicherheit is engaged in the collection and evaluation of light water reactor operating experience as well as analyses for the risk study of the pressurized water reactor (PWR). Within these activities, thermohydraulic calculations have been performed to show the influence of different boundary conditions and disturbances on the steam generator tube rupture (SGTR) transients. The analyses of these calculations have focused on the measures and systems needed to cope with an SGTR. The reference plant for this analysis is a 1300-MW(e) PWR of Kraftwerk Union design with four loops, each containing a U-tube steam generator (SG) andmore » a reactor cooling pump (RCP). The thermal-hydraulic code DRUFAN-02 was used for the transient calculations.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Boyack, B.E.; Steiner, J.L.; Harmony, S.C.

The PIUS advanced reactor is a 640-MWe pressurized water reactor developed by Asea Brown Boveri (ABB). A unique feature of the PIUS concept is the absence of mechanical control and shutdown rods. Reactivity is normally controlled by coolant boron concentration and the temperature of the moderator coolant. ABB submitted the PIUS design to the US Nuclear Regulatory Commission (NRC) for preapplication review, and Los Alamos supported the NRC`s review effort. Baseline analyses of small-break initiators at two locations were performed with the system neutronic and thermal-hydraulic analysis code TRAC-PF1/MOD2. In addition, sensitivity studies were performed to explore the robustness ofmore » the PIUS concept to severe off-normal conditions having a very low probability of occurrence.« less

Software Tools for Stochastic Simulations of Turbulence

DTIC Science & Technology

2015-08-28

client interface to FTI. Specefic client programs using this interface include the weather forecasting code WRF ; the high energy physics code, FLASH...client programs using this interface include the weather forecasting code WRF ; the high energy physics code, FLASH; and two locally constructed fluid...45 4.4.2.2 FLASH . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 45 4.4.2.3 WRF

Modeling of Non-Isothermal Cryogenic Fluid Sloshing

NASA Technical Reports Server (NTRS)

Agui, Juan H.; Moder, Jeffrey P.

2015-01-01

A computational fluid dynamic model was used to simulate the thermal destratification in an upright self-pressurized cryostat approximately half-filled with liquid nitrogen and subjected to forced sinusoidal lateral shaking. A full three-dimensional computational grid was used to model the tank dynamics, fluid flow and thermodynamics using the ANSYS Fluent code. A non-inertial grid was used which required the addition of momentum and energy source terms to account for the inertial forces, energy transfer and wall reaction forces produced by the shaken tank. The kinetics-based Schrage mass transfer model provided the interfacial mass transfer due to evaporation and condensation at the sloshing interface. The dynamic behavior of the sloshing interface, its amplitude and transition to different wave modes, provided insight into the fluid process at the interface. The tank pressure evolution and temperature profiles compared relatively well with the shaken cryostat experimental test data provided by the Centre National D'Etudes Spatiales.

Towards a Consolidated Approach for the Assessment of Evaluation Models of Nuclear Power Reactors

DOE PAGES

Epiney, A.; Canepa, S.; Zerkak, O.; ...

2016-11-02

The STARS project at the Paul Scherrer Institut (PSI) has adopted the TRACE thermal-hydraulic (T-H) code for best-estimate system transient simulations of the Swiss Light Water Reactors (LWRs). For analyses involving interactions between system and core, a coupling of TRACE with the SIMULATE-3K (S3K) LWR core simulator has also been developed. In this configuration, the TRACE code and associated nuclear power reactor simulation models play a central role to achieve a comprehensive safety analysis capability. Thus, efforts have now been undertaken to consolidate the validation strategy by implementing a more rigorous and structured assessment approach for TRACE applications involving eithermore » only system T-H evaluations or requiring interfaces to e.g. detailed core or fuel behavior models. The first part of this paper presents the preliminary concepts of this validation strategy. The principle is to systematically track the evolution of a given set of predicted physical Quantities of Interest (QoIs) over a multidimensional parametric space where each of the dimensions represent the evolution of specific analysis aspects, including e.g. code version, transient specific simulation methodology and model "nodalisation". If properly set up, such environment should provide code developers and code users with persistent (less affected by user effect) and quantified information (sensitivity of QoIs) on the applicability of a simulation scheme (codes, input models, methodology) for steady state and transient analysis of full LWR systems. Through this, for each given transient/accident, critical paths of the validation process can be identified that could then translate into defining reference schemes to be applied for downstream predictive simulations. In order to illustrate this approach, the second part of this paper presents a first application of this validation strategy to an inadvertent blowdown event that occurred in a Swiss BWR/6. The transient was initiated by the spurious actuation of the Automatic Depressurization System (ADS). The validation approach progresses through a number of dimensions here: First, the same BWR system simulation model is assessed for different versions of the TRACE code, up to the most recent one. The second dimension is the "nodalisation" dimension, where changes to the input model are assessed. The third dimension is the "methodology" dimension. In this case imposed power and an updated TRACE core model are investigated. For each step in each validation dimension, a common set of QoIs are investigated. For the steady-state results, these include fuel temperatures distributions. For the transient part of the present study, the evaluated QoIs include the system pressure evolution and water carry-over into the steam line.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hale, Richard Edward; Cetiner, Sacit M.; Fugate, David L.

The Small Modular Reactor (SMR) Dynamic System Modeling Tool project is in the third year of development. The project is designed to support collaborative modeling and study of various advanced SMR (non-light water cooled) concepts, including the use of multiple coupled reactors at a single site. The objective of the project is to provide a common simulation environment and baseline modeling resources to facilitate rapid development of dynamic advanced reactor SMR models, ensure consistency among research products within the Instrumentation, Controls, and Human-Machine Interface (ICHMI) technical area, and leverage cross-cutting capabilities while minimizing duplication of effort. The combined simulation environmentmore » and suite of models are identified as the Modular Dynamic SIMulation (MoDSIM) tool. The critical elements of this effort include (1) defining a standardized, common simulation environment that can be applied throughout the program, (2) developing a library of baseline component modules that can be assembled into full plant models using existing geometry and thermal-hydraulic data, (3) defining modeling conventions for interconnecting component models, and (4) establishing user interfaces and support tools to facilitate simulation development (i.e., configuration and parameterization), execution, and results display and capture.« less

Optimised Iteration in Coupled Monte Carlo - Thermal-Hydraulics Calculations

NASA Astrophysics Data System (ADS)

Hoogenboom, J. Eduard; Dufek, Jan

2014-06-01

This paper describes an optimised iteration scheme for the number of neutron histories and the relaxation factor in successive iterations of coupled Monte Carlo and thermal-hydraulic reactor calculations based on the stochastic iteration method. The scheme results in an increasing number of neutron histories for the Monte Carlo calculation in successive iteration steps and a decreasing relaxation factor for the spatial power distribution to be used as input to the thermal-hydraulics calculation. The theoretical basis is discussed in detail and practical consequences of the scheme are shown, among which a nearly linear increase per iteration of the number of cycles in the Monte Carlo calculation. The scheme is demonstrated for a full PWR type fuel assembly. Results are shown for the axial power distribution during several iteration steps. A few alternative iteration method are also tested and it is concluded that the presented iteration method is near optimal.

Instability in Immiscible Fluids Displacement from Cracks and Porous Samples

NASA Astrophysics Data System (ADS)

Smirnov, N. N.; Nikitin, V. F.; Ivashnyov, O. E.

2002-01-01

problems of terrestrial engineering and technology. Surface tension affected flows in porous media could be much better understood in microgravity studies eliminating the masking effects of gravity. Saffman-Taylor instability of the interface could bring to formation and growth of "fingers" of gas penetrating the bulk fluid. The growth of fingers and their further coalescence could not be described by the linear analysis. Growth of fingers causes irregularity of the mixing zone. The tangential velocity difference on the interface separating fluids of different densities and viscousities could bring to a Kelvin-Helmholtz instability resulting in "diffusion of fingers" partial regularization of the displacement mixing zone. Thus combination of the two effects would govern the flow in the displacement process. fracture under a pressure differential displacing the high viscosity residual fracturing fluid. There are inherent instability and scalability problems associated with viscous fingering that play a key role in the cleanup procedure. Entrapment of residual fracturing fluid by the gas flow lowers down the quality of a fracture treatment leaving most of fluid in the hydraulic fracture thus decreasing the production rate. The gravity effects could play essential role in vertical hydraulic fractures as the problem is scale dependent. displacement of viscous fluid by a less viscous one in a two-dimensional channel with vertical breaks, and to determine characteristic size of entrapment zones. Extensive direct numerical simulations allow to investigate the sensitivity of the displacement process to variation of values of the main governing parameters. were found for the two limiting cases: infinitely wide cell, and narrow cell with an infinitely small gap between the finger and the side walls. governing parameters. The obtained solutions allowed to explain the physical meaning of the exiting empirical criteria for the beginning of viscous fingering and the growth of a number of fingers in the cell, and allowed us to make some additional suggestions for the cleanup procedure. depending on the resident fluid properties, for which the displacement still remains stable. viscous one were carried out. Validation of the code was performed by comparing the results of model problems simulations with the existing solutions published in literature. Being in a good agreement with the previously obtained results, nevertheless, the developed code is an advanced one. While the existing codes could operate with linear equations and regular geometry and initial disturbances only, the new code permits taking into account non-linear effects as well. characterizing the quality of displacement. The functional dependence of the dimensionless criteria on the values of governing parameters needs further investigations. Services, an international company in the oil and gas industry.

Mixing, Combustion, and Other Interface Dominated Flows; Paragraphs 3.2.1 A, B, C and 3.2.2 A

DTIC Science & Technology

2014-04-09

Condensed Matter Physics , (12 2010): 43401. doi: H. Lim, Y. Yu, J. Glimm, X. L. Li, D.H. Sharp. Subgrid Models for Mass and Thermal Diffusion in...zone and a series of radial cracks in solid plates hit by high velocity projectiles). • Only 2D dimensional models • Serial codes for running on single ...exter- nal parallel packages TAO and Global Arrays, developed within DOE high performance computing initiatives. A Schwartz-type overlapping domain

TRACE/PARCS analysis of the OECD/NEA Oskarshamn-2 BWR stability benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kozlowski, T.; Downar, T.; Xu, Y.

2012-07-01

On February 25, 1999, the Oskarshamn-2 NPP experienced a stability event which culminated in diverging power oscillations with a decay ratio of about 1.4. The event was successfully modeled by the TRACE/PARCS coupled code system, and further analysis of the event is described in this paper. The results show very good agreement with the plant data, capturing the entire behavior of the transient including the onset of instability, growth of the oscillations (decay ratio) and oscillation frequency. This provides confidence in the prediction of other parameters which are not available from the plant records. The event provides coupled code validationmore » for a challenging BWR stability event, which involves the accurate simulation of neutron kinetics (NK), thermal-hydraulics (TH), and TH/NK. coupling. The success of this work has demonstrated the ability of the 3-D coupled systems code TRACE/PARCS to capture the complex behavior of BWR stability events. The problem was released as an international OECD/NEA benchmark, and it is the first benchmark based on measured plant data for a stability event with a DR greater than one. Interested participants are invited to contact authors for more information. (authors)« less

Steady state and LOCA analysis of Kartini reactor using RELAP5/SCDAP code: The role of passive system

NASA Astrophysics Data System (ADS)

Antariksawan, Anhar R.; Wahyono, Puradwi I.; Taxwim

2018-02-01

Safety is the priority for nuclear installations, including research reactors. On the other hand, many studies have been done to validate the applicability of nuclear power plant based best estimate computer codes to the research reactor. This study aims to assess the applicability of the RELAP5/SCDAP code to Kartini research reactor. The model development, steady state and transient due to LOCA calculations have been conducted by using RELAP5/SCDAP. The calculation results are compared with available measurements data from Kartini research reactor. The results show that the RELAP5/SCDAP model steady state calculation agrees quite well with the available measurement data. While, in the case of LOCA transient simulations, the model could result in reasonable physical phenomena during the transient showing the characteristics and performances of the reactor against the LOCA transient. The role of siphon breaker hole and natural circulation in the reactor tank as passive system was important to keep reactor in safe condition. It concludes that the RELAP/SCDAP could be use as one of the tool to analyse the thermal-hydraulic safety of Kartini reactor. However, further assessment to improve the model is still needed.

Simulation of Targets Feeding Pipe Rupture in Wendelstein 7-X Facility Using RELAP5 and COCOSYS Codes

NASA Astrophysics Data System (ADS)

Kaliatka, T.; Povilaitis, M.; Kaliatka, A.; Urbonavicius, E.

2012-10-01

Wendelstein nuclear fusion device W7-X is a stellarator type experimental device, developed by Max Planck Institute of plasma physics. Rupture of one of the 40 mm inner diameter coolant pipes providing water for the divertor targets during the "baking" regime of the facility operation is considered to be the most severe accident in terms of the plasma vessel pressurization. "Baking" regime is the regime of the facility operation during which plasma vessel structures are heated to the temperature acceptable for the plasma ignition in the vessel. This paper presents the model of W7-X cooling system (pumps, valves, pipes, hydro-accumulators, and heat exchangers), developed using thermal-hydraulic state-of-the-art RELAP5 Mod3.3 code, and model of plasma vessel, developed by employing the lumped-parameter code COCOSYS. Using both models the numerical simulation of processes in W7-X cooling system and plasma vessel has been performed. The results of simulation showed, that the automatic valve closure time 1 s is the most acceptable (no water hammer effect occurs) and selected area of the burst disk is sufficient to prevent pressure in the plasma vessel.

STAR-CCM+ Verification and Validation Plan

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pointer, William David

2016-09-30

The commercial Computational Fluid Dynamics (CFD) code STAR-CCM+ provides general purpose finite volume method solutions for fluid dynamics and energy transport. This document defines plans for verification and validation (V&V) of the base code and models implemented within the code by the Consortium for Advanced Simulation of Light water reactors (CASL). The software quality assurance activities described herein are port of the overall software life cycle defined in the CASL Software Quality Assurance (SQA) Plan [Sieger, 2015]. STAR-CCM+ serves as the principal foundation for development of an advanced predictive multi-phase boiling simulation capability within CASL. The CASL Thermal Hydraulics Methodsmore » (THM) team develops advanced closure models required to describe the subgrid-resolution behavior of secondary fluids or fluid phases in multiphase boiling flows within the Eulerian-Eulerian framework of the code. These include wall heat partitioning models that describe the formation of vapor on the surface and the forces the define bubble/droplet dynamic motion. The CASL models are implemented as user coding or field functions within the general framework of the code. This report defines procedures and requirements for V&V of the multi-phase CFD capability developed by CASL THM. Results of V&V evaluations will be documented in a separate STAR-CCM+ V&V assessment report. This report is expected to be a living document and will be updated as additional validation cases are identified and adopted as part of the CASL THM V&V suite.« less

Travel-time-based thermal tracer tomography

NASA Astrophysics Data System (ADS)

Somogyvári, Márk; Bayer, Peter; Brauchler, Ralf

2016-05-01

Active thermal tracer testing is a technique to get information about the flow and transport properties of an aquifer. In this paper we propose an innovative methodology using active thermal tracers in a tomographic setup to reconstruct cross-well hydraulic conductivity profiles. This is facilitated by assuming that the propagation of the injected thermal tracer is mainly controlled by advection. To reduce the effects of density and viscosity changes and thermal diffusion, early-time diagnostics are used and specific travel times of the tracer breakthrough curves are extracted. These travel times are inverted with an eikonal solver using the staggered grid method to reduce constraints from the pre-defined grid geometry and to improve the resolution. Finally, non-reliable pixels are removed from the derived hydraulic conductivity tomograms. The method is applied to successfully reconstruct cross-well profiles as well as a 3-D block of a high-resolution fluvio-aeolian aquifer analog data set. Sensitivity analysis reveals a negligible role of the injection temperature, but more attention has to be drawn to other technical parameters such as the injection rate. This is investigated in more detail through model-based testing using diverse hydraulic and thermal conditions in order to delineate the feasible range of applications for the new tomographic approach.

Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Arndt, S.A.

1997-07-01

The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for codemore » use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.« less

Multi-Material ALE with AMR for Modeling Hot Plasmas and Cold Fragmenting Materials

NASA Astrophysics Data System (ADS)

Alice, Koniges; Nathan, Masters; Aaron, Fisher; David, Eder; Wangyi, Liu; Robert, Anderson; David, Benson; Andrea, Bertozzi

2015-02-01

We have developed a new 3D multi-physics multi-material code, ALE-AMR, which combines Arbitrary Lagrangian Eulerian (ALE) hydrodynamics with Adaptive Mesh Refinement (AMR) to connect the continuum to the microstructural regimes. The code is unique in its ability to model hot radiating plasmas and cold fragmenting solids. New numerical techniques were developed for many of the physics packages to work efficiently on a dynamically moving and adapting mesh. We use interface reconstruction based on volume fractions of the material components within mixed zones and reconstruct interfaces as needed. This interface reconstruction model is also used for void coalescence and fragmentation. A flexible strength/failure framework allows for pluggable material models, which may require material history arrays to determine the level of accumulated damage or the evolving yield stress in J2 plasticity models. For some applications laser rays are propagating through a virtual composite mesh consisting of the finest resolution representation of the modeled space. A new 2nd order accurate diffusion solver has been implemented for the thermal conduction and radiation transport packages. One application area is the modeling of laser/target effects including debris/shrapnel generation. Other application areas include warm dense matter, EUV lithography, and material wall interactions for fusion devices.

Enhancements to the SHARP Build System and NEK5000 Coupling

DOE Office of Scientific and Technical Information (OSTI.GOV)

McCaskey, Alex; Bennett, Andrew R.; Billings, Jay Jay

The SHARP project for the Department of Energy's Nuclear Energy Advanced Modeling and Simulation (NEAMS) program provides a multiphysics framework for coupled simulations of advanced nuclear reactor designs. It provides an overall coupling environment that utilizes custom interfaces to couple existing physics codes through a common spatial decomposition and unique solution transfer component. As of this writing, SHARP couples neutronics, thermal hydraulics, and structural mechanics using PROTEUS, Nek5000, and Diablo respectively. This report details two primary SHARP improvements regarding the Nek5000 and Diablo individual physics codes: (1) an improved Nek5000 coupling interface that lets SHARP achieve a vast increase inmore » overall solution accuracy by manipulating the structure of the internal Nek5000 spatial mesh, and (2) the capability to seamlessly couple structural mechanics calculations into the framework through improvements to the SHARP build system. The Nek5000 coupling interface now uses a barycentric Lagrange interpolation method that takes the vertex-based power and density computed from the PROTEUS neutronics solver and maps it to the user-specified, general-order Nek5000 spectral element mesh. Before this work, SHARP handled this vertex-based solution transfer in an averaging-based manner. SHARP users can now achieve higher levels of accuracy by specifying any arbitrary Nek5000 spectral mesh order. This improvement takes the average percentage error between the PROTEUS power solution and the Nek5000 interpolated result down drastically from over 23 % to just above 2 %, and maintains the correct power profile. We have integrated Diablo into the SHARP build system to facilitate the future coupling of structural mechanics calculations into SHARP. Previously, simulations involving Diablo were done in an iterative manner, requiring a large amount manual work, and left only as a task for advanced users. This report will detail a new Diablo build system that was implemented using GNU Autotools, mirroring much of the current SHARP build system, and easing the use of structural mechanics calculations for end-users of the SHARP multiphysics framework. It lets users easily build and use Diablo as a stand-alone simulation, as well as fully couple with the other SHARP physics modules. The top-level SHARP build system was modified to allow Diablo to hook in directly. New dependency handlers were implemented to let SHARP users easily build the framework with these new simulation capabilities. The remainder of this report will describe this work in full, with a detailed discussion of the overall design philosophy of SHARP, the new solution interpolation method introduced, and the Diablo integration work. We will conclude with a discussion of possible future SHARP improvements that will serve to increase solution accuracy and framework capability.« less

Development of Web Interfaces for Analysis Codes

NASA Astrophysics Data System (ADS)

Emoto, M.; Watanabe, T.; Funaba, H.; Murakami, S.; Nagayama, Y.; Kawahata, K.

Several codes have been developed to analyze plasma physics. However, most of them are developed to run on supercomputers. Therefore, users who typically use personal computers (PCs) find it difficult to use these codes. In order to facilitate the widespread use of these codes, a user-friendly interface is required. The authors propose Web interfaces for these codes. To demonstrate the usefulness of this approach, the authors developed Web interfaces for two analysis codes. One of them is for FIT developed by Murakami. This code is used to analyze the NBI heat deposition, etc. Because it requires electron density profiles, electron temperatures, and ion temperatures as polynomial expressions, those unfamiliar with the experiments find it difficult to use this code, especially visitors from other institutes. The second one is for visualizing the lines of force in the LHD (large helical device) developed by Watanabe. This code is used to analyze the interference caused by the lines of force resulting from the various structures installed in the vacuum vessel of the LHD. This code runs on PCs; however, it requires that the necessary parameters be edited manually. Using these Web interfaces, users can execute these codes interactively.

Hydro-mechanical coupled simulation of hydraulic fracturing using the eXtended Finite Element Method (XFEM)

NASA Astrophysics Data System (ADS)

Youn, Dong Joon

This thesis presents the development and validation of an advanced hydro-mechanical coupled finite element program analyzing hydraulic fracture propagation within unconventional hydrocarbon formations under various conditions. The realistic modeling of hydraulic fracturing is necessarily required to improve the understanding and efficiency of the stimulation technique. Such modeling remains highly challenging, however, due to factors including the complexity of fracture propagation mechanisms, the coupled behavior of fracture displacement and fluid pressure, the interactions between pre-existing natural and initiated hydraulic fractures and the formation heterogeneity of the target reservoir. In this research, an eXtended Finite Element Method (XFEM) scheme is developed allowing for representation of single or multiple fracture propagations without any need for re-meshing. Also, the coupled flows through the fracture are considered in the program to account for their influence on stresses and deformations along the hydraulic fracture. In this research, a sequential coupling scheme is applied to estimate fracture aperture and fluid pressure with the XFEM. Later, the coupled XFEM program is used to estimate wellbore bottomhole pressure during fracture propagation, and the pressure variations are analyzed to determine the geometry and performance of the hydraulic fracturing as pressure leak-off test. Finally, material heterogeneity is included into the XFEM program to check the effect of random formation property distributions to the hydraulic fracture geometry. Random field theory is used to create the random realization of the material heterogeneity with the consideration of mean, standard deviation, and property correlation length. These analyses lead to probabilistic information on the response of unconventional reservoirs and offer a more scientific approach regarding risk management for the unconventional reservoir stimulation. The new stochastic approach combining XFEM and random field is named as eXtended Random Finite Element Method (XRFEM). All the numerical analysis codes in this thesis are written in Fortran 2003, and these codes are applicable as a series of sub-modules within a suite of finite element codes developed by Smith and Griffiths (2004).

A Large-scale Finite Element Model on Micromechanical Damage and Failure of Carbon Fiber/Epoxy Composites Including Thermal Residual Stress

NASA Astrophysics Data System (ADS)

Liu, P. F.; Li, X. K.

2018-06-01

The purpose of this paper is to study micromechanical progressive failure properties of carbon fiber/epoxy composites with thermal residual stress by finite element analysis (FEA). Composite microstructures with hexagonal fiber distribution are used for the representative volume element (RVE), where an initial fiber breakage is assumed. Fiber breakage with random fiber strength is predicted using Monte Carlo simulation, progressive matrix damage is predicted by proposing a continuum damage mechanics model and interface failure is simulated using Xu and Needleman's cohesive model. Temperature dependent thermal expansion coefficients for epoxy matrix are used. FEA by developing numerical codes using ANSYS finite element software is divided into two steps: 1. Thermal residual stresses due to mismatch between fiber and matrix are calculated; 2. Longitudinal tensile load is further exerted on the RVE to perform progressive failure analysis of carbon fiber/epoxy composites. Numerical convergence is solved by introducing the viscous damping effect properly. The extended Mori-Tanaka method that considers interface debonding is used to get homogenized mechanical responses of composites. Three main results by FEA are obtained: 1. the real-time matrix cracking, fiber breakage and interface debonding with increasing tensile strain is simulated. 2. the stress concentration coefficients on neighbouring fibers near the initial broken fiber and the axial fiber stress distribution along the broken fiber are predicted, compared with the results using the global and local load-sharing models based on the shear-lag theory. 3. the tensile strength of composite by FEA is compared with those by the shear-lag theory and experiments. Finally, the tensile stress-strain curve of composites by FEA is applied to the progressive failure analysis of composite pressure vessel.

A Large-scale Finite Element Model on Micromechanical Damage and Failure of Carbon Fiber/Epoxy Composites Including Thermal Residual Stress

NASA Astrophysics Data System (ADS)

Liu, P. F.; Li, X. K.

2017-09-01

The purpose of this paper is to study micromechanical progressive failure properties of carbon fiber/epoxy composites with thermal residual stress by finite element analysis (FEA). Composite microstructures with hexagonal fiber distribution are used for the representative volume element (RVE), where an initial fiber breakage is assumed. Fiber breakage with random fiber strength is predicted using Monte Carlo simulation, progressive matrix damage is predicted by proposing a continuum damage mechanics model and interface failure is simulated using Xu and Needleman's cohesive model. Temperature dependent thermal expansion coefficients for epoxy matrix are used. FEA by developing numerical codes using ANSYS finite element software is divided into two steps: 1. Thermal residual stresses due to mismatch between fiber and matrix are calculated; 2. Longitudinal tensile load is further exerted on the RVE to perform progressive failure analysis of carbon fiber/epoxy composites. Numerical convergence is solved by introducing the viscous damping effect properly. The extended Mori-Tanaka method that considers interface debonding is used to get homogenized mechanical responses of composites. Three main results by FEA are obtained: 1. the real-time matrix cracking, fiber breakage and interface debonding with increasing tensile strain is simulated. 2. the stress concentration coefficients on neighbouring fibers near the initial broken fiber and the axial fiber stress distribution along the broken fiber are predicted, compared with the results using the global and local load-sharing models based on the shear-lag theory. 3. the tensile strength of composite by FEA is compared with those by the shear-lag theory and experiments. Finally, the tensile stress-strain curve of composites by FEA is applied to the progressive failure analysis of composite pressure vessel.

Review of CTF s Fuel Rod Modeling Using FRAPCON-4.0 s Centerline Temperature Predictions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Toptan, Aysenur; Salko, Robert K; Avramova, Maria

Coolant Boiling in Rod Arrays Two Fluid (COBRA-TF), or CTF1 [1], is a nuclear thermal hydraulic subchannel code used throughout academia and industry. CTF s fuel rod modeling is originally developed for VIPRE code [2]. Its methodology is based on GAPCON [3] and FRAP [4] fuel performance codes, and material properties are included from MATPRO handbook [5]. This work focuses on review of CTF s fuel rod modeling to address shortcomings in CTF s temperature predictions. CTF is compared to FRAPCON which is U.S. NRC s steady-state fuel performance code for light-water reactor fuel rods. FRAPCON calculates the changes inmore » fuel rod variables and temperatures including the eects of cladding hoop strain, cladding oxidation, hydriding, fuel irradiation swelling, densification, fission gas release and rod internal gas pressure. It uses fuel, clad and gap material properties from MATPRO. Additionally, it has its own models for fission gas release, cladding corrosion and cladding hydrogen pickup. It allows finite dierence or finite element approaches for its mechanical model. In this study, FRAPCON-4.0 [6] is used as a reference fuel performance code. In comparison, Halden Reactor Data for IFA432 Rod 1 and Rod 3. CTF simulations are performed in two ways; informing CTF with gap conductance value from FRAPCON, and using CTF s dynamic gap conductance model. First case is chosen to show temperature is predicted correctly with CTF s models for thermal and cladding conductivities once gap conductance is provided. Latter is to review CTF s dynamic gap conductance model. These Halden test cases are selected to be representative of cases with and without any physical contact between fuel-pellet and clad while reviewing functionality of CTF s dynamic gap conductance model. Improving the CTF s dynamic gap conductance model will allow prediction of fuel and cladding thermo-mechanical behavior under irradiation, and better temperature feedbacks from CTF in transient calculations.« less

Groundwater exchanges near a channelized versus unmodified stream mouth discharging to a subalpine lake

USGS Publications Warehouse

Constantz, James; Naranjo, Ramon C.; Niswonger, Richard G.; Allander, Kip K.; Neilson, B.; Rosenberry, Donald O.; Smith, David W.; Rosecrans, C.; Stonestrom, David A.

2016-01-01

The terminus of a stream flowing into a larger river, pond, lake, or reservoir is referred to as the stream-mouth reach or simply the stream mouth. The terminus is often characterized by rapidly changing thermal and hydraulic conditions that result in abrupt shifts in surface water/groundwater (sw/gw) exchange patterns, creating the potential for unique biogeochemical processes and ecosystems. Worldwide shoreline development is changing stream-lake interfaces through channelization of stream mouths, i.e., channel straightening and bank stabilization to prevent natural meandering at the shoreline. In the central Sierra Nevada (USA), Lake Tahoe's shoreline has an abundance of both “unmodified” (i.e., not engineered though potentially impacted by broader watershed engineering) and channelized stream mouths. Two representative stream mouths along the lake's north shore, one channelized and one unmodified, were selected to compare and contrast water and heat exchanges. Hydraulic and thermal properties were monitored during separate campaigns in September 2012 and 2013 and sw/gw exchanges were estimated within the stream mouth-shoreline continuum. Heat-flow and water-flow patterns indicated clear differences in the channelized versus the unmodified stream mouth. For the channelized stream mouth, relatively modulated, cool-temperature, low-velocity longitudinal streambed flows discharged offshore beneath warmer buoyant lakeshore water. In contrast, a seasonal barrier bar formed across the unmodified stream mouth, creating higher-velocity subsurface flow paths and higher diurnal temperature variations relative to shoreline water. As a consequence, channelization altered sw/gw exchanges potentially altering biogeochemical processing and ecological systems in and near the stream mouth.

Investigation of wellbore microannulus permeability under stress via experimental wellbore mock-up and finite element modeling

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gomez, Steven P.; Sobolik, Steve R.; Matteo, Edward N.

This research aims to describe the microannulus region of the cement sheath-steel casing interface in terms of its compressibility and permeability. Here, a wellbore system mock-up was used for lab-scale testing, and was subjected to confining and casing pressures in a pressure vessel while measuring gas flow along the specimen’s axis. The flow was interpreted as the hydraulic aperture of the microannuli. Numerical joint models were used to calculate stress and displacement conditions of the microannulus region, where the mechanical stiffness and hydraulic aperture were altered in response to the imposed stress state and displacement across the joint interface.

Investigation of wellbore microannulus permeability under stress via experimental wellbore mock-up and finite element modeling

DOE PAGES

Gomez, Steven P.; Sobolik, Steve R.; Matteo, Edward N.; ...

2016-11-16

This research aims to describe the microannulus region of the cement sheath-steel casing interface in terms of its compressibility and permeability. Here, a wellbore system mock-up was used for lab-scale testing, and was subjected to confining and casing pressures in a pressure vessel while measuring gas flow along the specimen’s axis. The flow was interpreted as the hydraulic aperture of the microannuli. Numerical joint models were used to calculate stress and displacement conditions of the microannulus region, where the mechanical stiffness and hydraulic aperture were altered in response to the imposed stress state and displacement across the joint interface.

Deconstructing Temperature Gradients across Fluid Interfaces: The Structural Origin of the Thermal Resistance of Liquid-Vapor Interfaces

NASA Astrophysics Data System (ADS)

Muscatello, Jordan; Chacón, Enrique; Tarazona, Pedro; Bresme, Fernando

2017-07-01

The interfacial thermal resistance determines condensation-evaporation processes and thermal transport across material-fluid interfaces. Despite its importance in transport processes, the interfacial structure responsible for the thermal resistance is still unknown. By combining nonequilibrium molecular dynamics simulations and interfacial analyses that remove the interfacial thermal fluctuations we show that the thermal resistance of liquid-vapor interfaces is connected to a low density fluid layer that is adsorbed at the liquid surface. This thermal resistance layer (TRL) defines the boundary where the thermal transport mechanism changes from that of gases (ballistic) to that characteristic of dense liquids, dominated by frequent particle collisions involving very short mean free paths. We show that the thermal conductance is proportional to the number of atoms adsorbed in the TRL, and hence we explain the structural origin of the thermal resistance in liquid-vapor interfaces.

Phase Change Material Thermal Power Generator

NASA Technical Reports Server (NTRS)

Jones, Jack A. (Inventor); Chao, Yi (Inventor); Valdez, Thomas I. (Inventor)

2014-01-01

An energy producing device, for example a submersible vehicle for descending or ascending to different depths within water or ocean, is disclosed. The vehicle comprises a temperature-responsive material to which a hydraulic fluid is associated. A pressurized storage compartment stores the fluid as soon as the temperature-responsive material changes density. The storage compartment is connected with a hydraulic motor, and a valve allows fluid passage from the storage compartment to the hydraulic motor. An energy storage component, e.g. a battery, is connected with the hydraulic motor and is charged by the hydraulic motor when the hydraulic fluid passes through the hydraulic motor. Upon passage in the hydraulic motor, the fluid is stored in a further storage compartment and is then sent back to the area of the temperature-responsive material.

Phase change material thermal power generator

NASA Technical Reports Server (NTRS)

Jones, Jack A. (Inventor); Chao, Yi (Inventor); Valdez, Thomas I. (Inventor)

2011-01-01

An energy producing device, for example a submersible vehicle for descending or ascending to different depths within water or ocean, is disclosed. The vehicle comprises a temperature-responsive material to which a hydraulic fluid is associated. A pressurized storage compartment stores the fluid as soon as the temperature-responsive material changes density. The storage compartment is connected with a hydraulic motor, and a valve allows fluid passage from the storage compartment to the hydraulic motor. An energy storage component, e.g. a battery, is connected with the hydraulic motor and is charged by the hydraulic motor when the hydraulic fluid passes through the hydraulic motor. Upon passage in the hydraulic motor, the fluid is stored in a further storage compartment and is then sent back to the area of the temperature-responsive material.

RACLETTE: a model for evaluating the thermal response of plasma facing components to slow high power plasma transients. Part I: Theory and description of model capabilities

NASA Astrophysics Data System (ADS)

Raffray, A. René; Federici, Gianfranco

1997-04-01

RACLETTE (Rate Analysis Code for pLasma Energy Transfer Transient Evaluation), a comprehensive but relatively simple and versatile model, was developed to help in the design analysis of plasma facing components (PFCs) under 'slow' high power transients, such as those associated with plasma vertical displacement events. The model includes all the key surface heat transfer processes such as evaporation, melting, and radiation, and their interaction with the PFC block thermal response and the coolant behaviour. This paper represents part I of two sister and complementary papers. It covers the model description, calibration and validation, and presents a number of parametric analyses shedding light on and identifying trends in the PFC armour block response to high plasma energy deposition transients. Parameters investigated include the plasma energy density and deposition time, the armour thickness and the presence of vapour shielding effects. Part II of the paper focuses on specific design analyses of ITER plasma facing components (divertor, limiter, primary first wall and baffle), including improvements in the thermal-hydraulic modeling required for better understanding the consequences of high energy deposition transients in particular for the ITER limiter case.

Assessment and modification of an ion source grid design in KSTAR neutral beam system.

PubMed

Lee, Dong Won; Shin, Kyu In; Jin, Hyung Gon; Choi, Bo Guen; Kim, Tae-Seong; Jeong, Seung Ho

2014-02-01

A new 2 MW NB (Neutral Beam) ion source for supplying 3.5 MW NB heating for the KSTAR campaign was developed in 2012 and its grid was made from OFHC (Oxygen Free High Conductivity) copper with rectangular cooling channels. However, the plastic deformation such as a bulging in the plasma grid of the ion source was found during the overhaul period after the 2012 campaign. A thermal-hydraulic and a thermo-mechanical analysis using the conventional code, ANSYS, were carried out and the thermal fatigue life assessment was evaluated. It was found that the thermal fatigue life of the OFHC copper grid was about 335 cycles in case of 0.165 MW/m(2) heat flux and it gave too short fatigue life to be used as a KSTAR NB ion source grid. To overcome the limited fatigue life of the current design, the following methods were proposed in the present study: (1) changing the OHFC copper to CuCrZr, copper-alloy or (2) adopting a new design with a pure Mo metal grid and CuCrZr tubes. It is confirmed that the proposed methods meet the requirements by performing the same assessment.

Summary of the thermal evaluation of LWBR (LWBR Development Program)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lerner, S.; McWilliams, K.D.; Stout, J.W.

1980-03-01

This report describes the thermal evaluation of the core for the Shippingport Light Water Breeder Reactor. This core contains unique thermal-hydraulic features such as (1) close rod-to-rod proximity, (2) an open-lattice array of fuel rods with two different diameters and rod-to-rod spacings in the same flow region, (3) triplate orifices located at both the entrance and exit of fuel modules and (4) a hydraulically-balanced movable-fuel system coupled with (5) axial-and-radial fuel zoning for reactivity control. Performance studies used reactor thermal principles such as the hot-and-nominal channel concept and related nuclear/engineering design allowances. These were applied to models of three-dimensional roddedmore » arrays comprising the core fuel regions.« less

Interface-based two-way tuning of the in-plane thermal transport in nanofilms

NASA Astrophysics Data System (ADS)

Hua, Yu-Chao; Cao, Bing-Yang

2018-03-01

Here, the two-way tuning of in-plane thermal transport is obtained in the bi-layer nanofilms with an interfacial effect by using the Boltzmann transport equation (BTE) and the phonon Monte Carlo (MC) technique. A thermal conductivity model was derived from the BTE and verified by the MC simulations. Both the model and the MC simulations indicate that the tuning of the thermal transport can be bidirectional (reduced or enhanced), depending on the interface conditions (i.e., roughness and adhesion energy) and the phonon property dissimilarity at the interface. For the identical-material interface, the emergence of thermal conductivity variation requires two conditions: (a) the interface is not completely specular and (b) the transmission specularity parameter differs from the reflection specularity parameter at the interface. When the transmission specularity parameter is larger than the reflection specularity parameter at the interface, the thermal conductivity improvement effect emerges, whereas the thermal conductivity reduction effect occurs. For the disparate-material interface, the phonon property perturbation near the interface causes the thermal conductivity variation, even when neither the above two conditions are satisfied. The mean free path ratio (γ) between the disparate materials was defined to characterize the phonon property dissimilarity. γ > 1 can lead to the thermal conductivity improvement effect, while γ < 1 corresponds to the thermal conductivity reduction effect. Our work provides a more in-depth understanding of the interfacial effect on the nanoscale thermal transport, with an applicable predictive model, which can be helpful for predicting and manipulating phonon transport in nanofilms.

Temperature distribution by the effect of groundwater flow in an aquifer thermal energy storage system model

NASA Astrophysics Data System (ADS)

Shim, B.

2005-12-01

Aquifer thermal energy storage (ATES) can be a cost-effective and renewable energy source, depending on site-specific thermohydraulic conditions. To design an effective ATES system, the understanding of thermohydraulic processes is necessary. The heat transfer phenomena of an aquifer heat storage system are simulated with the scenario of heat pump operation of pumping and waste water reinjection in a two layered confined aquifer model having the effect of groundwater movement. Temperature distribution of the aquifer model is generated, and hydraulic heads and temperature variations are monitored at both wells during simulation days. The average groundwater velocities are determined with two assumed hydraulic gradients set by boundary conditions, and the effect of groundwater flow are shown at the generated thermal distributions at three different depth slices. The generated temperature contour lines at the hydraulic gradient of 0.001 are shaped circular, and the center is moved less than 5 m to the east in 365 days. However at the hydraulic gradient of 0.01, the contour centers of the east well at each depth slice are moved near the east boundary and the movement of temperature distribution is increased at the lower aquifer. By the analysis of thermal interference data between two wells the efficiency of a heat pump operation model is validated, and the variation of heads is monitored at injection, pumping and stabilized state. The thermal efficiency of the ATES system model is represented as highly depended on groundwater flow velocity and direction. Therefore the hydrogeologic condition for the system site should be carefully surveyed.

SlugIn 1.0: A Free Tool for Automated Slug Test Analysis.

PubMed

Martos-Rosillo, Sergio; Guardiola-Albert, Carolina; Padilla Benítez, Alberto; Delgado Pastor, Joaquín; Azcón González, Antonio; Durán Valsero, Juan José

2018-05-01

The correct characterization of aquifer parameters is essential for water-supply and water-quality investigations. Slug tests are widely used for these purposes. While free software is available to interpret slug tests, some codes are not user-friendly, or do not include a wide range of methods to interpret the results, or do not include automatic, inverse solutions to the test data. The private sector has also generated several good programs to interpret slug test data, but they are not free of charge. The computer program SlugIn 1.0 is available online for free download, and is demonstrated to aid in the analysis of slug tests to estimate hydraulic parameters. The program provides an easy-to-use Graphical User Interface. SlugIn 1.0 incorporates automated parameter estimation and facilitates the visualization of several interpretations of the same test. It incorporates solutions for confined and unconfined aquifers, partially penetrating wells, skin effects, shape factor, anisotropy, high hydraulic conductivity formations and the Mace test for large-diameter wells. It is available in English and Spanish and can be downloaded from the web site of the Geological Survey of Spain. Two field examples are presented to illustrate how the software operates. © 2018, National Ground Water Association.

CNT-based Thermal Interface Materials for Load-Bearing Aerospace Applications

DTIC Science & Technology

2012-08-01

CNT -based Thermal Interface Materials for Load-Bearing Aerospace Applications Michael Bifano, Pankaj Kaul and Vikas Prakash (PI) Department...4. TITLE AND SUBTITLE CNT -based Thermal Interface Materials for Load-Bearing Aerospace Applications 5a. CONTRACT NUMBER 5b. GRANT NUMBER 5c...Z39-18 Objective Develop multifunctional CNT -epoxy Thermal Interface Materials (TIMs) for load bearing aerospace applications. Emphasis - To

Thermal-Flow Code for Modeling Gas Dynamics and Heat Transfer in Space Shuttle Solid Rocket Motor Joints

NASA Technical Reports Server (NTRS)

Wang, Qunzhen; Mathias, Edward C.; Heman, Joe R.; Smith, Cory W.

2000-01-01

A new, thermal-flow simulation code, called SFLOW. has been developed to model the gas dynamics, heat transfer, as well as O-ring and flow path erosion inside the space shuttle solid rocket motor joints by combining SINDA/Glo, a commercial thermal analyzer. and SHARPO, a general-purpose CFD code developed at Thiokol Propulsion. SHARP was modified so that friction, heat transfer, mass addition, as well as minor losses in one-dimensional flow can be taken into account. The pressure, temperature and velocity of the combustion gas in the leak paths are calculated in SHARP by solving the time-dependent Navier-Stokes equations while the heat conduction in the solid is modeled by SINDA/G. The two codes are coupled by the heat flux at the solid-gas interface. A few test cases are presented and the results from SFLOW agree very well with the exact solutions or experimental data. These cases include Fanno flow where friction is important, Rayleigh flow where heat transfer between gas and solid is important, flow with mass addition due to the erosion of the solid wall, a transient volume venting process, as well as some transient one-dimensional flows with analytical solutions. In addition, SFLOW is applied to model the RSRM nozzle joint 4 subscale hot-flow tests and the predicted pressures, temperatures (both gas and solid), and O-ring erosions agree well with the experimental data. It was also found that the heat transfer between gas and solid has a major effect on the pressures and temperatures of the fill bottles in the RSRM nozzle joint 4 configuration No. 8 test.

Transient three-dimensional thermal-hydraulic analysis of nuclear reactor fuel rod arrays: general equations and numerical scheme

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wnek, W.J.; Ramshaw, J.D.; Trapp, J.A.

1975-11-01

A mathematical model and a numerical solution scheme for thermal- hydraulic analysis of fuel rod arrays are given. The model alleviates the two major deficiencies associated with existing rod array analysis models, that of a correct transverse momentum equation and the capability of handling reversing and circulatory flows. Possible applications of the model include steady state and transient subchannel calculations as well as analysis of flows in heat exchangers, other engineering equipment, and porous media. (auth)

Mercury Thermal Hydraulic Loop (MTHL) Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Felde, David K.; Crye, Jason Michael; Wendel, Mark W.

2017-03-01

The Spallation Neutron Source (SNS) is a high-power linear accelerator built at Oak Ridge National Laboratory (ORNL) which incorporates the use of a flowing liquid mercury target. The Mercury Thermal Hydraulic Loop (MTHL) was constructed to investigate and verify the heat transfer characteristics of liquid mercury in a rectangular channel. This report provides a compilation of previously reported results from the water-cooled and electrically heated straight and curved test sections that simulate the geometry of the window cooling channel in the target nose region.

An analysis of the sliding pressure start-up of SCWR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, F.; Yang, J.; Li, H.

In this paper, the preliminary sliding pressure start-up system and scheme of supercritical water-cooled reactor in CGNPC (CGN-SCWR) were proposed. Thermal-hydraulic behavior in start-up procedures was analyzed in detail by employing advanced reactor subchannel analysis software ATHAS. The maximum cladding temperature (MCT for short) and core power of fuel assembly during the whole start-up process were investigated comparatively. The results show that the recommended start-up scheme meets the design requirements from the perspective of thermal-hydraulic. (authors)

Geoengineering Research for a Deep Underground Science and Engineering Laboratory in Sedimentary Rock

NASA Astrophysics Data System (ADS)

Mauldon, M.

2004-12-01

A process to identify world-class research for a Deep Underground Science and Engineering Laboratory (DUSEL) in the USA has been initiated by NSF. While allowing physicists to study, inter alia, dark matter and dark energy, this laboratory will create unprecedented opportunities for biologists to study deep life, geoscientists to study crustal processes and geoengineers to study the behavior of rock, fluids and underground cavities at depth, on time scales of decades. A substantial portion of the nation's future infrastructure is likely to be sited underground because of energy costs, urban crowding and vulnerability of critical surface facilities. Economic and safe development of subsurface space will require an improved ability to engineer the geologic environment. Because of the prevalence of sedimentary rock in the upper continental crust, much of this subterranean infrastructure will be hosted in sedimentary rock. Sedimentary rocks are fundamentally anisotropic due to lithology and bedding, and to discontinuities ranging from microcracks to faults. Fractures, faults and bedding planes create structural defects and hydraulic pathways over a wide range of scales. Through experimentation, observation and monitoring in a sedimentary rock DUSEL, in conjunction with high performance computational models and visualization tools, we will explore the mechanical and hydraulic characteristics of layered rock. DUSEL will permit long-term experiments on 100 m blocks of rock in situ, accessed via peripheral tunnels. Rock volumes will be loaded to failure and monitored for post-peak behavior. The response of large rock bodies to stress relief-driven, time-dependent strain will be monitored over decades. Large block experiments will be aimed at measurement of fluid flow and particle/colloid transport, in situ mining (incl. mining with microbes), remediation technologies, fracture enhancement for resource extraction and large scale long-term rock mass response to induced stresses - with parallel geophysical imaging of the rock mass (and subsequent verification) flow and transport processes, and time-dependent stress and strain. An experimental advantage of sedimentary rock is the presence of pervasive mechanical interfaces (bedding planes), which suggest a host of experimental designs on large rock blocks and slabs (induced flexure, shear strength of interfaces, etc). Thus DUSEL will enable fundamental research about the behavior of a layered rock mass - the dominant structural architecture in near-surface environments worldwide. A further benefit is the natural suitability of sedimentary rocks for experiments related to oil and gas production, or to CO2 sequestration. For example, fluid-induced fracturing of sedimentary rock has long been used by the hydrocarbon industry to improve oil and coal bed methane recovery. Since some fracturing agents are potential contaminants, a major concern and legal responsibility in the US is ensuring the integrity of nearby aquifers. Hydraulic fracturing from a sedimentary rock DUSEL will be followed by injection of low viscosity grout. The rock mass will then be mined back to expose network characteristics of the induced hydraulic fractures. Key questions related to hydrocarbon extraction, CO2 sequestration, waste isolation, and remediation of subsurface contaminants depend critically on the connectivity and architecture of fractures and on coupled thermal, hydrological, mechanical and chemical processes. Fluid flow, particle transport and reaction transport processes are coupled to the stress across fractures, and to thermal, chemical and hydraulic gradients. All can best be studied via large block tests in a subterranean laboratory, ideally in a sedimentary environment.

Satellite freeze forecast system: Executive summary

NASA Technical Reports Server (NTRS)

Martsolf, J. D. (Principal Investigator)

1983-01-01

A satellite-based temperature monitoring and prediction system consisting of a computer controlled acquisition, processing, and display system and the ten automated weather stations called by that computer was developed and transferred to the national weather service. This satellite freeze forecasting system (SFFS) acquires satellite data from either one of two sources, surface data from 10 sites, displays the observed data in the form of color-coded thermal maps and in tables of automated weather station temperatures, computes predicted thermal maps when requested and displays such maps either automatically or manually, archives the data acquired, and makes comparisons with historical data. Except for the last function, SFFS handles these tasks in a highly automated fashion if the user so directs. The predicted thermal maps are the result of two models, one a physical energy budget of the soil and atmosphere interface and the other a statistical relationship between the sites at which the physical model predicts temperatures and each of the pixels of the satellite thermal map.

Effect of Freeze-Thaw Cycles on Soil Nitrogen Reactive Transport in a Polygonal Arctic Tundra Ecosystem at Barrow AK Using 3-D Coupled ALM-PFLOTRAN

NASA Astrophysics Data System (ADS)

Yuan, F.; Wang, G.; Painter, S. L.; Tang, G.; Xu, X.; Kumar, J.; Bisht, G.; Hammond, G. E.; Mills, R. T.; Thornton, P. E.; Wullschleger, S. D.

2017-12-01

In Arctic tundra ecosystem soil freezing-thawing is one of dominant physical processes through which biogeochemical (e.g., carbon and nitrogen) cycles are tightly coupled. Besides hydraulic transport, freezing-thawing can cause pore water movement and aqueous species gradients, which are additional mechanisms for soil nitrogen (N) reactive-transport in Tundra ecosystem. In this study, we have fully coupled an in-development ESM(i.e., Advanced Climate Model for Energy, ACME)'s Land Model (ALM) aboveground processes with a state-of-the-art massively parallel 3-D subsurface thermal-hydrology and reactive transport code, PFLOTRAN. The resulting coupled ALM-PFLOTRAN model is a Land Surface Model (LSM) capable of resolving 3-D soil thermal-hydrological-biogeochemical cycles. This specific version of PFLOTRAN has incorporated CLM-CN Converging Trophic Cascade (CTC) model and a full and simple but robust soil N cycle. It includes absorption-desorption for soil NH4+ and gas dissolving-degasing process as well. It also implements thermal-hydrology mode codes with three newly-modified freezing-thawing algorithms which can greatly improve computing performance in regarding to numerical stiffness at freezing-point. Here we tested the model in fully 3-D coupled mode at the Next Generation Ecosystem Experiment-Arctic (NGEE-Arctic) field intensive study site at the Barrow Environmental Observatory (BEO), AK. The simulations show that: (1) synchronous coupling of soil thermal-hydrology and biogeochemistry in 3-D can greatly impact ecosystem dynamics across polygonal tundra landscape; and (2) freezing-thawing cycles can add more complexity to the system, resulting in greater mobility of soil N vertically and laterally, depending upon local micro-topography. As a preliminary experiment, the model is also implemented for Pan-Arctic region in 1-D column mode (i.e. no lateral connection), showing significant differences compared to stand-alone ALM. The developed ALM-PFLOTRAN coupling codes embeded within ESM will be used for Pan-Arctic regional evaluation of climate change-caused ecosystem responses and their feedbacks to climate system at various scales.

Thermal force induced by the presence of a particle near a solidifying interface.

PubMed

Hadji, L

2001-11-01

The presence of a foreign particle in the melt, ahead of a solid-liquid interface, leads to the onset of interfacial deformations if the thermal conductivity of the particle, k(p), differs from that of the melt, k(l). In this paper, the influence of the thermal conductivity contrast on the interaction between the solidifying interface and the particle is quantified. We show that the interface distortion gives rise to a thermal force whose expression is given by F(th)=2piLGa3(1-alpha)/(2+alpha)T(m), where L is the latent heat of fusion per unit volume, T(m) is the melting point, a is the particle's radius, G the thermal gradient in the liquid phase and alpha=k(p)/k(l). The derivation makes use of the following assumptions: (i) the particle is small compared to the horizontal extent of the interface, (ii) the particle is placed in the near proximity of the deformable solid-liquid interface, and (iii) the interface is practically immobile in the calculation of the thermal field, i.e., V

Thermal Hydraulics Design and Analysis Methodology for a Solid-Core Nuclear Thermal Rocket Engine Thrust Chamber

NASA Technical Reports Server (NTRS)

Wang, Ten-See; Canabal, Francisco; Chen, Yen-Sen; Cheng, Gary; Ito, Yasushi

2013-01-01

Nuclear thermal propulsion is a leading candidate for in-space propulsion for human Mars missions. This chapter describes a thermal hydraulics design and analysis methodology developed at the NASA Marshall Space Flight Center, in support of the nuclear thermal propulsion development effort. The objective of this campaign is to bridge the design methods in the Rover/NERVA era, with a modern computational fluid dynamics and heat transfer methodology, to predict thermal, fluid, and hydrogen environments of a hypothetical solid-core, nuclear thermal engine the Small Engine, designed in the 1960s. The computational methodology is based on an unstructured-grid, pressure-based, all speeds, chemically reacting, computational fluid dynamics and heat transfer platform, while formulations of flow and heat transfer through porous and solid media were implemented to describe those of hydrogen flow channels inside the solid24 core. Design analyses of a single flow element and the entire solid-core thrust chamber of the Small Engine were performed and the results are presented herein

Modeling of Flow Blockage in a Liquid Metal-Cooled Reactor Subassembly with a Subchannel Analysis Code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Jeong, Hae-Yong; Ha, Kwi-Seok; Chang, Won-Pyo

The local blockage in a subassembly of a liquid metal-cooled reactor (LMR) is of importance to the plant safety because of the compact design and the high power density of the core. To analyze the thermal-hydraulic parameters in a subassembly of a liquid metal-cooled reactor with a flow blockage, the Korea Atomic Energy Research Institute has developed the MATRA-LMR-FB code. This code uses the distributed resistance model to describe the sweeping flow formed by the wire wrap around the fuel rods and to model the recirculation flow after a blockage. The hybrid difference scheme is also adopted for the descriptionmore » of the convective terms in the recirculating wake region of low velocity. Some state-of-the-art turbulent mixing models were implemented in the code, and the models suggested by Rehme and by Zhukov are analyzed and found to be appropriate for the description of the flow blockage in an LMR subassembly. The MATRA-LMR-FB code predicts accurately the experimental data of the Oak Ridge National Laboratory 19-pin bundle with a blockage for both the high-flow and low-flow conditions. The influences of the distributed resistance model, the hybrid difference method, and the turbulent mixing models are evaluated step by step with the experimental data. The appropriateness of the models also has been evaluated through a comparison with the results from the COMMIX code calculation. The flow blockage for the KALIMER design has been analyzed with the MATRA-LMR-FB code and is compared with the SABRE code to guarantee the design safety for the flow blockage.« less

MDANSE: An Interactive Analysis Environment for Molecular Dynamics Simulations.

PubMed

Goret, G; Aoun, B; Pellegrini, E

2017-01-23

The MDANSE software-Molecular Dynamics Analysis of Neutron Scattering Experiments-is presented. It is an interactive application for postprocessing molecular dynamics (MD) simulations. Given the widespread use of MD simulations in material and biomolecular sciences to get a better insight for experimental techniques such as thermal neutron scattering (TNS), the development of MDANSE has focused on providing a user-friendly, interactive, graphical user interface for analyzing many trajectories in the same session and running several analyses simultaneously independently of the interface. This first version of MDANSE already proposes a broad range of analyses, and the application has been designed to facilitate the introduction of new analyses in the framework. All this makes MDANSE a valuable tool for extracting useful information from trajectories resulting from a wide range of MD codes.

NGNP Data Management and Analysis System Modeling Capabilities

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cynthia D. Gentillon

2009-09-01

Projects for the very-high-temperature reactor (VHTR) program provide data in support of Nuclear Regulatory Commission licensing of the VHTR. Fuel and materials to be used in the reactor are tested and characterized to quantify performance in high temperature and high fluence environments. In addition, thermal-hydraulic experiments are conducted to validate codes used to assess reactor safety. The VHTR Program has established the NGNP Data Management and Analysis System (NDMAS) to ensure that VHTR data are (1) qualified for use, (2) stored in a readily accessible electronic form, and (3) analyzed to extract useful results. This document focuses on the thirdmore » NDMAS objective. It describes capabilities for displaying the data in meaningful ways and identifying relationships among the measured quantities that contribute to their understanding.« less

Estimating spatially distributed soil texture using time series of thermal remote sensing - a case study in central Europe

NASA Astrophysics Data System (ADS)

Müller, Benjamin; Bernhardt, Matthias; Jackisch, Conrad; Schulz, Karsten

2016-09-01

For understanding water and solute transport processes, knowledge about the respective hydraulic properties is necessary. Commonly, hydraulic parameters are estimated via pedo-transfer functions using soil texture data to avoid cost-intensive measurements of hydraulic parameters in the laboratory. Therefore, current soil texture information is only available at a coarse spatial resolution of 250 to 1000 m. Here, a method is presented to derive high-resolution (15 m) spatial topsoil texture patterns for the meso-scale Attert catchment (Luxembourg, 288 km2) from 28 images of ASTER (advanced spaceborne thermal emission and reflection radiometer) thermal remote sensing. A principle component analysis of the images reveals the most dominant thermal patterns (principle components, PCs) that are related to 212 fractional soil texture samples. Within a multiple linear regression framework, distributed soil texture information is estimated and related uncertainties are assessed. An overall root mean squared error (RMSE) of 12.7 percentage points (pp) lies well within and even below the range of recent studies on soil texture estimation, while requiring sparser sample setups and a less diverse set of basic spatial input. This approach will improve the generation of spatially distributed topsoil maps, particularly for hydrologic modeling purposes, and will expand the usage of thermal remote sensing products.

Interface test series: An in situ study of factors affecting the containment of hydraulic fractures

NASA Astrophysics Data System (ADS)

Warpinski, N. R.; Finley, S. J.; Vollendorf, W. C.; Obrien, M.; Eshom, E.

1982-02-01

In situ experiments, which are accessible for direct observation by mineback, were conducted to determine the effect that material-property interfaces and in situ stress differences have on hydraulic fracture propagation and the resultant overall geometry. These experiments show conclusively that a difference in elastic modulus at a geologic interface has little or no effect on crack growth and, therefore, is not a feature which would promote containment of fractures within a specified reservoir zone. However, differences in the in situ stress between adjacent layers is shown to have a considerable influence on fracture propagation. Experiments were conducted in a low modulus ash-fall tuff which contained two layers of high minimum principal in situ stress and which was overlain by a formation with at least a factor of 5 increase in elastic modulus. Fractures were observed to terminate in regions of high minimum principal in situ stress in nearly every case.

Integrated Thermal Response Tool for Earth Entry Vehicles

NASA Technical Reports Server (NTRS)

Chen, Y.-K.; Milos, F. S.; Partridge, Harry (Technical Monitor)

2001-01-01

A system is presented for multi-dimensional, fully-coupled thermal response modeling of hypersonic entry vehicles. The system consists of a two-dimensional implicit thermal response, pyrolysis and ablation program (TITAN), a commercial finite-element thermal and mechanical analysis code (MARC), and a high fidelity Navier-Stokes equation solver (GIANTS). The simulations performed by this integrated system include hypersonic flow-field, fluid and solid interaction, ablation, shape change, pyrolysis gas generation and flow, and thermal response of heatshield and structure. The thermal response of the ablating and charring heatshield material is simulated using TITAN, and that of the underlying structural is simulated using MARC. The ablating heatshield is treated as an outer boundary condition of the structure, and continuity conditions of temperature and heat flux are imposed at the interface between TITAN and MARC. Aerothermal environments with fluid and solid interaction are predicted by coupling TITAN and GIANTS through surface energy balance equations. With this integrated system, the aerothermal environments for an entry vehicle and the thermal response of both the heatshield and the structure can be obtained simultaneously. Representative computations for a proposed blunt body earth entry vehicle are presented and discussed in detail.

Thermal Response Modeling System for a Mars Sample Return Vehicle

NASA Technical Reports Server (NTRS)

Chen, Y.-K.; Miles, Frank S.; Arnold, Jim (Technical Monitor)

2001-01-01

A multi-dimensional, coupled thermal response modeling system for analysis of hypersonic entry vehicles is presented. The system consists of a high fidelity Navier-Stokes equation solver (GIANTS), a two-dimensional implicit thermal response, pyrolysis and ablation program (TITAN), and a commercial finite-element thermal and mechanical analysis code (MARC). The simulations performed by this integrated system include hypersonic flowfield, fluid and solid interaction, ablation, shape change, pyrolysis gas eneration and flow, and thermal response of heatshield and structure. The thermal response of the heatshield is simulated using TITAN, and that of the underlying structural is simulated using MARC. The ablating heatshield is treated as an outer boundary condition of the structure, and continuity conditions of temperature and heat flux are imposed at the interface between TITAN and MARC. Aerothermal environments with fluid and solid interaction are predicted by coupling TITAN and GIANTS through surface energy balance equations. With this integrated system, the aerothermal environments for an entry vehicle and the thermal response of the entire vehicle can be obtained simultaneously. Representative computations for a flat-faced arc-jet test model and a proposed Mars sample return capsule are presented and discussed.

Thermal Response Modeling System for a Mars Sample Return Vehicle

NASA Technical Reports Server (NTRS)

Chen, Y.-K.; Milos, F. S.

2002-01-01

A multi-dimensional, coupled thermal response modeling system for analysis of hypersonic entry vehicles is presented. The system consists of a high fidelity Navier-Stokes equation solver (GIANTS), a two-dimensional implicit thermal response, pyrolysis and ablation program (TITAN), and a commercial finite element thermal and mechanical analysis code (MARC). The simulations performed by this integrated system include hypersonic flowfield, fluid and solid interaction, ablation, shape change, pyrolysis gas generation and flow, and thermal response of heatshield and structure. The thermal response of the heatshield is simulated using TITAN, and that of the underlying structural is simulated using MARC. The ablating heatshield is treated as an outer boundary condition of the structure, and continuity conditions of temperature and heat flux are imposed at the interface between TITAN and MARC. Aerothermal environments with fluid and solid interaction are predicted by coupling TITAN and GIANTS through surface energy balance equations. With this integrated system, the aerothermal environments for an entry vehicle and the thermal response of the entire vehicle can be obtained simultaneously. Representative computations for a flat-faced arc-jet test model and a proposed Mars sample return capsule are presented and discussed.

Thermal hydraulic feasibility assessment of the hot conditioning system and process

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heard, F.J.

1996-10-10

The Spent Nuclear Fuel Project was established to develop engineered solutions for the expedited removal, stabilization, and storage of spent nuclear fuel from the K Basins at the U.S. Department of Energy`s Hanford Site in Richland, Washington. A series of analyses have been completed investigating the thermal-hydraulic performance and feasibility of the proposed Hot Conditioning System and process for the Spent Nuclear Fuel Project. The analyses were performed using a series of thermal-hydraulic models that could respond to all process and safety-related issues that may arise pertaining to the Hot Conditioning System. The subject efforts focus on independently investigating, quantifying,more » and establishing the governing heat production and removal mechanisms, flow distributions within the multi-canister overpack, and performing process simulations for various purge gases under consideration for the Hot Conditioning System, as well as obtaining preliminary results for comparison with and verification of other analyses, and providing technology- based recommendations for consideration and incorporation into the Hot Conditioning System design bases.« less

Thermal Hydraulic Design and Analysis of a Water-Cooled Ceramic Breeder Blanket with Superheated Steam for CFETR

NASA Astrophysics Data System (ADS)

Cheng, Xiaoman; Ma, Xuebin; Jiang, Kecheng; Chen, Lei; Huang, Kai; Liu, Songlin

2015-09-01

The water-cooled ceramic breeder blanket (WCCB) is one of the blanket candidates for China fusion engineering test reactor (CFETR). In order to improve power generation efficiency and tritium breeding ratio, WCCB with superheated steam is under development. The thermal-hydraulic design is the key to achieve the purpose of safe heat removal and efficient power generation under normal and partial loading operation conditions. In this paper, the coolant flow scheme was designed and one self-developed analytical program was developed, based on a theoretical heat transfer model and empirical correlations. Employing this program, the design and analysis of related thermal-hydraulic parameters were performed under different fusion power conditions. The results indicated that the superheated steam water-cooled blanket is feasible. supported by the National Special Project for Magnetic Confined Nuclear Fusion Energy of China (Nos. 2013GB108004, 2014GB122000 and 2014GB119000), and National Natural Science Foundation of China (No. 11175207)

Study of steam condensation at sub-atmospheric pressure: setting a basic research using MELCOR code

NASA Astrophysics Data System (ADS)

Manfredini, A.; Mazzini, M.

2017-11-01

One of the most serious accidents that can occur in the experimental nuclear fusion reactor ITER is the break of one of the headers of the refrigeration system of the first wall of the Tokamak. This results in water-steam mixture discharge in vacuum vessel (VV), with consequent pressurization of this container. To prevent the pressure in the VV exceeds 150 KPa absolute, a system discharges the steam inside a suppression pool, at an absolute pressure of 4.2 kPa. The computer codes used to analyze such incident (eg. RELAP 5 or MELCOR) are not validated experimentally for such conditions. Therefore, we planned a basic research, in order to have experimental data useful to validate the heat transfer correlations used in these codes. After a thorough literature search on this topic, ACTA, in collaboration with the staff of ITER, defined the experimental matrix and performed the design of the experimental apparatus. For the thermal-hydraulic design of the experiments, we executed a series of calculations by MELCOR. This code, however, was used in an unconventional mode, with the development of models suited respectively to low and high steam flow-rate tests. The article concludes with a discussion of the placement of experimental data within the map featuring the phenomenon characteristics, showing the importance of the new knowledge acquired, particularly in the case of chugging.

Analysis of the influence of the heat transfer phenomena on the late phase of the ThAI Iod-12 test

NASA Astrophysics Data System (ADS)

Gonfiotti, B.; Paci, S.

2014-11-01

Iodine is one of the major contributors to the source term during a severe accident in a Nuclear Power Plant for its volatility and high radiological consequences. Therefore, large efforts have been made to describe the Iodine behaviour during an accident, especially in the containment system. Due to the lack of experimental data, in the last years many attempts were carried out to fill the gaps on the knowledge of Iodine behaviour. In this framework, two tests (ThAI Iod-11 and Iod-12) were carried out inside a multi-compartment steel vessel. A quite complex transient characterizes these two tests; therefore they are also suitable for thermal- hydraulic benchmarks. The two tests were originally released for a benchmark exercise during the SARNET2 EU Project. At the end of this benchmark a report covering the main findings was issued, stating that the common codes employed in SA studies were able to simulate the tests but with large discrepancies. The present work is then related to the application of the new versions of ASTEC and MELCOR codes with the aim of carry out a new code-to-code comparison vs. ThAI Iod-12 experimental data, focusing on the influence of the heat exchanges with the outer environment, which seems to be one of the most challenging issues to cope with.

Development of concepts for the management of shallow geothermal resources in urban areas - Experience gained from the Basel and Zaragoza case studies

NASA Astrophysics Data System (ADS)

García-Gil, Alejandro; Epting, Jannis; Mueller, Matthias H.; Huggenberger, Peter; Vázquez-Suñé, Enric

2015-04-01

In urban areas the shallow subsurface often is used as a heat resource (shallow geothermal energy), i.e. for the installation and operation of a broad variety of geothermal systems. Increasingly, groundwater is used as a low-cost heat sink, e.g. for building acclimatization. Together with other shallow geothermal exploitation systems significantly increased groundwater temperatures have been observed in many urban areas (urban heat island effect). The experience obtained from two selected case study cities in Basel (CH) and Zaragoza (ES) has allowed developing concepts and methods for the management of thermal resources in urban areas. Both case study cities already have a comprehensive monitoring network operating (hydraulics and temperature) as well as calibrated high-resolution numerical groundwater flow and heat-transport models. The existing datasets and models have allowed to compile and compare the different hydraulic and thermal boundary conditions for both groundwater bodies, including: (1) River boundaries (River Rhine and Ebro), (2) Regional hydraulic and thermal settings, (3) Interaction with the atmosphere under consideration of urbanization and (4) Anthropogenic quantitative and thermal groundwater use. The potential natural states of the considered groundwater bodies also have been investigated for different urban settings and varying processes concerning groundwater flow and thermal regimes. Moreover, concepts for the management of thermal resources in urban areas and the transferability of the applied methods to other urban areas are discussed. The methods used provide an appropriate selection of parameters (spatiotemporal resolution) that have to be measured for representative interpretations of groundwater flow and thermal regimes of specific groundwater bodies. From the experience acquired from the case studies it is shown that understanding the variable influences of the specific geological and hydrogeological as well as hydraulic and thermal boundary conditions in urban settings is crucial. It also could be shown that good quality data are necessary to appropriately define and investigate thermal boundary conditions and the temperature development in urban systems. Groundwater temperatures in both investigated groundwater bodies are already over-heated and essentially impede further thermal groundwater use for cooling purposes. Current legislation approaches are not suitable to evaluate new concessions for thermal exploitation. Therefore, novel approaches for the assessment of new concessions which take into account the complex interaction of natural boundaries as well as existing shallow geothermal systems have to be developed.

Engineering Interface Structures and Thermal Stabilities via SPD Processing in Bulk Nanostructured Metals

DOE PAGES

Zheng, Shijian; Carpenter, John S.; McCabe, Rodney J.; ...

2014-02-27

Nanostructured metals achieve extraordinary strength but suffer from low thermal stability, both a consequence of a high fraction of interfaces. Overcoming this tradeoff relies on making the interfaces themselves thermally stable. In this paper, we show that the atomic structures of bi-metal interfaces in macroscale nanomaterials suitable for engineering structures can be significantly altered via changing the severe plastic deformation (SPD) processing pathway. Two types of interfaces are formed, both exhibiting a regular atomic structure and providing for excellent thermal stability, up to more than half the melting temperature of one of the constituents. Most importantly, the thermal stability ofmore » one is found to be significantly better than the other, indicating the exciting potential to control and optimize macroscale robustness via atomic-scale bimetal interface tuning. As a result, we demonstrate an innovative way to engineer pristine bimetal interfaces for a new class of simultaneously strong and thermally stable materials.« less

Thermal Interface Comparisons Under Flight Like Conditions

NASA Technical Reports Server (NTRS)

Rodriquez-Ruiz, Juan

2008-01-01

Thermal interface materials are used in bolted interfaces to promote good thermal conduction between the two. The mounting surface can include panels, heat pipes, electronics boxes, etc.. . On Lunar Reconnaissance Orbiter (LRO) project the results are directly applicable: a) Several high power avionics boxes b) Several interfaces from RWA to radiator through heat pipe network

Application of multiple-point geostatistics to simulate the effect of small-scale aquifer heterogeneity on the efficiency of aquifer thermal energy storage

NASA Astrophysics Data System (ADS)

Possemiers, Mathias; Huysmans, Marijke; Batelaan, Okke

2015-08-01

Adequate aquifer characterization and simulation using heat transport models are indispensible for determining the optimal design for aquifer thermal energy storage (ATES) systems and wells. Recent model studies indicate that meter-scale heterogeneities in the hydraulic conductivity field introduce a considerable uncertainty in the distribution of thermal energy around an ATES system and can lead to a reduction in the thermal recoverability. In a study site in Bierbeek, Belgium, the influence of centimeter-scale clay drapes on the efficiency of a doublet ATES system and the distribution of the thermal energy around the ATES wells are quantified. Multiple-point geostatistical simulation of edge properties is used to incorporate the clay drapes in the models. The results show that clay drapes have an influence both on the distribution of thermal energy in the subsurface and on the efficiency of the ATES system. The distribution of the thermal energy is determined by the strike of the clay drapes, with the major axis of anisotropy parallel to the clay drape strike. The clay drapes have a negative impact (3.3-3.6 %) on the energy output in the models without a hydraulic gradient. In the models with a hydraulic gradient, however, the presence of clay drapes has a positive influence (1.6-10.2 %) on the energy output of the ATES system. It is concluded that it is important to incorporate small-scale heterogeneities in heat transport models to get a better estimate on ATES efficiency and distribution of thermal energy.

Application of multiple-point geostatistics to simulate the effect of small scale aquifer heterogeneity on the efficiency of Aquifer Thermal Energy Storage (ATES)

NASA Astrophysics Data System (ADS)

Possemiers, Mathias; Huysmans, Marijke; Batelaan, Okke

2015-04-01

Adequate aquifer characterization and simulation using heat transport models are indispensible for determining the optimal design for Aquifer Thermal Energy Storage (ATES) systems and wells. Recent model studies indicate that meter scale heterogeneities in the hydraulic conductivity field introduce a considerable uncertainty in the distribution of thermal energy around an ATES system and can lead to a reduction in the thermal recoverability. In this paper, the influence of centimeter scale clay drapes on the efficiency of a doublet ATES system and the distribution of the thermal energy around the ATES wells are quantified. Multiple-point geostatistical simulation of edge properties is used to incorporate the clay drapes in the models. The results show that clay drapes have an influence both on the distribution of thermal energy in the subsurface and on the efficiency of the ATES system. The distribution of the thermal energy is determined by the strike of the clay drapes, with the major axis of anisotropy parallel to the clay drape strike. The clay drapes have a negative impact (3.3 - 3.6%) on the energy output in the models without a hydraulic gradient. In the models with a hydraulic gradient, however, the presence of clay drapes has a positive influence (1.6 - 10.2%) on the energy output of the ATES system. It is concluded that it is important to incorporate small scale heterogeneities in heat transport models to get a better estimate on ATES efficiency and distribution of thermal energy.

Lead Coolant Test Facility Systems Design, Thermal Hydraulic Analysis and Cost Estimate

DOE Office of Scientific and Technical Information (OSTI.GOV)

Soli Khericha; Edwin Harvego; John Svoboda

2012-01-01

The Idaho National Laboratory prepared a preliminary technical and functional requirements (T&FR), thermal hydraulic design and cost estimate for a lead coolant test facility. The purpose of this small scale facility is to simulate lead coolant fast reactor (LFR) coolant flow in an open lattice geometry core using seven electrical rods and liquid lead or lead-bismuth eutectic coolant. Based on review of current world lead or lead-bismuth test facilities and research needs listed in the Generation IV Roadmap, five broad areas of requirements were identified as listed: (1) Develop and Demonstrate Feasibility of Submerged Heat Exchanger; (2) Develop and Demonstratemore » Open-lattice Flow in Electrically Heated Core; (3) Develop and Demonstrate Chemistry Control; (4) Demonstrate Safe Operation; and (5) Provision for Future Testing. This paper discusses the preliminary design of systems, thermal hydraulic analysis, and simplified cost estimate. The facility thermal hydraulic design is based on the maximum simulated core power using seven electrical heater rods of 420 kW; average linear heat generation rate of 300 W/cm. The core inlet temperature for liquid lead or Pb/Bi eutectic is 4200 C. The design includes approximately seventy-five data measurements such as pressure, temperature, and flow rates. The preliminary estimated cost of construction of the facility is $3.7M (in 2006 $). It is also estimated that the facility will require two years to be constructed and ready for operation.« less

Supplemental Thermal-Hydraulic Transient Analyses of BR2 in Support of Conversion to LEU Fuel

DOE Office of Scientific and Technical Information (OSTI.GOV)

Licht, J.; Dionne, B.; Sikik, E.

2016-01-01

Belgian Reactor 2 (BR2) is a research and test reactor located in Mol, Belgium and is primarily used for radioisotope production and materials testing. The Materials Management and Minimization (M3) Reactor Conversion Program of the National Nuclear Security Administration (NNSA) is supporting the conversion of the BR2 reactor from Highly Enriched Uranium (HEU) fuel to Low Enriched Uranium (LEU) fuel. The RELAP5/Mod 3.3 code has been used to perform transient thermal-hydraulic safety analyses of the BR2 reactor to support reactor conversion. A RELAP5 model of BR2 has been validated against select transient BR2 reactor experiments performed in 1963 by showingmore » agreement with measured cladding temperatures. Following the validation, the RELAP5 model was then updated to represent the current use of the reactor; taking into account core configuration, neutronic parameters, trip settings, component changes, etc. Simulations of the 1963 experiments were repeated with this updated model to re-evaluate the boiling risks associated with the currently allowed maximum heat flux limit of 470 W/cm 2 and temporary heat flux limit of 600 W/cm 2. This document provides analysis of additional transient simulations that are required as part of a modern BR2 safety analysis report (SAR). The additional simulations included in this report are effect of pool temperature, reduced steady-state flow rate, in-pool loss of coolant accidents, and loss of external cooling. The simulations described in this document have been performed for both an HEU- and LEU-fueled core.« less

XPI: The Xanadu Parameter Interface

NASA Technical Reports Server (NTRS)

White, N.; Barrett, P.; Oneel, B.; Jacobs, P.

1992-01-01

XPI is a table driven parameter interface which greatly simplifies both command driven programs such as BROWSE and XIMAGE as well as stand alone single-task programs. It moves all of the syntax and semantic parsing of commands and parameters out of the users code into common code and externally defined tables. This allows the programmer to concentrate on writing the code unique to the application rather than reinventing the user interface and for external graphical interfaces to interface with no changes to the command driven program. XPI also includes a compatibility library which allows programs written using the IRAF host interface (Mandel and Roll) to use XPI in place of the IRAF host interface.

Development of Adaptive Model Refinement (AMoR) for Multiphysics and Multifidelity Problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turinsky, Paul

This project investigated the development and utilization of Adaptive Model Refinement (AMoR) for nuclear systems simulation applications. AMoR refers to utilization of several models of physical phenomena which differ in prediction fidelity. If the highest fidelity model is judged to always provide or exceeded the desired fidelity, than if one can determine the difference in a Quantity of Interest (QoI) between the highest fidelity model and lower fidelity models, one could utilize the fidelity model that would just provide the magnitude of the QoI desired. Assuming lower fidelity models require less computational resources, in this manner computational efficiency can bemore » realized provided the QoI value can be accurately and efficiently evaluated. This work utilized Generalized Perturbation Theory (GPT) to evaluate the QoI, by convoluting the GPT solution with the residual of the highest fidelity model determined using the solution from lower fidelity models. Specifically, a reactor core neutronics problem and thermal-hydraulics problem were studied to develop and utilize AMoR. The highest fidelity neutronics model was based upon the 3D space-time, two-group, nodal diffusion equations as solved in the NESTLE computer code. Added to the NESTLE code was the ability to determine the time-dependent GPT neutron flux. The lower fidelity neutronics model was based upon the point kinetics equations along with utilization of a prolongation operator to determine the 3D space-time, two-group flux. The highest fidelity thermal-hydraulics model was based upon the space-time equations governing fluid flow in a closed channel around a heat generating fuel rod. The Homogenous Equilibrium Mixture (HEM) model was used for the fluid and Finite Difference Method was applied to both the coolant and fuel pin energy conservation equations. The lower fidelity thermal-hydraulic model was based upon the same equations as used for the highest fidelity model but now with coarse spatial meshing, corrected somewhat by employing effective fuel heat conduction values. The effectiveness of switching between the highest fidelity model and lower fidelity model as a function of time was assessed using the neutronics problem. Based upon work completed to date, one concludes that the time switching is effective in annealing out differences between the highest and lower fidelity solutions. The effectiveness of using a lower fidelity GPT solution, along with a prolongation operator, to estimate the QoI was also assessed. The utilization of a lower fidelity GPT solution was done in an attempt to avoid the high computational burden associated with solving for the highest fidelity GPT solution. Based upon work completed to date, one concludes that the lower fidelity adjoint solution is not sufficiently accurate with regard to estimating the QoI; however, a formulation has been revealed that may provide a path for addressing this shortcoming.« less

Jet and electromagnetic tomography (JET) of extreme phases of matter in heavy-ion collisions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Heinz, Ulrich

2015-08-31

The Ohio State University (OSU) group contributed to the deliverables of the JET Collaboration three major products: 1. The code package iEBE-VISHNU for modeling the dynamical evolution of the soft medium created in relativistic heavy-ion collisions, from its creation all the way to final freeze-out using a hybrid approach that interfaces a free-streaming partonic pre-equilbrium stage with a (2+1)-dimensional viscous relativistic fluid dynamical stage for the quark-gluon plasma (QGP) phase and the microscopic hadron cascade UrQMD for the hadronic rescattering and freeze-out stage. Except for UrQMD, all dynamical evolution components and interfaces were developed at OSU and tested and implementedmore » in collaboration with the Duke University group. 2. An electromagnetic radiation module for the calculation of thermal photon emission from the QGP and hadron resonance gas stages of a heavy-ion collision, with emission rates that have been corrected for viscous effects in the expanding medium consistent with the bulk evolution. The electromagnetic radiation module was developed under OSU leadership in collaboration with the McGill group and has been integrated in the iEBE-VISHNU code package. 3. An interface between the Monte Carlo jet shower evolution and hadronization codes developed by the Wayne State University (WSU), McGill and Texas A&M groups and the iEBE-VISHNU bulk evolution code, for performing jet quenching and jet shape modification studies in a realistically modeled evolving medium that was tuned to measured soft hadron data. Building on work performed at OSU for the theoretical framework used to describe the interaction of jets with the medium, initial work on the jet shower Monte Carlo was started at OSU and moved to WSU when OSU Visiting Assistant Professor Abhijit Majumder accepted a tenure track faculty position at WSU in September 2011. The jet-hydro interface was developed at OSU and WSU and tested and implemented in collaboration with the McGill, Texas A&M, and LBNL groups.« less

Development of an electronic seepage chamber for extended use in a river.

PubMed

Fritz, Brad G; Mendoza, Donaldo P; Gilmore, Tyler J

2009-01-01

Seepage chambers have been used to characterize the flux of water across the water-sediment interface in a variety of settings. In this work, an electronic seepage chamber was developed specifically for long-term use in a large river where hydraulic gradient reversals occur frequently with river-stage variations. A bidirectional electronic flowmeter coupled with a seepage chamber was used to measure temporal changes in the magnitude and direction of water flux across the water-sediment interface over an 8-week period. The specific discharge measured from the seepage chamber compared favorably with measurements of vertical hydraulic gradient and previous specific discharge calculations. This, as well as other supporting data, demonstrates the effectiveness of the electronic seepage chamber to accurately quantify water flux in two directions over a multimonth period in this setting. The ability to conduct multimonth measurements of water flux at a subhourly frequency in a river system is a critical capability for a seepage chamber in a system where hydraulic gradients change on a daily and seasonal basis.

LARGE-SCALE NATURAL GRADIENT TRACER TEST IN SAND AND GRAVEL, CAPE CODE, MASSACHUSETTS 3. HYDRAULIC CONDUCTI- VITY AND CALCULATED MACRODISPERSIVITIES

EPA Science Inventory

Hydraulic conductivity (K) variability in a sand and gravel aquifer on Cape Cod, Massachusetts, was measured and subsequently used in stochastic transport theories to estimate macrodispersivities. Nearly 1500 K measurements were obtained by borehole flowmeter tests ...

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

Molecular Tagging Velocimetry Development for In-situ Measurement in High-Temperature Test Facility

NASA Technical Reports Server (NTRS)

Andre, Matthieu A.; Bardet, Philippe M.; Burns, Ross A.; Danehy, Paul M.

2015-01-01

The High Temperature Test Facility, HTTF, at Oregon State University (OSU) is an integral-effect test facility designed to model the behavior of a Very High Temperature Gas Reactor (VHTR) during a Depressurized Conduction Cooldown (DCC) event. It also has the ability to conduct limited investigations into the progression of a Pressurized Conduction Cooldown (PCC) event in addition to phenomena occurring during normal operations. Both of these phenomena will be studied with in-situ velocity field measurements. Experimental measurements of velocity are critical to provide proper boundary conditions to validate CFD codes, as well as developing correlations for system level codes, such as RELAP5 (http://www4vip.inl.gov/relap5/). Such data will be the first acquired in the HTTF and will introduce a diagnostic with numerous other applications to the field of nuclear thermal hydraulics. A laser-based optical diagnostic under development at The George Washington University (GWU) is presented; the technique is demonstrated with velocity data obtained in ambient temperature air, and adaptation to high-pressure, high-temperature flow is discussed.

Condensation model for the ESBWR passive condensers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Revankar, S. T.; Zhou, W.; Wolf, B.

2012-07-01

In the General Electric's Economic simplified boiling water reactor (GE-ESBWR) the passive containment cooling system (PCCS) plays a major role in containment pressure control in case of an loss of coolant accident. The PCCS condenser must be able to remove sufficient energy from the reactor containment to prevent containment from exceeding its design pressure following a design basis accident. There are three PCCS condensation modes depending on the containment pressurization due to coolant discharge; complete condensation, cyclic venting and flow through mode. The present work reviews the models and presents model predictive capability along with comparison with existing data frommore » separate effects test. The condensation models in thermal hydraulics code RELAP5 are also assessed to examine its application to various flow modes of condensation. The default model in the code predicts complete condensation well, and basically is Nusselt solution. The UCB model predicts through flow well. None of condensation model in RELAP5 predict complete condensation, cyclic venting, and through flow condensation consistently. New condensation correlations are given that accurately predict all three modes of PCCS condensation. (authors)« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE PAGES

Piro, M.H.A; Wassermann, F.; Grundmann, S.; ...

2017-05-23

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

MELCOR simulations of the severe accident at Fukushima Daiichi Unit 3

DOE PAGES

Cardoni, Jeffrey; Gauntt, Randall; Kalinich, Donald; ...

2014-05-01

In response to the accident at the Fukushima Daiichi nuclear power station in Japan, the U.S. Nuclear Regulatory Commission and U.S. Department of Energy agreed to jointly sponsor an accident reconstruction study as a means of assessing the severe accident modeling capability of the MELCOR code. Objectives of the project included reconstruction of the accident progressions using computer models and accident data, and validation of the MELCOR code and the Fukushima models against plant data. A MELCOR 2.1 model of the Fukushima Daiichi Unit 3 reactor is developed using plant-specific information and accident-specific boundary conditions, which involve considerable uncertainty duemore » to the inherent nature of severe accidents. Publicly available thermal-hydraulic data and radioactivity release estimates have evolved significantly since the accidents. Such data are expected to continually change as the reactors are decommissioned and more measurements are performed. As a result, the MELCOR simulations in this work primarily use boundary conditions that are based on available plant data as of May 2012.« less

Fluid Flow Investigations within a 37 Element CANDU Fuel Bundle Supported by Magnetic Resonance Velocimetry and Computational Fluid Dynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Piro, M.H.A; Wassermann, F.; Grundmann, S.

The current work presents experimental and computational investigations of fluid flow through a 37 element CANDU nuclear fuel bundle. Experiments based on Magnetic Resonance Velocimetry (MRV) permit three-dimensional, three-component fluid velocity measurements to be made within the bundle with sub-millimeter resolution that are non-intrusive, do not require tracer particles or optical access of the flow field. Computational fluid dynamic (CFD) simulations of the foregoing experiments were performed with the hydra-th code using implicit large eddy simulation, which were in good agreement with experimental measurements of the fluid velocity. Greater understanding has been gained in the evolution of geometry-induced inter-subchannel mixing,more » the local effects of obstructed debris on the local flow field, and various turbulent effects, such as recirculation, swirl and separation. These capabilities are not available with conventional experimental techniques or thermal-hydraulic codes. Finally, the overall goal of this work is to continue developing experimental and computational capabilities for further investigations that reliably support nuclear reactor performance and safety.« less

Extending a CAD-Based Cartesian Mesh Generator for the Lattice Boltzmann Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Cantrell, J Nathan; Inclan, Eric J; Joshi, Abhijit S

2012-01-01

This paper describes the development of a custom preprocessor for the PaRAllel Thermal Hydraulics simulations using Advanced Mesoscopic methods (PRATHAM) code based on an open-source mesh generator, CartGen [1]. PRATHAM is a three-dimensional (3D) lattice Boltzmann method (LBM) based parallel flow simulation software currently under development at the Oak Ridge National Laboratory. The LBM algorithm in PRATHAM requires a uniform, coordinate system-aligned, non-body-fitted structured mesh for its computational domain. CartGen [1], which is a GNU-licensed open source code, already comes with some of the above needed functionalities. However, it needs to be further extended to fully support the LBM specificmore » preprocessing requirements. Therefore, CartGen is being modified to (i) be compiler independent while converting a neutral-format STL (Stereolithography) CAD geometry to a uniform structured Cartesian mesh, (ii) provide a mechanism for PRATHAM to import the mesh and identify the fluid/solid domains, and (iii) provide a mechanism to visually identify and tag the domain boundaries on which to apply different boundary conditions.« less

Development of fission-products transport model in severe-accident scenarios for Scdap/Relap5

NASA Astrophysics Data System (ADS)

Honaiser, Eduardo Henrique Rangel

The understanding and estimation of the release of fission products during a severe accident became one of the priorities of the nuclear community after 1980, with the events of the Three-mile Island unit 2 (TMI-2), in 1979, and Chernobyl accidents, in 1986. Since this time, theoretical developments and experiments have shown that the primary circuit systems of light water reactors (LWR) have the potential to attenuate the release of fission products, a fact that had been neglected before. An advanced tool, compatible with nuclear thermal-hydraulics integral codes, is developed to predict the retention and physical evolution of the fission products in the primary circuit of LWRs, without considering the chemistry effects. The tool embodies the state-of-the-art models for the involved phenomena as well as develops new models. The capabilities acquired after the implementation of this tool in the Scdap/Relap5 code can be used to increase the accuracy of probability safety assessment (PSA) level 2, enhance the reactor accident management procedures and design new emergency safety features.

Design and Construction of Experiment for Direct Electron Irradiation of Uranyl Sulfate Solution: Bubble Formation and Thermal Hydraulics Studies

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chemerisov, Sergey; Gromov, Roman; Makarashvili, Vakho

Argonne is assisting SHINE Medical Technologies in developing SHINE, a system for producing fission-product 99Mo using a D/T-accelerator to produce fission in a non-critical target solution of aqueous uranyl sulfate. We have developed an experimental setup for studying thermal-hydraulics and bubble formation in the uranyl sulfate solution to simulate conditions expected in the SHINE target solution during irradiation. A direct electron beam from the linac accelerator will be used to irradiate a 20 L solution (sector of the solution vessel). Because the solution will undergo radiolytic decomposition, we will be able to study bubble formation and dynamics and effects ofmore » convection and temperature on bubble behavior. These experiments will serve as a verification/ validation tool for the thermal-hydraulic model. Utilization of the direct electron beam for irradiation allows homogeneous heating of a large solution volume and simplifies observation of the bubble dynamics simultaneously with thermal-hydraulic data collection, which will complement data collected during operation of the miniSHINE experiment. Irradiation will be conducted using a 30-40 MeV electron beam from the high-power linac accelerator. The total electron-beam power will be 20 kW, which will yield a power density on the order of 1 kW/L. The solution volume will be cooled on the front and back surfaces and central tube to mimic the geometry of the proposed SHINE solution vessel. Also, multiple thermocouples will be inserted into the solution vessel to map thermal profiles. The experimental design is now complete, and installation and testing are in progress.« less

Experimental investigation of the hydraulic and heat-transfer properties of artificially fractured granite.

PubMed

Luo, Jin; Zhu, Yongqiang; Guo, Qinghai; Tan, Long; Zhuang, Yaqin; Liu, Mingliang; Zhang, Canhai; Xiang, Wei; Rohn, Joachim

2017-01-05

In this paper, the hydraulic and heat-transfer properties of two sets of artificially fractured granite samples are investigated. First, the morphological information is determined using 3D modelling technology. The area ratio is used to describe the roughness of the fracture surface. Second, the hydraulic properties of fractured granite are tested by exposing samples to different confining pressures and temperatures. The results show that the hydraulic properties of the fractures are affected mainly by the area ratio, with a larger area ratio producing a larger fracture aperture and higher hydraulic conductivity. Both the hydraulic apertureand the hydraulic conductivity decrease with an increase in the confining pressure. Furthermore, the fracture aperture decreases with increasing rock temperature, but the hydraulic conductivity increases owing to a reduction of the viscosity of the fluid flowing through. Finally, the heat-transfer efficiency of the samples under coupled hydro-thermal-mechanical conditions is analysed and discussed.

Experimental investigation of the hydraulic and heat-transfer properties of artificially fractured granite

PubMed Central

Luo, Jin; Zhu, Yongqiang; Guo, Qinghai; Tan, Long; Zhuang, Yaqin; Liu, Mingliang; Zhang, Canhai; Xiang, Wei; Rohn, Joachim

2017-01-01

In this paper, the hydraulic and heat-transfer properties of two sets of artificially fractured granite samples are investigated. First, the morphological information is determined using 3D modelling technology. The area ratio is used to describe the roughness of the fracture surface. Second, the hydraulic properties of fractured granite are tested by exposing samples to different confining pressures and temperatures. The results show that the hydraulic properties of the fractures are affected mainly by the area ratio, with a larger area ratio producing a larger fracture aperture and higher hydraulic conductivity. Both the hydraulic apertureand the hydraulic conductivity decrease with an increase in the confining pressure. Furthermore, the fracture aperture decreases with increasing rock temperature, but the hydraulic conductivity increases owing to a reduction of the viscosity of the fluid flowing through. Finally, the heat-transfer efficiency of the samples under coupled hydro-thermal-mechanical conditions is analysed and discussed. PMID:28054594

Groundwater flow modeling of periods with periglacial and glacial climate conditions for the safety assessment of the proposed high-level nuclear waste repository site at Forsmark, Sweden

NASA Astrophysics Data System (ADS)

Vidstrand, Patrik; Follin, Sven; Selroos, Jan-Olof; Näslund, Jens-Ove

2014-09-01

The impact of periglacial and glacial climate conditions on groundwater flow in fractured crystalline rock is studied by means of groundwater flow modeling of the Forsmark site, which was recently proposed as a repository site for the disposal of spent high-level nuclear fuel in Sweden. The employed model uses a thermal-hydraulically coupled approach for permafrost modeling and discusses changes in groundwater flow implied by the climate conditions found over northern Europe at different times during the last glacial cycle (Weichselian glaciation). It is concluded that discharge of particles released at repository depth occurs very close to the ice-sheet margin in the absence of permafrost. If permafrost is included, the greater part discharges into taliks in the periglacial area. During a glacial cycle, hydraulic gradients at repository depth reach their maximum values when the ice-sheet margin passes over the site; at this time, also, the interface between fresh and saline waters is distorted the most. The combined effect of advances and retreats during several glaciations has not been studied in the present work; however, the results indicate that hydrochemical conditions at depth in the groundwater flow model are almost restored after a single event of ice-sheet advance and retreat.

X-Antenna: A graphical interface for antenna analysis codes

NASA Technical Reports Server (NTRS)

Goldstein, B. L.; Newman, E. H.; Shamansky, H. T.

1995-01-01

This report serves as the user's manual for the X-Antenna code. X-Antenna is intended to simplify the analysis of antennas by giving the user graphical interfaces in which to enter all relevant antenna and analysis code data. Essentially, X-Antenna creates a Motif interface to the user's antenna analysis codes. A command-file allows new antennas and codes to be added to the application. The menu system and graphical interface screens are created dynamically to conform to the data in the command-file. Antenna data can be saved and retrieved from disk. X-Antenna checks all antenna and code values to ensure they are of the correct type, writes an output file, and runs the appropriate antenna analysis code. Volumetric pattern data may be viewed in 3D space with an external viewer run directly from the application. Currently, X-Antenna includes analysis codes for thin wire antennas (dipoles, loops, and helices), rectangular microstrip antennas, and thin slot antennas.

Surface Cracking and Interface Reaction Associated Delamination Failure of Thermal and Environmental Barrier Coatings

NASA Technical Reports Server (NTRS)

Zhu, Dong-Ming; Choi, Sung R.; Eldridge, Jeffrey I.; Lee, Kang N.; Miller, Robert A.

2003-01-01

In this paper, surface cracking and interface reactions of a BSAS coating and a multi-layer ZrO2-8wt%Y2O3 and mullite/BSAS/Si thermal and environmental barrier coating system on SiC/SiC ceramic matrix composites were characterized after long-term combined laser thermal gradient and furnace cyclic tests in a water vapor containing environment. The surface cracking was analyzed based on the coating thermal gradient sintering behavior and thermal expansion mismatch stress characteristics under the thermal cyclic conditions. The interface reactions, which were largely enhanced by the coating surface cracking in the water vapor environment, were investigated in detail, and the reaction phases were identified for the coating system after the long-term exposure. The accelerated coating delamination failure was attributed to the increased delamination driving force under the thermal gradient cyclic loading and the reduced interface adhesion due to the detrimental interface reactions.

Surface Cracking and Interface Reaction Associated Delamination Failure of Thermal and Environmental Barrier Coatings

NASA Technical Reports Server (NTRS)

Zhu, Dongming; Choi, Sung R.; Eldridge, Jeffrey I.; Lee, Kang N.; Miller, Robert A.

2003-01-01

In this paper, surface cracking and interface reactions of a BSAS coating and a multi-layer ZTO2-8wt%Y2O3 and mullite/BSAS/Si thermal and environmental barrier coating system on SiC/SiC ceramic matrix composites were characterized after long-term combined laser thermal gradient and furnace cyclic tests in a water vapor containing environment. The surface cracking was analyzed based on the coating thermal gradient sintering behavior and thermal expansion mismatch stress characteristics under the thermal cyclic conditions. The interface reactions, which were largely enhanced by the coating surface cracking in the water vapor environment, were investigated in detail, and the reaction phases were identified for the coating system after the long- term exposure. The accelerated coating delamination failure was attributed to the increased delamination driving force under the thermal gradient cyclic loading and the reduced interface adhesion due to the detrimental interface reactions.

Thermal conductance at atomically clean and disordered silicon/aluminum interfaces: A molecular dynamics simulation study

NASA Astrophysics Data System (ADS)

Ih Choi, Woon; Kim, Kwiseon; Narumanchi, Sreekant

2012-09-01

Thermal resistance between layers impedes effective heat dissipation in electronics packaging applications. Thermal conductance for clean and disordered interfaces between silicon (Si) and aluminum (Al) was computed using realistic Si/Al interfaces and classical molecular dynamics with the modified embedded atom method potential. These realistic interfaces, which include atomically clean as well as disordered interfaces, were obtained using density functional theory. At 300 K, the magnitude of interfacial conductance due to phonon-phonon scattering obtained from the classical molecular dynamics simulations was approximately five times higher than the conductance obtained using analytical elastic diffuse mismatch models. Interfacial disorder reduced the thermal conductance due to increased phonon scattering with respect to the atomically clean interface. Also, the interfacial conductance, due to electron-phonon scattering at the interface, was greater than the conductance due to phonon-phonon scattering. This indicates that phonon-phonon scattering is the bottleneck for interfacial transport at the semiconductor/metal interfaces. The molecular dynamics modeling predictions for interfacial thermal conductance for a 5-nm disordered interface between Si/Al were in-line with recent experimental data in the literature.

Numerical modeling of consolidation processes in hydraulically deposited soils

NASA Astrophysics Data System (ADS)

Brink, Nicholas Robert

Hydraulically deposited soils are encountered in many common engineering applications including mine tailing and geotextile tube fills, though the consolidation process for such soils is highly nonlinear and requires the use of advanced numerical techniques to provide accurate predictions. Several commercially available finite element codes poses the ability to model soil consolidation, and it was the goal of this research to assess the ability of two of these codes, ABAQUS and PLAXIS, to model the large-strain, two-dimensional consolidation processes which occur in hydraulically deposited soils. A series of one- and two-dimensionally drained rectangular models were first created to assess the limitations of ABAQUS and PLAXIS when modeling consolidation of highly compressible soils. Then, geotextile tube and TSF models were created to represent actual scenarios which might be encountered in engineering practice. Several limitations were discovered, including the existence of a minimum preconsolidation stress below which numerical solutions become unstable.

Calculation of the Phenix end-of-life test 'Control Rod Withdrawal' with the ERANOS code

DOE Office of Scientific and Technical Information (OSTI.GOV)

Tiberi, V.

2012-07-01

The Inst. of Radiological Protection and Nuclear Safety (IRSN) acts as technical support to French public authorities. As such, IRSN is in charge of safety assessment of operating and under construction reactors, as well as future projects. In this framework, one current objective of IRSN is to evaluate the ability and accuracy of numerical tools to foresee consequences of accidents. Neutronic studies step in the safety assessment from different points of view among which the core design and its protection system. They are necessary to evaluate the core behavior in case of accident in order to assess the integrity ofmore » the first barrier and the absence of a prompt criticality risk. To reach this objective one main physical quantity has to be evaluated accurately: the neutronic power distribution in core during whole reactor lifetime. Phenix end of life tests, carried out in 2009, aim at increasing the experience feedback on sodium cooled fast reactors. These experiments have been done in the framework of the development of the 4. generation of nuclear reactors. Ten tests have been carried out: 6 on neutronic and fuel aspects, 2 on thermal hydraulics and 2 for the emergency shutdown. Two of them have been chosen for an international exercise on thermal hydraulics and neutronics in the frame of an IAEA Coordinated Research Project. Concerning neutronics, the Control Rod Withdrawal test is relevant for safety because it allows evaluating the capability of calculation tools to compute the radial power distribution on fast reactors core configurations in which the flux field is very deformed. IRSN participated to this benchmark with the ERANOS code developed by CEA for fast reactors studies. This paper presents the results obtained in the framework of the benchmark activity. A relatively good agreement was found with available measures considering the approximations done in the modeling. The work underlines the importance of burn-up calculations in order to have a fine core concentrations mesh for the calculation of the power distribution. (authors)« less

A linearized microstructural model for hydraulic conductivity evolution due to brittle damage: application to Hydraulic Fracturing treatments

NASA Astrophysics Data System (ADS)

Caramiello, G.; Montanino, A.; Della Vecchia, G., Sr.; Pandolfi, A., Sr.

2017-12-01

Among the features of geological structures, fractures and discontinuities play a dominant role, due to their significant influence on both the hydraulic and the mechanical behavior of the rock mass. Despite the current availability of fault and fracture mappings, the understanding of the influence of faults on fluid flow is nowadays not satisfactory, in particular when hydro-mechanical coupling is significant. In engineering technology fracture processes are often exploited. Hydraulic fracturing is one of the most important example. Hydraulic fracturing is a process characterized by the inception and propagation of fractures as a consequence of a hydraulic driven solicitation and it is used to improve the production and optimize well stimulation in low permeability reservoirs. Due to the coupling of several different phenomena (hydro-thermo-chemical coupling) there is not a reliable complete mathematical model able to simulate in a proper way the process. To design hydraulic fracturing treatments, it is necessary to predict the growth of fracture geometry as a function of treatment parameters. In this contribution we present a recently developed model of brittle damage of confined rock masses, with particular emphasis on the influence of mechanical damage on the evolution of porosity and permeability. The model is based on an explicit micromechanical construction of connected patterns of parallel equi-spaced cracks. A relevant feature of the model is that the fracture patterns are not arbitrary, but their inception, orientation and spacing follow from energetic consideration. The model, based on the Terzaghi effective stress concepts, has been then implemented into a coupled hydro-mechanical finite element code, where the linear momentum and the fluid mass balance equations are numerically solved via a staggered approach. The coupled code is used to simulate fracturing processes induced by an increase in pore pressure. The examples show the capability of the model in reproducing three-dimensional multiscale complex fracture patterns and permeability enhancement in the damaged porous medium. The numerical code, has been used to verify the influence of the distance between the different perforation slots as well of the wellbore-deviation from the minimum stress axis on the propagation of multiple.

Interface Shape Control Using Localized Heating during Bridgman Growth

NASA Technical Reports Server (NTRS)

Volz, M. P.; Mazuruk, K.; Aggarwal, M. D.; Croll, A.

2008-01-01

Numerical calculations were performed to assess the effect of localized radial heating on the melt-crystal interface shape during vertical Bridgman growth. System parameters examined include the ampoule, melt and crystal thermal conductivities, the magnitude and width of localized heating, and the latent heat of crystallization. Concave interface shapes, typical of semiconductor systems, could be flattened or made convex with localized heating. Although localized heating caused shallower thermal gradients ahead of the interface, the magnitude of the localized heating required for convexity was less than that which resulted in a thermal inversion ahead of the interface. A convex interface shape was most readily achieved with ampoules of lower thermal conductivity. Increasing melt convection tended to flatten the interface, but the amount of radial heating required to achieve a convex interface was essentially independent of the convection intensity.

Imaging hydraulic fractures using temperature transients in the Belridge Diatomite

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shahin, G.T.; Johnston, R.M.

1995-12-31

Results of a temperature transient analysis of Shell`s Phase 1 and Phase 2 Diatomite Steamdrive Pilots are used to image hydraulic injection fracture lengths, angles, and heat injectivities into the low-permeability formation. The Phase 1 Pilot is a limited-interval injection test. In Phase 2, steam is injected into two 350 ft upper and lower zones through separate hydraulic fractures. Temperature response of both pilots is monitored with sixteen logging observation wells. A perturbation analysis of the non-linear pressure diffusion and heat transport equations indicates that at a permeability of about 0.1 md or less, heat transport in the Diatomite tendsmore » to be dominated by thermal diffusivity, and pressure diffusion is dominated by the ratio of thermal expansion to fluid compressibility. Under these conditions, the temperature observed at a logging observation well is governed by a dimensionless quantity that depends on the perpendicular distance between the observation well and the hydraulic fracture, divided by the square root of time. Using this dependence, a novel method is developed for imaging hydraulic fracture geometry and relative heat injectivity from the temperature history of the pilot.« less

A fault constitutive relation accounting for thermal pressurization of pore fluid

USGS Publications Warehouse

Andrews, D.J.

2002-01-01

The heat generated in a slip zone during an earthquake can raise fluid pressure and thereby reduce frictional resistance to slip. The amount of fluid pressure rise depends on the associated fluid flow. The heat generated at a given time produces fluid pressure that decreases inversely with the square root of hydraulic diffusivity times the elapsed time. If the slip velocity function is crack-like, there is a prompt fluid pressure rise at the onset of slip, followed by a slower increase. The stress drop associated with the prompt fluid pressure rise increases with rupture propagation distance. The threshold propagation distance at which thermally induced stress drop starts to dominate over frictionally induced stress drop is proportional to hydraulic diffusivity. If hydraulic diffusivity is 0.02 m2/s, estimated from borehole samples of fault zone material, the threshold propagation distance is 300 m. The stress wave in an earthquake will induce an unknown amount of dilatancy and will increase hydraulic diffusivity, both of which will lessen the fluid pressure effect. Nevertheless, if hydraulic diffusivity is no more than two orders of magnitude larger than the laboratory value, then stress drop is complete in large earthquakes.

Preliminary study on new configuration with LEU fuel assemblies for the Dalat nuclear research reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Van Lam Pham; Vinh Vinh Le; Ton Nghiem Huynh

2008-07-15

The fuel conversion of the Dalat Nuclear Research Reactor (DNRR) is being realized. The DNRR is a pool type research reactor which was reconstructed from the 250 kW TRIGA- MARK II reactor. The reconstructed reactor attained its nominal power of 500 kW in February 1984. According to the results of design and safety analyses performed by the joint study between RERTR Program at Argonne National Laboratory (ANL) and Vietnam Atomic Energy Commission (VAEC) the mixed core of irradiated HEU and new LEU WWR-M2 fuel assemblies will be created soon. This paper presents the results of preliminary study on new configurationmore » with only LEU fuel assemblies for the DNRR. The codes MCNP, REBUS and VARI3D are used to calculate neutron flux performance in irradiation positions and kinetics parameters. The idea of change of Beryllium rod reloading enables to get working configuration assured shutdown margin, thermal-hydraulic safety and increase in thermal neutron flux in neutron trap at the center of DNRR active core. (author)« less

The development of an intelligent interface to a computational fluid dynamics flow-solver code

NASA Technical Reports Server (NTRS)

Williams, Anthony D.

1988-01-01

Researchers at NASA Lewis are currently developing an 'intelligent' interface to aid in the development and use of large, computational fluid dynamics flow-solver codes for studying the internal fluid behavior of aerospace propulsion systems. This paper discusses the requirements, design, and implementation of an intelligent interface to Proteus, a general purpose, 3-D, Navier-Stokes flow solver. The interface is called PROTAIS to denote its introduction of artificial intelligence (AI) concepts to the Proteus code.

The development of an intelligent interface to a computational fluid dynamics flow-solver code

NASA Technical Reports Server (NTRS)

Williams, Anthony D.

1988-01-01

Researchers at NASA Lewis are currently developing an 'intelligent' interface to aid in the development and use of large, computational fluid dynamics flow-solver codes for studying the internal fluid behavior of aerospace propulsion systems. This paper discusses the requirements, design, and implementation of an intelligent interface to Proteus, a general purpose, three-dimensional, Navier-Stokes flow solver. The interface is called PROTAIS to denote its introduction of artificial intelligence (AI) concepts to the Proteus code.

Flight evaluation of Spacelab 1 payload thermal/ECS interfaces

NASA Technical Reports Server (NTRS)

Ray, C. D.; Humphries, W. R.; Patterson, W. C.

1984-01-01

The Spacelab (SL-1) thermal/Environmental Control Systems (ECS) are discussed. Preflight analyses and flight data are compared in order to validate payload to Spacelab interfaces as well as corroborate modeling/analysis techniques. In doing so, a brief description of the Spacelab 1 payload configuration and the interactive Spacelab thermal/ECS systems are given. In particular, these interfaces address equipment cooling air, thermal and fluid conditions, humidity levels, both freon and water loop temperatures and load states, as well as passive radiant environment interfaces.

Reduction of thermal stresses in continuous fiber reinforced metal matrix composites with interface layers

NASA Technical Reports Server (NTRS)

Jansson, S.; Leckie, F. A.

1990-01-01

The potential of using an interface layer to reduce thermal stresses in the matrix of composites with a mismatch in coefficients of thermal expansion of fiber and matrix was investigated. It was found that compliant layers, with properties of readily available materials, do not have the potential to reduce thermal stresses significantly. However, interface layers with high coefficient of thermal expansion can compensate for the mismatch and reduce thermal stresses in the matrix significantly.

Analysis of ITER NbTi and Nb3Sn CICCs experimental minimum quench energy with JackPot, MCM and THEA models

NASA Astrophysics Data System (ADS)

Bagni, T.; Duchateau, J. L.; Breschi, M.; Devred, A.; Nijhuis, A.

2017-09-01

Cable-in-conduit conductors (CICCs) for ITER magnets are subjected to fast changing magnetic fields during the plasma-operating scenario. In order to anticipate the limitations of conductors under the foreseen operating conditions, it is essential to have a better understanding of the stability margin of magnets. In the last decade ITER has launched a campaign for characterization of several types of NbTi and Nb3Sn CICCs comprising quench tests with a singular sine wave fast magnetic field pulse and relatively small amplitude. The stability tests, performed in the SULTAN facility, were reproduced and analyzed using two codes: JackPot-AC/DC, an electromagnetic-thermal numerical model for CICCs, developed at the University of Twente (van Lanen and Nijhuis 2010 Cryogenics 50 139-148) and multi-constant-model (MCM) (Turck and Zani 2010 Cryogenics 50 443-9), an analytical model for CICCs coupling losses. The outputs of both codes were combined with thermal, hydraulic and electric analysis of superconducting cables to predict the minimum quench energy (MQE) (Bottura et al 2000 Cryogenics 40 617-26). The experimental AC loss results were used to calibrate the JackPot and MCM models and to reproduce the energy deposited in the cable during an MQE test. The agreement between experiments and models confirm a good comprehension of the various CICCs thermal and electromagnetic phenomena. The differences between the analytical MCM and numerical JackPot approaches are discussed. The results provide a good basis for further investigation of CICC stability under plasma scenario conditions using magnetic field pulses with lower ramp rate and higher amplitude.

Evolution of Cement-Casing Interface in Wellbore Microannuli under Stress

NASA Astrophysics Data System (ADS)

Matteo, E. N.; Gomez, S. P.; Sobolik, S. R.; Taha, M. R.; Stormont, J.

2017-12-01

Laboratory tests measured the compressibility and flow characteristics of wellbore microannuli. Specimens, consisting of a cement sheath cast on a steel casing with microannuli, were subjected to confining pressures and casing pressures in a pressure vessel that allows simultaneous measurement of gas flow along the axis of the specimen. The flow was interpreted as the hydraulic aperture of the microannuli. We found the hydraulic aperture decreases as confining stress is increased. The larger the initial hydraulic aperture, the more it decreases as confining stress increases. The changes in measured hydraulic aperture correspond to changes of many orders of magnitude in permeability of the wellbore system, suggesting that microannulus response to stress changes may have a significant impact on estimates of wellbore leakage. A finite element model of a wellbore system was developed that included elements representing the microannulus that incorporated the hyperbolic joint model. The thickness of the microannulus elements is equivalent to the hydraulic aperture. The calculated normal stress across the microannulus used in the numerical implementation was found to be similar to the applied confining pressure in the laboratory tests. The microannulus elements were found to reasonably reproduce laboratory behavior during loading from confining pressure increases. The calculated microannulus response to internal casing pressure changes was less stiff than measured, which may be due to hardening of the microannulus during testing. In particular, the microannulus model could be used to estimate CO2 leakage as a function of formation stress changes and/or displacements, or loading from casing expansion or contraction during wellbore operations. Recommendations for future work include an application of the joint model with a thermally active large-scale reservoir coupled with pore pressure caused by dynamic CO2 injection and subsequent microannulus region affects. Sandia National Laboratories is a multimission laboratory managed and operated by National Technology and 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-NA-0003525. SAND2017-8090 A.

L3.PHI.CTF.P10.02-rev2 Coupling of Subchannel T/H (CTF) and CRUD Chemistry (MAMBA1D)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Salko, Robert K.; Palmtag, Scott; Collins, Benjamin S.

2015-05-15

The purpose of this milestone is to create a preliminary capability for modeling light water reactor (LWR) thermal-hydraulic (T/H) and CRUD growth using the CTF subchannel code and the subgrid version of the MAMBA CRUD chemistry code, MAMBA1D. In part, this is a follow-on to Milestone L3.PHI.VCS.P9.01, which is documented in Report CASL-U-2014-0188-000, titled "Development of CTF Capability for Modeling Reactor Operating Cycles with Crud Growth". As the title suggests, the previous milestone set up a framework for modeling reactor operation cycles with CTF. The framework also facilitated coupling to a CRUD chemistry capability for modeling CRUD growth throughout themore » reactor operating cycle. To demonstrate the capability, a simple CRUD \\surrogate" tool was developed and coupled to CTF; however, it was noted that CRUD growth predictions by the surrogate were not considered realistic. This milestone builds on L3.PHI.VCS.P9.01 by replacing this simple surrogate tool with the more advanced MAMBA1D CRUD chemistry code. Completing this task involves addressing unresolved tasks from Milestone L3.PHI.VCS.P9.01, setting up an interface to MAMBA1D, and extracting new T/H information from CTF that was not previously required in the simple surrogate tool. Speci c challenges encountered during this milestone include (1) treatment of the CRUD erosion model, which requires local turbulent kinetic energy (TKE) (a value that CTF does not calculate) and (2) treatment of the MAMBA1D CRUD chimney boiling model in the CTF rod heat transfer solution. To demonstrate this new T/H, CRUD modeling capability, two sets of simulations were performed: (1) an 18 month cycle simulation of a quarter symmetry model of Watts Bar and (2) a simulation of Assemblies G69 and G70 from Seabrook Cycle 5. The Watts Bar simulation is merely a demonstration of the capability. The simulation of the Seabrook cycle, which had experienced CRUD-related fuel rod failures, had actual CRUD-scrape data to compare with results. As results show, the initial CTF/MAMBA1D-predicted CRUD thicknesses were about half of their expected values, so further investigation will be required for this simulation.« less

Thermal flux limited electron Kapitza conductance in copper-niobium multilayers

DOE PAGES

Cheaito, Ramez; Hattar, Khalid Mikhiel; Gaskins, John T.; ...

2015-03-05

The interplay between the contributions of electron thermal flux and interface scattering to the Kapitza conductance across metal-metal interfaces through measurements of thermal conductivity of copper-niobium multilayers was studied. Thermal conductivities of copper-niobium multilayer films of period thicknesses ranging from 5.4 to 96.2 nm and sample thicknesses ranging from 962 to 2677 nm are measured by time-domain thermoreflectance over a range of temperatures from 78 to 500 K. The Kapitza conductances between the Cu and Nb interfaces in multilayer films are determined from the thermal conductivities using a series resistor model and are in good agreement with the electron diffusemore » mismatch model. The results for the thermal boundary conductance between Cu and Nb are compared to literature values for the thermal boundary conductance across Al-Cu and Pd-Ir interfaces, and demonstrate that the interface conductance in metallic systems is dictated by the temperature derivative of the electron energy flux in the metallic layers, rather than electron mean free path or scattering processes at the interface.« less

ELM - A SIMPLE TOOL FOR THERMAL-HYDRAULIC ANALYSIS OF SOLID-CORE NUCLEAR ROCKET FUEL ELEMENTS

NASA Technical Reports Server (NTRS)

Walton, J. T.

1994-01-01

ELM is a simple computational tool for modeling the steady-state thermal-hydraulics of propellant flow through fuel element coolant channels in nuclear thermal rockets. Written for the nuclear propulsion project of the Space Exploration Initiative, ELM evaluates the various heat transfer coefficient and friction factor correlations available for turbulent pipe flow with heat addition. In the past, these correlations were found in different reactor analysis codes, but now comparisons are possible within one program. The logic of ELM is based on the one-dimensional conservation of energy in combination with Newton's Law of Cooling to determine the bulk flow temperature and the wall temperature across a control volume. Since the control volume is an incremental length of tube, the corresponding pressure drop is determined by application of the Law of Conservation of Momentum. The size, speed, and accuracy of ELM make it a simple tool for use in fuel element parametric studies. ELM is a machine independent program written in FORTRAN 77. It has been successfully compiled on an IBM PC compatible running MS-DOS using Lahey FORTRAN 77, a DEC VAX series computer running VMS, and a Sun4 series computer running SunOS UNIX. ELM requires 565K of RAM under SunOS 4.1, 360K of RAM under VMS 5.4, and 406K of RAM under MS-DOS. Because this program is machine independent, no executable is provided on the distribution media. The standard distribution medium for ELM is one 5.25 inch 360K MS-DOS format diskette. ELM was developed in 1991. DEC, VAX, and VMS are trademarks of Digital Equipment Corporation. Sun4 and SunOS are trademarks of Sun Microsystems, Inc. IBM PC is a registered trademark of International Business Machines. MS-DOS is a registered trademark of Microsoft Corporation.

The Influence of a TiN Film on the Electronic Contribution to the Thermal Conductivity of a TiC Film in a TiN-TiC Layer System

NASA Astrophysics Data System (ADS)

Jagannadham, K.

2018-01-01

TiC and TiN films were deposited by reactive magnetron sputtering on Si substrates. Scanning electron microscopy (SEM) and transmission electron microscopy (TEM) characterization of the microstructure and interface structure have been carried out and the stoichiometric composition of TiC is determined. Thermal conductivity and interface thermal conductance between different layers in the films are evaluated by the transient thermo reflectance (TTR) and three-omega (3- ω) methods. The results showed that the thermal conductivity of the TiC films increased with temperature. The thermal conductivity of TiC in the absence of TiN is dominated by phonon contribution. The electronic contribution to the thermal conductivity of TiC in the presence of TiN is found to be more significant. The interface thermal conductance of the TiC/TiN interface is much larger than that of interfaces at Au/TiC, TiC/Si, or TiN/Si. The interface thermal conductance between TiC and TiN is reduced by the layer formed as a result of interdiffusion.

Numerical Optimization of the Thermal Field in Bridgman Detached Growth

NASA Technical Reports Server (NTRS)

Stelian, C.; Volz, M. P.; Derby, J. J.

2009-01-01

The global modeling of the thermal field in two vertical Bridgman-like crystal growth configurations, has been performed to get optimal thermal conditions for a successful detached growth of Ge and CdTe crystals. These computations are performed using the CrysMAS code and expand upon our previous analysis [1] that propose a new mechanism involving the thermal field and meniscus position to explain stable conditions for dewetted Bridgman growth. The analysis of the vertical Bridgman configuration with two heaters, used by Palosz et al. for the detached growth of Ge, shows, consistent with their results, that the large wetting angle of germanium on boron nitride surfaces was an important factor to promote a successful detached growth. Our computations predict that by initiating growth much higher into the hot zone of the furnace, the thermal conditions will be favorable for continued detachment even for systems that did not exhibit high contact angles. The computations performed for a vertical gradient freeze configuration with three heaters representative of that used for the detached growth of CdTe, show favorable thermal conditions for dewetting during the entirely growth run described. Improved thermal conditions are also predicted for coated silica crucibles when the solid-liquid interface advances higher into the hot zone during the solidification process. The second set of experiments on CdTe growth described elsewhere has shown the reattachment of the crystal to the crucible after few centimeters of dewetted growth. The thermal modeling of this configuration shows a second solidification front appearing at the top of the sample and approaching the middle line across the third heater. In these conditions, the crystal grows detached from the bottom, but will be attached to the crucible in the upper part because of the solidification without gap in this region. The solidification with two interfaces can be avoided when the top of the sample is positioned below the middle position of the third furnace.

Rocketdyne/Westinghouse nuclear thermal rocket engine modeling

NASA Technical Reports Server (NTRS)

Glass, James F.

1993-01-01

The topics are presented in viewgraph form and include the following: systems approach needed for nuclear thermal rocket (NTR) design optimization; generic NTR engine power balance codes; rocketdyne nuclear thermal system code; software capabilities; steady state model; NTR engine optimizer code-logic; reactor power calculation logic; sample multi-component configuration; NTR design code output; generic NTR code at Rocketdyne; Rocketdyne NTR model; and nuclear thermal rocket modeling directions.

The Python Sky Model: software for simulating the Galactic microwave sky

NASA Astrophysics Data System (ADS)

Thorne, B.; Dunkley, J.; Alonso, D.; Næss, S.

2017-08-01

We present a numerical code to simulate maps of Galactic emission in intensity and polarization at microwave frequencies, aiding in the design of cosmic microwave background experiments. This python code builds on existing efforts to simulate the sky by providing an easy-to-use interface and is based on publicly available data from the WMAP (Wilkinson Microwave Anisotropy Probe) and Planck satellite missions. We simulate synchrotron, thermal dust, free-free and anomalous microwave emission over the whole sky, in addition to the cosmic microwave background, and include a set of alternative prescriptions for the frequency dependence of each component, for example, polarized dust with multiple temperatures and a decorrelation of the signals with frequency, which introduce complexity that is consistent with current data. We also present a new prescription for adding small-scale realizations of these components at resolutions greater than current all-sky measurements. The usefulness of the code is demonstrated by forecasting the impact of varying foreground complexity on the recovered tensor-to-scalar ratio for the LiteBIRD satellite. The code is available at: https://github.com/bthorne93/PySM_public.

Pulsed thermography detection of water and hydraulic oil intrusion in the honeycomb sandwich structure composite

NASA Astrophysics Data System (ADS)

Zhao, Shi-bin; Zhang, Cun-lin; Wu, Nai-ming

2011-08-01

Water and hydraulic oil intrusion inside honeycomb sandwich Structure Composite during service has been linked to in-flight failure in some aircraft. There is an ongoing effort to develop nondestructive testing methods to detect the presence of water and hydraulic oil within the sandwich panels. Pulsed thermography(PT) represents an attractive approach in that it is sensitive to the change of thermal properties. Using a flash lamp PT, testing can be applied directly to the surface of the panel. The viability of PT is demonstrated through laboratory imaging of both water and hydraulic oil within sandwich panels. The detection of water and hydraulic oil intrusion using a one-sided flash lamp PT is presented. It is shown that simple detection, as well as spatial localization of water and hydraulic oil within sandwich panels, and assign the quantity of water and hydraulic oil is possible.

Thermal Stratification Analysis for Sodium Fast Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schneider, James; Anderson, Mark; Baglietto, Emilio

The sodium fast reactor (SFR) is the most mature reactor concept of all the generation-IV nuclear systems and is a promising reactor design that is currently under development by several organizations. The majority of sodium fast reactor designs utilize a pool type arrangement which incorporates the primary coolant pumps and intermediate heat exchangers within the sodium pool. These components typically protrude into the pool thus reducing the risk and severity of a loss of coolant accidents. To further ensure safe operation under even the most severe transients a more comprehensive understanding of key thermal hydraulic phenomena in this pool ismore » desired. One of the key technology gaps identified for SFR safety is determining the extent and the effects of thermal stratification developing in the pool during postulated accident scenarios such as a protected or unprotected loss of flow incident. In an effort to address these issues, detailed flow models of transient stratification in the pool during an accident can be developed. However, to develop the calculation models, and ensure they can reproduce the underlying physics, highly spatially resolved data is needed. This data can be used in conjunction with advanced computational fluid dynamic calculations to aid in the development of simple reduced dimensional models for systems codes such as SAM and SAS4A/SASSYS-1.« less

RERTR-10 Irradiation Summary Report

DOE Office of Scientific and Technical Information (OSTI.GOV)

D. M. Perez

2011-05-01

The Reduced Enrichment for Research and Test Reactor (RERTR) experiment RERTR-10 was designed to further test the effectiveness of modified fuel/clad interfaces in monolithic fuel plates. The experiment was conducted in two campaigns: RERTR-10A and RERTR-10B. The fuel plates tested in RERTR-10A were all fabricated by Hot Isostatic Pressing (HIP) and were designed to evaluate the effect of various Si levels in the interlayer and the thickness of the Zr interlayer (0.001”) using 0.010” and 0.020” nominal foil thicknesses. The fuel plates in RERTR-10B were fabricated by Friction Bonding (FB) with two different thickness Si layers and Nb and Zrmore » diffusion barriers.1 The following report summarizes the life of the RERTR-10A/B experiment through end of irradiation, including as-run neutronic analysis results, thermal analysis results and hydraulic testing results.« less

CFD Aided Design and Production of Hydraulic Turbines

NASA Astrophysics Data System (ADS)

Kaplan, Alper; Cetinturk, Huseyin; Demirel, Gizem; Ayli, Ece; Celebioglu, Kutay; Aradag, Selin; ETU Hydro Research Center Team

2014-11-01

Hydraulic turbines are turbo machines which produce electricity from hydraulic energy. Francis type turbines are the most common one in use today. The design of these turbines requires high engineering effort since each turbine is tailor made due to different head and discharge. Therefore each component of the turbine is designed specifically. During the last decades, Computational Fluid Dynamics (CFD) has become very useful tool to predict hydraulic machinery performance and save time and money for designers. This paper describes a design methodology to optimize a Francis turbine by integrating theoretical and experimental fundamentals of hydraulic machines and commercial CFD codes. Specific turbines are designed and manufactured with the help of a collaborative CFD/CAD/CAM methodology based on computational fluid dynamics and five-axis machining for hydraulic electric power plants. The details are presented in this study. This study is financially supported by Turkish Ministry of Development.

Development of Multi-physics (Multiphase CFD + MCNP) simulation for generic solution vessel power calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Seung Jun; Buechler, Cynthia Eileen

The current study aims to predict the steady state power of a generic solution vessel and to develop a corresponding heat transfer coefficient correlation for a Moly99 production facility by conducting a fully coupled multi-physics simulation. A prediction of steady state power for the current application is inherently interconnected between thermal hydraulic characteristics (i.e. Multiphase computational fluid dynamics solved by ANSYS-Fluent 17.2) and the corresponding neutronic behavior (i.e. particle transport solved by MCNP6.2) in the solution vessel. Thus, the development of a coupling methodology is vital to understand the system behavior at a variety of system design and postulated operatingmore » scenarios. In this study, we report on the k-effective (keff) calculation for the baseline solution vessel configuration with a selected solution concentration using MCNP K-code modeling. The associated correlation of thermal properties (e.g. density, viscosity, thermal conductivity, specific heat) at the selected solution concentration are developed based on existing experimental measurements in the open literature. The numerical coupling methodology between multiphase CFD and MCNP is successfully demonstrated, and the detailed coupling procedure is documented. In addition, improved coupling methods capturing realistic physics in the solution vessel thermal-neutronic dynamics are proposed and tested further (i.e. dynamic height adjustment, mull-cell approach). As a key outcome of the current study, a multi-physics coupling methodology between MCFD and MCNP is demonstrated and tested for four different operating conditions. Those different operating conditions are determined based on the neutron source strength at a fixed geometry condition. The steady state powers for the generic solution vessel at various operating conditions are reported, and a generalized correlation of the heat transfer coefficient for the current application is discussed. The assessment of multi-physics methodology and preliminary results from various coupled calculations (power prediction and heat transfer coefficient) can be further utilized for the system code validation and generic solution vessel design improvement.« less

Analysis of discontinuities across thin inhomogeneities, groundwater/surface water interactions in river networks, and circulation about slender bodies using slit elements in the Analytic Element Method

USDA-ARS?s Scientific Manuscript database

Groundwater and surface water contain interfaces across which hydrologic functions are discontinuous. Thin elements with high hydraulic conductivity in a porous media focus groundwater, which flows through such inhomogeneities and causes an abrupt change in stream function across their interfaces, a...

Assimilation of temperature and hydraulic gradients for quantifying the spatial variability of streambed hydraulics

NASA Astrophysics Data System (ADS)

Huang, Xiang; Andrews, Charles B.; Liu, Jie; Yao, Yingying; Liu, Chuankun; Tyler, Scott W.; Selker, John S.; Zheng, Chunmiao

2016-08-01

Understanding the spatial and temporal characteristics of water flux into or out of shallow aquifers is imperative for water resources management and eco-environmental conservation. In this study, the spatial variability in the vertical specific fluxes and hydraulic conductivities in a streambed were evaluated by integrating distributed temperature sensing (DTS) data and vertical hydraulic gradients into an ensemble Kalman filter (EnKF) and smoother (EnKS) and an empirical thermal-mixing model. The formulation of the EnKF/EnKS assimilation scheme is based on a discretized 1D advection-conduction equation of heat transfer in the streambed. We first systematically tested a synthetic case and performed quantitative and statistical analyses to evaluate the performance of the assimilation schemes. Then a real-world case was evaluated to calculate assimilated specific flux. An initial estimate of the spatial distributions of the vertical hydraulic gradients was obtained from an empirical thermal-mixing model under steady-state conditions using a constant vertical hydraulic conductivity. Then, this initial estimate was updated by repeatedly dividing the assimilated specific flux by estimates of the vertical hydraulic gradients to obtain a refined spatial distribution of vertical hydraulic gradients and vertical hydraulic conductivities. Our results indicate that optimal parameters can be derived with fewer iterations but greater simulation effort using the EnKS compared with the EnKF. For the field application in a stream segment of the Heihe River Basin in northwest China, the average vertical hydraulic conductivities in the streambed varied over three orders of magnitude (5 × 10-1 to 5 × 102 m/d). The specific fluxes ranged from near zero (qz < ±0.05 m/d) to ±1.0 m/d, while the vertical hydraulic gradients were within the range of -0.2 to 0.15 m/m. The highest and most variable fluxes occurred adjacent to a debris-dam and bridge pier. This phenomenon is very likely the result of heterogeneous streambed hydraulic characteristics in these areas. Our results have significant implications for hyporheic micro-habitats, fish spawning and other wildlife incubation, regional flow and hyporheic solute transport models in the Heihe River Basin, as well as in other similar hydrologic settings.

Benchmark of Atucha-2 PHWR RELAP5-3D control rod model by Monte Carlo MCNP5 core calculation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pecchia, M.; D'Auria, F.; Mazzantini, O.

2012-07-01

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and peculiarities require the adoption of advanced Monte Carlo codes for performing realistic neutronic simulations. Therefore core models of Atucha-2 PHWR were developed using MCNP5. In this work a methodology was set up to collect the flux in the hexagonal mesh by which the Atucha-2 core is represented. The scope of this activity is to evaluate the effect of obliquely inserted control rod on neutron flux in order to validate the RELAP5-3D{sup C}/NESTLE three dimensional neutron kinetic coupled thermal-hydraulic model, applied by GRNSPG/UNIPI formore » performing selected transients of Chapter 15 FSAR of Atucha-2. (authors)« less

Contribution to the thermodynamic description of the corium - The U-Zr-O system

NASA Astrophysics Data System (ADS)

Quaini, A.; Guéneau, C.; Gossé, S.; Dupin, N.; Sundman, B.; Brackx, E.; Domenger, R.; Kurata, M.; Hodaj, F.

2018-04-01

In order to understand the stratification process that may occur in the late phase of the fuel degradation during a severe accident in a PWR, the thermodynamic knowledge of the U-Zr-O system is crucial. The presence of a miscibility gap in the U-Zr-O liquid phase may lead to a stratified configuration, which will impact the accidental scenario management. The aim of this work was to obtain new experimental data in the U-Zr-O liquid miscibility gap. New tie-line data were provided at 2567 ± 25 K. The related thermodynamic models were reassessed using present data and literature values. The reassessed model will be implemented in the TAF-ID international database. The composition and density of phases potentially formed during stratification will be predicted by coupling current thermodynamic model with thermal-hydraulics codes.

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Clarno, Kevin T; Palmtag, Scott; Davidson, Gregory G

2014-01-01

The Consortium for Advanced Simulation of Light Water Reactors (CASL) is developing a core simulator called VERA-CS to model operating PWR reactors with high resolution. This paper describes how the development of VERA-CS is being driven by a set of progression benchmark problems that specify the delivery of useful capability in discrete steps. As part of this development, this paper will describe the current capability of VERA-CS to perform a multiphysics simulation of an operating PWR at Hot Full Power (HFP) conditions using a set of existing computer codes coupled together in a novel method. Results for several single-assembly casesmore » are shown that demonstrate coupling for different boron concentrations and power levels. Finally, high-resolution results are shown for a full-core PWR reactor modeled in quarter-symmetry.« less

Analysis of the SL-1 Accident Using RELAPS5-3D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Francisco, A.D. and Tomlinson, E. T.

2007-11-08

On January 3, 1961, at the National Reactor Testing Station, in Idaho Falls, Idaho, the Stationary Low Power Reactor No. 1 (SL-1) experienced a major nuclear excursion, killing three people, and destroying the reactor core. The SL-1 reactor, a 3 MW{sub t} boiling water reactor, was shut down and undergoing routine maintenance work at the time. This paper presents an analysis of the SL-1 reactor excursion using the RELAP5-3D thermal-hydraulic and nuclear analysis code, with the intent of simulating the accident from the point of reactivity insertion to destruction and vaporization of the fuel. Results are presented, along with amore » discussion of sensitivity to some reactor and transient parameters (many of the details are only known with a high level of uncertainty).« less

TEMPEST. Transient 3-D Thermal-Hydraulic

DOE Office of Scientific and Technical Information (OSTI.GOV)

Eyler, L.L.

TEMPEST is a transient, three-dimensional, hydrothermal program that is designed to analyze a range of coupled fluid dynamic and heat transfer systems of particular interest to the Fast Breeder Reactor (FBR) thermal-hydraulic design community. The full three-dimensional, time-dependent equations of motion, continuity, and heat transport are solved for either laminar or turbulent fluid flow, including heat diffusion and generation in both solid and liquid materials. The equations governing mass, momentum, and energy conservation for incompressible flows and small density variations (Boussinesq approximation) are solved using finite-difference techniques. Analyses may be conducted in either cylindrical or Cartesian coordinate systems. Turbulence ismore » treated using a two-equation model. Two auxiliary plotting programs, SEQUEL and MANPLOT, for use with TEMPEST output are included. SEQUEL may be operated in batch or interactive mode; it generates data required for vector plots, contour plots of scalar quantities, line plots, grid and boundary plots, and time-history plots. MANPLOT reads the SEQUEL-generated data and creates the hardcopy plots. TEMPEST can be a valuable hydrothermal design analysis tool in areas outside the intended FBR thermal-hydraulic design community.« less

Influence of engineered interfaces on residual stresses and mechanical response in metal matrix composites

NASA Technical Reports Server (NTRS)

Arnold, Steven M.; Wilt, Thomas E.

1992-01-01

Because of the inherent coefficient of thermal expansion (CTE) mismatch between fiber and matrix within metal and intermetallic matrix composite systems, high residual stresses can develop under various thermal loading conditions. These conditions include cooling from processing temperature to room temperature as well as subsequent thermal cycling. As a result of these stresses, within certain composite systems, radial, circumferential, and/or longitudinal cracks have been observed to form at the fiber matrix interface region. A number of potential solutions for reducing this thermally induced residual stress field have been proposed recently. Examples of some potential solutions are high CTE fibers, fiber preheating, thermal anneal treatments, and an engineered interface. Here the focus is on designing an interface (by using a compensating/compliant layer concept) to reduce or eliminate the thermal residual stress field and, therefore, the initiation and propagation of cracks developed during thermal loading. Furthermore, the impact of the engineered interface on the composite's mechanical response when subjected to isothermal mechanical load histories is examined.

CBP PHASE I CODE INTEGRATION

DOE Office of Scientific and Technical Information (OSTI.GOV)

Smith, F.; Brown, K.; Flach, G.

The goal of the Cementitious Barriers Partnership (CBP) is to develop a reasonable and credible set of software tools to predict the structural, hydraulic, and chemical performance of cement barriers used in nuclear applications over extended time frames (greater than 100 years for operating facilities and greater than 1000 years for waste management). The simulation tools will be used to evaluate and predict the behavior of cementitious barriers used in near surface engineered waste disposal systems including waste forms, containment structures, entombments, and environmental remediation. These cementitious materials are exposed to dynamic environmental conditions that cause changes in material propertiesmore » via (i) aging, (ii) chloride attack, (iii) sulfate attack, (iv) carbonation, (v) oxidation, and (vi) primary constituent leaching. A set of state-of-the-art software tools has been selected as a starting point to capture these important aging and degradation phenomena. Integration of existing software developed by the CBP partner organizations was determined to be the quickest method of meeting the CBP goal of providing a computational tool that improves the prediction of the long-term behavior of cementitious materials. These partner codes were selected based on their maturity and ability to address the problems outlined above. The GoldSim Monte Carlo simulation program (GTG 2010a, GTG 2010b) was chosen as the code integration platform (Brown & Flach 2009b). GoldSim (current Version 10.5) is a Windows based graphical object-oriented computer program that provides a flexible environment for model development (Brown & Flach 2009b). The linking of GoldSim to external codes has previously been successfully demonstrated (Eary 2007, Mattie et al. 2007). GoldSim is capable of performing deterministic and probabilistic simulations and of modeling radioactive decay and constituent transport. As part of the CBP project, a general Dynamic Link Library (DLL) interface was developed to link GoldSim with external codes (Smith III et al. 2010). The DLL uses a list of code inputs provided by GoldSim to create an input file for the external application, runs the external code, and returns a list of outputs (read from files created by the external application) back to GoldSim. In this way GoldSim provides: (1) a unified user interface to the applications, (2) the capability of coupling selected codes in a synergistic manner, and (3) the capability of performing probabilistic uncertainty analysis with the codes. GoldSim is made available by the GoldSim Technology Group as a free 'Player' version that allows running but not editing GoldSim models. The player version makes the software readily available to a wider community of users that would wish to use the CBP application but do not have a license for GoldSim.« less

Demonstration of optimum fuel-to-moderator ratio in a PWR unit fuel cell

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feltus, M.A.; Pozsgai, C.

1992-01-01

Nuclear engineering students at The Pennsylvania State University develop scaled-down [[approx]350 MW(thermal)] pressurized water reactors (PWRs) using actual plants as references. The design criteria include maintaining the clad temperature below 2200[degree]F, fuel temperature below melting point, sufficient departure from nucleate boiling ratio (DNBR) margin, a beginning-of-life boron concentration that yields a negative moderator temperature coefficient, an adequate cycle power production (330 effective full-power days), and a batch loading scheme that is economical. The design project allows for many degrees of freedom (e.g., assembly number, pitch and height and batch enrichments) so that each student's result is unique. The iterative naturemore » of the design process is stressed in the course. The LEOPARD code is used for the unit cell depletion, critical boron, and equilibrium xenon calculations. Radial two-group diffusion equations are solved with the TWIDDLE-DEE code. The steady-state ZEBRA thermal-hydraulics program is used for calculating DNBR. The unit fuel cell pin radius and pitch (fuel-to-moerator ratio) for the scaled-down design, however, was set equal to the already optimized ratio for the reference PWR. This paper describes an honors project that shows how the optimum fuel-to-moderator ratio is found for a unit fuel cell shown in terms of neutron economics. This exercise illustrates the impact of fuel-to-moderator variations on fuel utilization factor and the effect of assuming space and energy separability.« less

Critical heat flux in subcooled flow boiling

NASA Astrophysics Data System (ADS)

Hall, David Douglas

The critical heat flux (CHF) phenomenon was investigated for water flow in tubes with particular emphasis on the development of methods for predicting CHF in the subcooled flow boiling regime. The Purdue University Boiling and Two-Phase Flow Laboratory (PU-BTPFL) CHF database for water flow in a uniformly heated tube was compiled from the world literature dating back to 1949 and represents the largest CHF database ever assembled with 32,544 data points from over 100 sources. The superiority of this database was proven via a detailed examination of previous databases. The PU-BTPFL CHF database is an invaluable tool for the development of CHF correlations and mechanistic models that are superior to existing ones developed with smaller, less comprehensive CHF databases. In response to the many inaccurate and inordinately complex correlations, two nondimensional, subcooled CHF correlations were formulated, containing only five adjustable constants and whose unique functional forms were determined without using a statistical analysis but rather using the parametric trends observed in less than 10% of the subcooled CHF data. The correlation based on inlet conditions (diameter, heated length, mass velocity, pressure, inlet quality) was by far the most accurate of all known subcooled CHF correlations, having mean absolute and root-mean-square (RMS) errors of 10.3% and 14.3%, respectively. The outlet (local) conditions correlation was the most accurate correlation based on local CHF conditions (diameter, mass velocity, pressure, outlet quality) and may be used with a nonuniform axial heat flux. Both correlations proved more accurate than a recent CHF look-up table commonly employed in nuclear reactor thermal hydraulic computer codes. An interfacial lift-off, subcooled CHF model was developed from a consideration of the instability of the vapor-liquid interface and the fraction of heat required for liquid-vapor conversion as opposed to that for bulk liquid heating. Severe vapor effusion in an upstream wetting front lifts the vapor-liquid interface off the surface, triggering CHF. Since the model is entirely based on physical observations, it has the potential to accurately predict CHF for other fluids and flow geometries which are beyond the conditions for which it was validated.

An Accessible User Interface for Geoscience and Programming

NASA Astrophysics Data System (ADS)

Sevre, E. O.; Lee, S.

2012-12-01

The goal of this research is to develop an interface that will simplify user interaction with software for scientists. The motivating factor of the research is to develop tools that assist scientists with limited motor skills with the efficient generation and use of software tools. Reliance on computers and programming is increasing in the world of geology, and it is increasingly important for geologists and geophysicists to have the computational resources to use advanced software and edit programs for their research. I have developed a prototype of a program to help geophysicists write programs using a simple interface that requires only simple single-mouse-clicks to input code. It is my goal to minimize the amount of typing necessary to create simple programs and scripts to increase accessibility for people with disabilities limiting fine motor skills. This interface can be adapted for various programming and scripting languages. Using this interface will simplify development of code for C/C++, Java, and GMT, and can be expanded to support any other text based programming language. The interface is designed around the concept of maximizing the amount of code that can be written using a minimum number of clicks and typing. The screen is split into two sections: a list of click-commands is on the left hand side, and a text area is on the right hand side. When the user clicks on a command on the left hand side the applicable code is automatically inserted at the insertion point in the text area. Currently in the C/C++ interface, there are commands for common code segments that are often used, such as for loops, comments, print statements, and structured code creation. The primary goal is to provide an interface that will work across many devices for developing code. A simple prototype has been developed for the iPad. Due to the limited number of devices that an iOS application can be used with, the code has been re-written in Java to run on a wider range of devices. Currently, the software works in a prototype mode, and it is our goal to further development to create software that can benefit a wide range of people working in geosciences, which will make code development practical and accessible for a wider audience of scientists. By using an interface like this, it reduces potential for errors by reusing known working code.

Comparison of the PLTEMP code flow instability predictions with measurements made with electrically heated channels for the advanced test reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feldman, E.

When the University of Missouri Research Reactor (MURR) was designed in the 1960s the potential for fuel element burnout by a phenomenon referred to at that time as 'autocatalytic vapor binding' was of serious concern. This type of burnout was observed to occur at power levels considerably lower than those that were known to cause critical heat flux. The conversion of the MURR from HEU fuel to LEU fuel will probably require significant design changes, such as changes in coolant channel thicknesses, that could affect the thermal-hydraulic behavior of the reactor core. Therefore, the redesign of the MURR to accommodatemore » an LEU core must address the same issues of fuel element burnout that were of concern in the 1960s. The Advanced Test Reactor (ATR) was designed at about the same time as the MURR and had similar concerns with regard to fuel element burnout. These concerns were addressed in the ATR by two groups of thermal-hydraulic tests that employed electrically heated simulated fuel channels. The Croft (1964), Reference 1, tests were performed at ANL. The Waters (1966), Reference 2, tests were performed at Hanford Laboratories in Richland Washington. Since fuel element surface temperatures rise rapidly as burnout conditions are approached, channel surface temperatures were carefully monitored in these experiments. For self-protection, the experimental facilities were designed to cut off the electric power when rapidly increasing surface temperatures were detected. In both the ATR reactor and in the tests with electrically heated channels, the heated length of the fuel plate was 48 inches, which is about twice that of the MURR. Whittle and Forgan (1967) independently conducted tests with electrically heated rectangular channels that were similar to the tests by Croft and by Walters. In the Whittle and Forgan tests the heated length of the channel varied among the tests and was between 16 and 24 inches. Both Waters and Whittle and Forgan show that the cause of the fuel element burnout is due to a form of flow instability. Whittle and Forgan provide a formula that predicts when this flow instability will occur. This formula is included in the PLTEMP/ANL code.Error! Reference source not found. Olson has shown that the PLTEMP/ANL code accurately predicts the powers at which flow instability occurs in the Whittle and Forgan experiments. He also considered the electrically heated tests performed in the ANS Thermal-Hydraulic Test Loop at ORNL and report by M. Siman-Tov et al. The purpose of this memorandum is to demonstrate that the PLTEMP/ANL code accurately predicts the Croft and the Waters tests. This demonstration should provide sufficient confidence that the PLTEMP/ANL code can adequately predict the onset of flow instability for the converted MURR. The MURR core uses light water as a coolant, has a 24-inch active fuel length, downward flow in the core, and an average core velocity of about 7 m/s. The inlet temperature is about 50 C and the peak outlet is about 20 C higher than the inlet for reactor operation at 10 MW. The core pressures range from about 4 to about 5 bar. The peak heat flux is about 110 W/cm{sup 2}. Section 2 describes the mechanism that causes flow instability. Section 3 describes the Whittle and Forgan formula for flow instability. Section 4 briefly describes both the Croft and the Waters experiments. Section 5 describes the PLTEMP/ANL models. Section 6 compares the PLTEMP/ANL predictions based on the Whittle and Forgan formula with the Croft measurements. Section 7 does the same for the Waters measurements. Section 8 provides the range of parameters for the Whittle and Forgan tests. Section 9 discusses the results and provides conclusions. In conclusion, although there is no single test that by itself closely matches the limiting conditions in the MURR, the preponderance of measured data and the ability of the Whittle and Forgan correlation, as implemented in PLTEMP/ANL, to predict the onset of flow instability for these tests leads one to the conclusion that the same method should be able to predict the onset of flow instability in the MURR reasonably well.« less

Improved understanding of the relationship between hydraulic properties and streaming potentials

NASA Astrophysics Data System (ADS)

Cassiani, G.; Brovelli, A.

2009-12-01

Streaming potential (SP) measurements have been satisfactorily used in a number of recent studies as a non-invasive tool to monitor fluid movement in both the vadose and the saturated zone. SPs are generated from the coupling between two independent physical processes oc-curring at the pore-level, namely water flow and excess of ions at the negatively charged solid matrix-water interface. The intensity of the measured potentials depends on physical proper-ties of the medium, including the internal micro-geometry of the system, the charge density of the interface and the composition of the pore fluid, which affects its ionic strength, pH and redox potential. The goal of this work is to investigate whether a relationship between the intensity of the SPs and the saturated hydraulic conductivity can be identified. Both properties are - at least to some extent - dependent on the pore-size distribution and connectivity of the pores, and there-fore some degree of correlation is expected. We used a pore-scale numerical model previously developed to simulate both the bulk hydraulic conductivity and the intensity of the SPs gener-ated in a three-dimensional pore-network. The chemical-physical properties of both the inter-face (Zeta-potential) and of the aqueous phase are computed using an analytical, physically based model that has shown good agreement with experimental data. Modelling results were satisfactorily compared with experimental data, showing that the model, although simplified retains the key properties and mechanisms that control SP generation. A sensitivity analysis with respect to the key geometrical and chemical parameters was conducted to evaluate how the correlation between the two studied variables changes and to ascertain whether the bulk hydraulic conductivity can be estimated from SP measurements alone.

4D synchrotron X-ray imaging to understand porosity development in shales during exposure to hydraulic fracturing fluid

NASA Astrophysics Data System (ADS)

Kiss, A. M.; Bargar, J.; Kohli, A. H.; Harrison, A. L.; Jew, A. D.; Lim, J. H.; Liu, Y.; Maher, K.; Zoback, M. D.; Brown, G. E.

2016-12-01

Unconventional (shale) reservoirs have emerged as the most important source of petroleum resources in the United States and represent a two-fold decrease in greenhouse gas emissions compared to coal. Despite recent progress, hydraulic fracturing operations present substantial technical, economic, and environmental challenges, including inefficient recovery, wastewater production and disposal, contaminant and greenhouse gas pollution, and induced seismicity. A relatively unexplored facet of hydraulic fracturing operations is the fluid-rock interface, where hydraulic fracturing fluid (HFF) contacts shale along faults and fractures. Widely used, water-based fracturing fluids contain oxidants and acid, which react strongly with shale minerals. Consequently, fluid injection and soaking induces a host of fluid-rock interactions, most notably the dissolution of carbonates and sulfides, producing enhanced or "secondary" porosity networks, as well as mineral precipitation. The competition between these mechanisms determines how HFF affects reactive surface area and permeability of the shale matrix. The resultant microstructural and chemical changes may also create capillary barriers that can trap hydrocarbons and water. A mechanistic understanding of the microstructure and chemistry of the shale-HFF interface is needed to design new methodologies and fracturing fluids. Shales were imaged using synchrotron micro-X-ray computed tomography before, during, and after exposure to HFF to characterize changes to the initial 3D structure. CT reconstructions reveal how the secondary porosity networks advance into the shale matrix. Shale samples span a range of lithologies from siliceous to calcareous to organic-rich. By testing shales of different lithologies, we have obtained insights into the mineralogic controls on secondary pore network development and the morphologies at the shale-HFF interface and the ultimate composition of produced water from different facies. These results show that mineral texture is a major control over secondary porosity network morphology.

INTRACORPOREAL HEAT DISSIPATION FROM A RADIOISOTOPE-POWERED ARTIFICIAL HEART.

PubMed

Huffman, Fred N.; Hagen, Kenneth G.; Whalen, Robert L.; Fuqua, John M.; Norman, John C.

1974-01-01

The feasibility of radioisotope-fueled circulatory support systems depends on the ability of the body to dissipate the reject heat from the power source driving the blood pump as well as to tolerate chronic intracorporeal radiation. Our studies have focused on the use of the circulating blood as a heat sink. Initial in vivo heat transfer studies utilized straight tube heat exchangers (electrically and radioisotope energized) to replace a segment of the descending aorta. More recent studies have used a left ventricular assist pump as a blood-cooled heat exchanger. This approach minimizes trauma, does not increase the area of prosthetic interface with the blood, and minimizes system volume. Heat rejected from the thermal engine (vapor or gas cycle) is transported from the nuclear power source in the abdomen to the pump in the thoracic cavity via hydraulic lines. Adjacent tissue is protected from the fuel capsule temperature (900 to 1200 degrees F) by vacuum foil insulation and polyurethane foam. The in vivo thermal management problems have been studied using a simulated thermal system (STS) which approximates the heat rejection and thermal transport mechanisms of the nuclear circulatory support systems under development by NHLI. Electric heaters simulate the reject heat from the thermal engines. These studies have been essential in establishing the location, suspension, surgical procedures, and postoperative care for implanting prototype nuclear heart assist systems in calves. The pump has a thermal impedance of 0.12 degrees C/watt. Analysis of the STS data in terms of an electrical analog model implies a heat transfer coefficient of 4.7 x 10(-3) watt/cm(2) degrees C in the abdomen compared to a value of 14.9 x 10(-3) watt/cm(2) degrees C from the heat exchanger plenum into the diaphragm.

Investigation of Thermal Interface Materials Using Phase-Sensitive Transient Thermoreflectance Technique: Preprint

DOE Office of Scientific and Technical Information (OSTI.GOV)

Feng, X.; King, C.; DeVoto, D.

2014-08-01

With increasing power density in electronics packages/modules, thermal resistances at multiple interfaces are a bottleneck to efficient heat removal from the package. In this work, the performance of thermal interface materials such as grease, thermoplastic adhesives and diffusion-bonded interfaces are characterized using the phase-sensitive transient thermoreflectance technique. A multi-layer heat conduction model was constructed and theoretical solutions were derived to obtain the relation between phase lag and the thermal/physical properties. This technique enables simultaneous extraction of the contact resistance and bulk thermal conductivity of the TIMs. With the measurements, the bulk thermal conductivity of Dow TC-5022 thermal grease (70 tomore » 75 um bondline thickness) was 3 to 5 W/(m-K) and the contact resistance was 5 to 10 mm2-K/W. For the Btech thermoplastic material (45 to 80 μm bondline thickness), the bulk thermal conductivity was 20 to 50 W/(m-K) and the contact resistance was 2 to 5 mm2-K/W. Measurements were also conducted to quantify the thermal performance of diffusion-bonded interface for power electronics applications. Results with the diffusion-bonded sample showed that the interfacial thermal resistance is more than one order of magnitude lower than those of traditional TIMs, suggesting potential pathways to efficient thermal management.« less

Tidal Boundary Conditions in SEAWAT

USGS Publications Warehouse

Mulligan, Ann E.; Langevin, Christian; Post, Vincent E.A.

2011-01-01

SEAWAT, a U.S. Geological Survey groundwater flow and transport code, is increasingly used to model the effects of tidal motion on coastal aquifers. Different options are available to simulate tidal boundaries but no guidelines exist nor have comparisons been made to identify the most effective approach. We test seven methods to simulate a sloping beach and a tidal flat. The ocean is represented in one of the three ways: directly using a high hydraulic conductivity (high-K) zone and indirect simulation via specified head boundaries using either the General Head Boundary (GHB) or the new Periodic Boundary Condition (PBC) package. All beach models simulate similar water fluxes across the upland boundary and across the sediment-water interface although the ratio of intertidal to subtidal flow is different at low tide. Simulating a seepage face results in larger intertidal fluxes and influences near-shore heads and salinity. Major differences in flow occur in the tidal flat simulations. Because SEAWAT does not simulate unsaturated flow the water table only rises via flow through the saturated zone. This results in delayed propagation of the rising tidal signal inland. Inundation of the tidal flat is delayed as is flow into the aquifer across the flat. This is severe in the high-K and PBC models but mild in the GHB models. Results indicate that any of the tidal boundary options are fine if the ocean-aquifer interface is steep. However, as the slope of that interface decreases, the high-K and PBC approaches perform poorly and the GHB boundary is preferable.

Elimination of numerical diffusion in 1 - phase and 2 - phase flows

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rajamaeki, M.

1997-07-01

The new hydraulics solution method PLIM (Piecewise Linear Interpolation Method) is capable of avoiding the excessive errors, numerical diffusion and also numerical dispersion. The hydraulics solver CFDPLIM uses PLIM and solves the time-dependent one-dimensional flow equations in network geometry. An example is given for 1-phase flow in the case when thermal-hydraulics and reactor kinetics are strongly coupled. Another example concerns oscillations in 2-phase flow. Both the example computations are not possible with conventional methods.

A Cooling System for the EAPU Shuttle Upgrade

NASA Technical Reports Server (NTRS)

Tongue, Stephen; Guyette, Greg; Irbeck, Bradley

2001-01-01

The Shuttle orbiter currently uses hydrazine-powered APU's for powering its hydraulic system pumps. To enhance vehicle safety and reliability, NASA is pursuing an APU upgrade where the hydrazine powered turbine is replaced by an electric motor pump and battery power supply. This EAPU (Electric APU) upgrade presents several thermal control challenges most notably the new requirement for moderate temperature control of high-power electron ics at 132 of (55.6 C). This paper describes how the existing Water Spray Boiler (WSB), which currently cools the hydraulic fluid and APU lubrication oil, is being modified to provide EAPU thermal management.

Application of the TEMPEST computer code to canister-filling heat transfer problems

DOE Office of Scientific and Technical Information (OSTI.GOV)

Farnsworth, R.K.; Faletti, D.W.; Budden, M.J.

Pacific Northwest Laboratory (PNL) researchers used the TEMPEST computer code to simulate thermal cooldown behavior of nuclear waste glass after it was poured into steel canisters for long-term storage. The objective of this work was to determine the accuracy and applicability of the TEMPEST code when used to compute canister thermal histories. First, experimental data were obtained to provide the basis for comparing TEMPEST-generated predictions. Five canisters were instrumented with appropriately located radial and axial thermocouples. The canister were filled using the pilot-scale ceramic melter (PSCM) at PNL. Each canister was filled in either a continous or a batch fillingmore » mode. One of the canisters was also filled within a turntable simulant (a group of cylindrical shells with heat transfer resistances similar to those in an actual melter turntable). This was necessary to provide a basis for assessing the ability of the TEMPEST code to also model the transient cooling of canisters in a melter turntable. The continous-fill model, Version M, was found to predict temperatures with more accuracy. The turntable simulant experiment demonstrated that TEMPEST can adequately model the asymmetric temperature field caused by the turntable geometry. Further, TEMPEST can acceptably predict the canister cooling history within a turntable, despite code limitations in computing simultaneous radiation and convection heat transfer between shells, along with uncertainty in stainless-steel surface emissivities. Based on the successful performance of TEMPEST Version M, development was initiated to incorporate 1) full viscous glass convection, 2) a dynamically adaptive grid that automatically follows the glass/air interface throughout the transient, and 3) a full enclosure radiation model to allow radiation heat transfer to non-nearest neighbor cells. 5 refs., 47 figs., 17 tabs.« less

Standard interface files and procedures for reactor physics codes, version III

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carmichael, B.M.

Standards and procedures for promoting the exchange of reactor physics codes are updated to Version-III status. Standards covering program structure, interface files, file handling subroutines, and card input format are included. The implementation status of the standards in codes and the extension of the standards to new code areas are summarized. (15 references) (auth)

Thermal transport across metal–insulator interface via electron–phonon interaction.

PubMed

Zhang, Lifa; Lü, Jing-Tao; Wang, Jian-Sheng; Li, Baowen

2013-11-06

The thermal transport across a metal–insulator interface can be characterized by electron–phonon interaction through which an electron lead is coupled to a phonon lead if phonon–phonon coupling at the interface is very weak. We investigate the thermal conductance and rectification between the electron part and the phonon part using the nonequilibrium Green's function method. It is found that the thermal conductance has a nonmonotonic behavior as a function of average temperature or the coupling strength between the phonon leads in the metal part and the insulator part. The metal–insulator interface shows a clear thermal rectification effect, which can be reversed by a change in average temperature or the electron–phonon coupling.

IAC - INTEGRATED ANALYSIS CAPABILITY

NASA Technical Reports Server (NTRS)

Frisch, H. P.

1994-01-01

The objective of the Integrated Analysis Capability (IAC) system is to provide a highly effective, interactive analysis tool for the integrated design of large structures. With the goal of supporting the unique needs of engineering analysis groups concerned with interdisciplinary problems, IAC was developed to interface programs from the fields of structures, thermodynamics, controls, and system dynamics with an executive system and database to yield a highly efficient multi-disciplinary system. Special attention is given to user requirements such as data handling and on-line assistance with operational features, and the ability to add new modules of the user's choice at a future date. IAC contains an executive system, a data base, general utilities, interfaces to various engineering programs, and a framework for building interfaces to other programs. IAC has shown itself to be effective in automatic data transfer among analysis programs. IAC 2.5, designed to be compatible as far as possible with Level 1.5, contains a major upgrade in executive and database management system capabilities, and includes interfaces to enable thermal, structures, optics, and control interaction dynamics analysis. The IAC system architecture is modular in design. 1) The executive module contains an input command processor, an extensive data management system, and driver code to execute the application modules. 2) Technical modules provide standalone computational capability as well as support for various solution paths or coupled analyses. 3) Graphics and model generation interfaces are supplied for building and viewing models. Advanced graphics capabilities are provided within particular analysis modules such as INCA and NASTRAN. 4) Interface modules provide for the required data flow between IAC and other modules. 5) User modules can be arbitrary executable programs or JCL procedures with no pre-defined relationship to IAC. 6) Special purpose modules are included, such as MIMIC (Model Integration via Mesh Interpolation Coefficients), which transforms field values from one model to another; LINK, which simplifies incorporation of user specific modules into IAC modules; and DATAPAC, the National Bureau of Standards statistical analysis package. The IAC database contains structured files which provide a common basis for communication between modules and the executive system, and can contain unstructured files such as NASTRAN checkpoint files, DISCOS plot files, object code, etc. The user can define groups of data and relations between them. A full data manipulation and query system operates with the database. The current interface modules comprise five groups: 1) Structural analysis - IAC contains a NASTRAN interface for standalone analysis or certain structural/control/thermal combinations. IAC provides enhanced structural capabilities for normal modes and static deformation analysis via special DMAP sequences. IAC 2.5 contains several specialized interfaces from NASTRAN in support of multidisciplinary analysis. 2) Thermal analysis - IAC supports finite element and finite difference techniques for steady state or transient analysis. There are interfaces for the NASTRAN thermal analyzer, SINDA/SINFLO, and TRASYS II. FEMNET, which converts finite element structural analysis models to finite difference thermal analysis models, is also interfaced with the IAC database. 3) System dynamics - The DISCOS simulation program which allows for either nonlinear time domain analysis or linear frequency domain analysis, is fully interfaced to the IAC database management capability. 4) Control analysis - Interfaces for the ORACLS, SAMSAN, NBOD2, and INCA programs allow a wide range of control system analyses and synthesis techniques. Level 2.5 includes EIGEN, which provides tools for large order system eigenanalysis, and BOPACE, which allows for geometric capabilities and finite element analysis with nonlinear material. Also included in IAC level 2.5 is SAMSAN 3.1, an engineering analysis program which contains a general purpose library of over 600 subroutines for numerical analysis. 5) Graphics - The graphics package IPLOT is included in IAC. IPLOT generates vector displays of tabular data in the form of curves, charts, correlation tables, etc. Either DI3000 or PLOT-10 graphics software is required for full graphic capability. In addition to these analysis tools, IAC 2.5 contains an IGES interface which allows the user to read arbitrary IGES files into an IAC database and to edit and output new IGES files. IAC is available by license for a period of 10 years to approved U.S. licensees. The licensed program product includes one set of supporting documentation. Additional copies may be purchased separately. IAC is written in FORTRAN 77 and has been implemented on a DEC VAX series computer operating under VMS. IAC can be executed by multiple concurrent users in batch or interactive mode. The program is structured to allow users to easily delete those program capabilities and "how to" examples they do not want in order to reduce the size of the package. The basic central memory requirement for IAC is approximately 750KB. The following programs are also available from COSMIC as separate packages: NASTRAN, SINDA/SINFLO, TRASYS II, DISCOS, ORACLS, SAMSAN, NBOD2, and INCA. The development of level 2.5 of IAC was completed in 1989.

Thermal conductivity of III-V semiconductor superlattices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mei, S., E-mail: song.mei@wisc.edu; Knezevic, I., E-mail: irena.knezevic@wisc.edu

2015-11-07

This paper presents a semiclassical model for the anisotropic thermal transport in III-V semiconductor superlattices (SLs). An effective interface rms roughness is the only adjustable parameter. Thermal transport inside a layer is described by the Boltzmann transport equation in the relaxation time approximation and is affected by the relevant scattering mechanisms (three-phonon, mass-difference, and dopant and electron scattering of phonons), as well as by diffuse scattering from the interfaces captured via an effective interface scattering rate. The in-plane thermal conductivity is obtained from the layer conductivities connected in parallel. The cross-plane thermal conductivity is calculated from the layer thermal conductivitiesmore » in series with one another and with thermal boundary resistances (TBRs) associated with each interface; the TBRs dominate cross-plane transport. The TBR of each interface is calculated from the transmission coefficient obtained by interpolating between the acoustic mismatch model (AMM) and the diffuse mismatch model (DMM), where the weight of the AMM transmission coefficient is the same wavelength-dependent specularity parameter related to the effective interface rms roughness that is commonly used to describe diffuse interface scattering. The model is applied to multiple III-arsenide superlattices, and the results are in very good agreement with experimental findings. The method is both simple and accurate, easy to implement, and applicable to complicated SL systems, such as the active regions of quantum cascade lasers. It is also valid for other SL material systems with high-quality interfaces and predominantly incoherent phonon transport.« less

Hydraulic and Mechanical Effects from Gas Hydrate Conversion and Secondary Gas Hydrate Formation during Injection of CO2 into CH4-Hydrate-Bearing Sediments

NASA Astrophysics Data System (ADS)

Bigalke, N.; Deusner, C.; Kossel, E.; Schicks, J. M.; Spangenberg, E.; Priegnitz, M.; Heeschen, K. U.; Abendroth, S.; Thaler, J.; Haeckel, M.

2014-12-01

The injection of CO2 into CH4-hydrate-bearing sediments has the potential to drive natural gas production and simultaneously sequester CO2 by hydrate conversion. The process aims at maintaining the in situ hydrate saturation and structure and causing limited impact on soil hydraulic properties and geomechanical stability. However, to increase hydrate conversion yields and rates it must potentially be assisted by thermal stimulation or depressurization. Further, secondary formation of CO2-rich hydrates from pore water and injected CO2 enhances hydrate conversion and CH4 production yields [1]. Technical stimulation and secondary hydrate formation add significant complexity to the bulk conversion process resulting in spatial and temporal effects on hydraulic and geomechanical properties that cannot be predicted by current reservoir simulation codes. In a combined experimental and numerical approach, it is our objective to elucidate both hydraulic and mechanical effects of CO2 injection and CH4-CO2-hydrate conversion in CH4-hydrate bearing soils. For the experimental approach we used various high-pressure flow-through systems equipped with different online and in situ monitoring tools (e.g. Raman microscopy, MRI and ERT). One particular focus was the design of triaxial cell experimental systems, which enable us to study sample behavior even during large deformations and particle flow. We present results from various flow-through high-pressure experimental studies on different scales, which indicate that hydraulic and geomechanical properties of hydrate-bearing sediments are drastically altered during and after injection of CO2. We discuss the results in light of the competing processes of hydrate dissociation, hydrate conversion and secondary hydrate formation. Our results will also contribute to the understanding of effects of temperature and pressure changes leading to dissociation of gas hydrates in ocean and permafrost systems. [1] Deusner C, Bigalke N, Kossel E, Haeckel M. Methane Production from Gas Hydrate Deposits through Injection of Supercritical CO2. Energies 2012:5(7): 2112-2140.

Stomatal control and leaf thermal and hydraulic capacitances under rapid environmental fluctuations.

PubMed

Schymanski, Stanislaus J; Or, Dani; Zwieniecki, Maciej

2013-01-01

Leaves within a canopy may experience rapid and extreme fluctuations in ambient conditions. A shaded leaf, for example, may become exposed to an order of magnitude increase in solar radiation within a few seconds, due to sunflecks or canopy motions. Considering typical time scales for stomatal adjustments, (2 to 60 minutes), the gap between these two time scales raised the question whether leaves rely on their hydraulic and thermal capacitances for passive protection from hydraulic failure or over-heating until stomata have adjusted. We employed a physically based model to systematically study effects of short-term fluctuations in irradiance on leaf temperatures and transpiration rates. Considering typical amplitudes and time scales of such fluctuations, the importance of leaf heat and water capacities for avoiding damaging leaf temperatures and hydraulic failure were investigated. The results suggest that common leaf heat capacities are not sufficient to protect a non-transpiring leaf from over-heating during sunflecks of several minutes duration whereas transpirative cooling provides effective protection. A comparison of the simulated time scales for heat damage in the absence of evaporative cooling with observed stomatal response times suggested that stomata must be already open before arrival of a sunfleck to avoid over-heating to critical leaf temperatures. This is consistent with measured stomatal conductances in shaded leaves and has implications for water use efficiency of deep canopy leaves and vulnerability to heat damage during drought. Our results also suggest that typical leaf water contents could sustain several minutes of evaporative cooling during a sunfleck without increasing the xylem water supply and thus risking embolism. We thus submit that shaded leaves rely on hydraulic capacitance and evaporative cooling to avoid over-heating and hydraulic failure during exposure to typical sunflecks, whereas thermal capacitance provides limited protection for very short sunflecks (tens of seconds).

A sharp interface immersed boundary method for vortex-induced vibration in the presence of thermal buoyancy

NASA Astrophysics Data System (ADS)

Garg, Hemanshul; Soti, Atul K.; Bhardwaj, Rajneesh

2018-02-01

We report the development of an in-house fluid-structure interaction solver and its application to vortex-induced vibration (VIV) of an elastically mounted cylinder in the presence of thermal buoyancy. The flow solver utilizes a sharp interface immersed boundary method, and in the present work, we extend it to account for the thermal buoyancy using Boussinesq approximation and couple it with a spring-mass system of the VIV. The one-way coupling utilizes an explicit time integration scheme and is computationally efficient. We present benchmark code verifications of the solver for natural convection, mixed convection, and VIV. In addition, we verify a coupled VIV-thermal buoyancy problem at a Reynolds number, Re = 150. We numerically demonstrate the onset of the VIV in the presence of the thermal buoyancy for an insulated cylinder at low Re. The buoyancy is induced by two parallel plates, kept in the direction of flow and symmetrically placed around the cylinder. The plates are maintained at the hot and cold temperature to the same degree relative to the ambient. In the absence of the thermal buoyancy (i.e., the plates are at ambient temperature), the VIV does not occur for Re ≤ 20 due to stable shear layers. By contrast, the thermal buoyancy induces flow instability and the vortex shedding helps us to achieve the VIV at Re ≤ 20, lower than the critical value of Re (≈21.7), reported in the literature, for a self-sustained VIV in the absence of the thermal buoyancy. The present simulations show that the lowest Re to achieve VIV in the presence of the thermal buoyancy is around Re ≈ 3, at Richardson number, Ri = 1. We examine the effect of the reduced velocity (UR), mass ratio (m), Prandtl number (Pr), Richardson number (Ri) on the displacement of the cylinder, lift coefficient, oscillation frequency, the phase difference between displacement and lift force, and wake structures. We obtain a significantly larger vibration amplitude of the cylinder over a wide range of UR as compared to that in the absence of the thermal buoyancy.

MuSim, a Graphical User Interface for Multiple Simulation Programs

DOE Office of Scientific and Technical Information (OSTI.GOV)

Roberts, Thomas; Cummings, Mary Anne; Johnson, Rolland

2016-06-01

MuSim is a new user-friendly program designed to interface to many different particle simulation codes, regardless of their data formats or geometry descriptions. It presents the user with a compelling graphical user interface that includes a flexible 3-D view of the simulated world plus powerful editing and drag-and-drop capabilities. All aspects of the design can be parametrized so that parameter scans and optimizations are easy. It is simple to create plots and display events in the 3-D viewer (with a slider to vary the transparency of solids), allowing for an effortless comparison of different simulation codes. Simulation codes: G4beamline, MAD-X,more » and MCNP; more coming. Many accelerator design tools and beam optics codes were written long ago, with primitive user interfaces by today's standards. MuSim is specifically designed to make it easy to interface to such codes, providing a common user experience for all, and permitting the construction and exploration of models with very little overhead. For today's technology-driven students, graphical interfaces meet their expectations far better than text-based tools, and education in accelerator physics is one of our primary goals.« less

Progress on DCLL Blanket Concept

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wong, Clement; Abdou, M.; Katoh, Yutai

2013-09-01

Under the US Fusion Nuclear Science and Technology Development program, we have selected the Dual Coolant Lead Lithium concept (DCLL) as a reference blanket, which has the potential to be a high performance DEMO blanket design with a projected thermal efficiency of >40%. Reduced activation ferritic/martensitic (RAF/M) steel is used as the structural material. The self-cooled breeder PbLi is circulated for power conversion and for tritium breeding. A SiC-based flow channel insert (FCI) is used as a means for magnetohydrodynamic pressure drop reduction from the circulating liquid PbLi and as a thermal insulator to separate the high-temperature PbLi (~700°C) frommore » the helium-cooled RAF/M steel structure. We are making progress on related R&D needs to address critical Fusion Nuclear Science and Facility (FNSF) and DEMO blanket development issues. When performing the function as the Interface Coordinator for the DCLL blanket concept, we had been developing the mechanical design and performing neutronics, structural and thermal hydraulics analyses of the DCLL TBM module. We had estimated the necessary ancillary equipment that will be needed at the ITER site and a detailed safety impact report has been prepared. This provided additional understanding of the DCLL blanket concept in preparation for the FNSF and DEMO. This paper will be a summary report on the progress of the DCLL TBM design and R&Ds for the DCLL blanket concept.« less

Coupled Finite Element ? Potts Model Simulations of Grain Growth in Copper Interconnects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Radhakrishnan, Balasubramaniam; Gorti, Sarma B

The paper addresses grain growth in copper interconnects in the presence of thermal expansion mismatch stresses. The evolution of grain structure and texture in copper in the simultaneous presence of two driving forces, curvature and elastic stored energy difference, is modeled by using a hybrid Potts model simulation approach. The elastic stored energy is calculated by using the commercial finite element code ABAQUS, where the effect of elastic anisotropy on the thermal mismatch stress and strain distribution within a polycrystalline grain structure is modeled through a user material (UMAT) interface. Parametric studies on the effect of trench width and themore » height of the overburden were carried out. The results show that the grain structure and texture evolution are significantly altered by the presence of elastic strain energy.« less

User's manual for a two-dimensional, ground-water flow code on the Octopus computer network

DOE Office of Scientific and Technical Information (OSTI.GOV)

Naymik, T.G.

1978-08-30

A ground-water hydrology computer code, programmed by R.L. Taylor (in Proc. American Society of Civil Engineers, Journal of Hydraulics Division, 93(HY2), pp. 25-33 (1967)), has been adapted to the Octopus computer system at Lawrence Livermore Laboratory. Using an example problem, this manual details the input, output, and execution options of the code.

Interface For Fault-Tolerant Control System

NASA Technical Reports Server (NTRS)

Shaver, Charles; Williamson, Michael

1989-01-01

Interface unit and controller emulator developed for research on electronic helicopter-flight-control systems equipped with artificial intelligence. Interface unit interrupt-driven system designed to link microprocessor-based, quadruply-redundant, asynchronous, ultra-reliable, fault-tolerant control system (controller) with electronic servocontrol unit that controls set of hydraulic actuators. Receives digital feedforward messages from, and transmits digital feedback messages to, controller through differential signal lines or fiber-optic cables (thus far only differential signal lines have been used). Analog signals transmitted to and from servocontrol unit via coaxial cables.

A comparison of the CHF between tubes and annuli under PWR thermal-hydraulic conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Herer, C.; Souyri, A.; Garnier, J.

1995-09-01

Critical Heat Flux (CHF) tests were carried out in three tubes with inside diameters of 8, 13, and 19.2 mm and in two annuli with an inner tube of 9.5 mm and an outer tube of 13 or 19.2 mm. All axial heat flux distributions in the test sections were uniform. The coolant fluid was Refrigerant 12 (Freon-12) under PWR thermal-hydraulic conditions (equivalent water conditions - Pressure: 7 to 20 MPa, Mass Velocity: 1000 to 6000 kg/m2/s, Local Quality: -75% to +45%). The effect of tube diameter is correlated for qualities under 15%. The change from the tube to themore » annulus configuration is correctly taken into account by the equivalent hydraulic diameter. Useful information is also provided concerning the effect of a cold wall in an annulus.« less

Fatigue and Damage Tolerance Analysis of a Hybrid Composite Tapered Flexbeam

NASA Technical Reports Server (NTRS)

Murri, Gretchen B.; Schaff, Jeffrey R.; Dobyns, Al

2001-01-01

The behavior of nonlinear tapered composite flexbeams under combined axial tension and cyclic bending loading was studied using coupon test specimens and finite element (FE) analyses. The flexbeams used a hybrid material system of graphite/epoxy and glass/epoxy and had internal dropped plies, dropped in an overlapping stepwise pattern. Two material configurations, differing only in the use of glass or graphite plies in the continuous plies near the midplane, were studied. Test specimens were cut from a full-size helicopter tail-rotor flexbeam and were tested in a hydraulic load frame under combined constant axialtension load and transverse cyclic bending loads. The first determination damage observed in the specimens occurred at the area around the tip of the outermost ply-drop group in the tapered region of the flexbeam, near the thick end. Delaminations grew slowly and stably, toward the thick end of the flexbeam, at the interfaces above and below the dropped-ply region. A 2D finite element model of the flexbeam was developed. The model was analyzed using a geometrically non-linear analysis with both the ANSYS and ABAQUS FE codes. The global responses of each analysis agreed well with the test results. The ANSYS model was used to calculate strain energy release rates (G) for delaminations initiating at two different ply-ending locations. The results showed that delaminations were more inclined to grow at the locations where they were observed in the test specimens. Both ANSYS and ABAQUS were used to calculate G values associated with delamination initiating at the observed location but growing in different interfaces, either above or below the ply-ending group toward the thick end, or toward the thin end from the tip of the resin pocket. The different analysis codes generated the same trends and comparable peak values, within 5-11 % for each delamination path. Both codes showed that delamination toward the thick region was largely mode II, and toward the thin region was predominantly mode I. The calculated peak G-values from either analysis predict delamination is most likely to occur along the same interface where it was observed in the test specimens. Calculated peak G values were used with material characterization data to calculate a curve relating the fatigue life of the specimens, N, to the applied transverse load, V, for a given constant axial load.

Intercalated water layers promote thermal dissipation at bio-nano interfaces.

PubMed

Wang, Yanlei; Qin, Zhao; Buehler, Markus J; Xu, Zhiping

2016-09-23

The increasing interest in developing nanodevices for biophysical and biomedical applications results in concerns about thermal management at interfaces between tissues and electronic devices. However, there is neither sufficient knowledge nor suitable tools for the characterization of thermal properties at interfaces between materials of contrasting mechanics, which are essential for design with reliability. Here we use computational simulations to quantify thermal transfer across the cell membrane-graphene interface. We find that the intercalated water displays a layered order below a critical value of ∼1 nm nanoconfinement, mediating the interfacial thermal coupling, and efficiently enhancing the thermal dissipation. We thereafter develop an analytical model to evaluate the critical value for power generation in graphene before significant heat is accumulated to disturb living tissues. These findings may provide a basis for the rational design of wearable and implantable nanodevices in biosensing and thermotherapic treatments where thermal dissipation and transport processes are crucial.

HSTRESS: A computer program to calculate the height of a hydraulic fracture in a multi-layered stress medium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Warpinski, N.R.

A computer code for calculating hydraulic fracture height and width in a stressed-layer medium has been modified for easy use on a personal computer. HSTRESS allows for up to 51 layers having different thicknesses, stresses and fracture toughnesses. The code can calculate fracture height versus pressure or pressure versus fracture height, depending on the design model in which the data will be used. At any pressure/height, a width profile is calculated and an equivalent width factor and flow resistance factor are determined. This program is written in FORTRAN. Graphics use PLOT88 software by Plotworks, Inc., but the graphics software mustmore » be obtained by the user because of licensing restrictions. A version without graphics can also be run. This code is available through the National Energy Software Center (NESC), operated by Argonne National Laboratory. 14 refs., 21 figs.« less

Superior Thermal Interface via Vertically Aligned Carbon Nanotubes Grown on Graphite Foils

DTIC Science & Technology

2012-01-01

accepted 12 November 2012) In an attempt to study the thermal transport at the interface between nanotubes and graphene, vertically aligned multiwalled...tually increases the thermal barrier in a significant manner. On the other hand, thermal transport properties of thermal tapes and thermally conductive...aforementioned study achieved superior thermal transport properties, the processing and scale-up of the developed process would be prohibitively

Interfacial Thermal Conductance Limit and Thermal Rectification Across Vertical Carbon Nanotube/Graphene Nanoribbon-Silicon Interfaces

DTIC Science & Technology

2013-01-01

Interfacial thermal conductance limit and thermal rectification across vertical carbon nanotube/graphene nanoribbon-silicon interfaces Ajit K...054308 (2013) Investigation on interfacial thermal resistance and phonon scattering at twist boundary of silicon J. Appl. Phys. 113, 053513 (2013...2013 to 00-00-2013 4. TITLE AND SUBTITLE Interfacial thermal conductance limit and thermal rectification across vertical carbon nanotube/graphene

Uncertainty Quantification For Physical and Numerical Diffusion Models In Inertial Confinement Fusion Simulations

NASA Astrophysics Data System (ADS)

Rana, Verinder S.

This thesis concerns simulations of Inertial Confinement Fusion. Inertial confinement is carried out in a large scale facility at National Ignition Facility. The experiments have failed to reproduce design calculations, and so uncertainty quantification of calculations is an important asset. Uncertainties can be classified as aleatoric or epistemic. This thesis is concerned with aleatoric uncertainty quantification. Among the many uncertain aspects that affect the simulations, we have narrowed our study of possible uncertainties. The first source of uncertainty we present is the amount of pre-heating of the fuel done by hot electrons. The second source of uncertainty we consider is the effect of the algorithmic and physical transport diffusion and their effect on the hot spot thermodynamics. Physical transport mechanisms play an important role for the entire duration of the ICF capsule, so modeling them correctly becomes extremely vital. In addition, codes that simulate material mixing introduce numerical (algorithmically) generated transport across the material interfaces. This adds another layer of uncertainty in the solution through the artificially added diffusion. The third source of uncertainty we consider is physical model uncertainty. The fourth source of uncertainty we focus on a single localized surface perturbation (a divot) which creates a perturbation to the solution that can potentially enter the hot spot to diminish the thermonuclear environment. Jets of ablator material are hypothesized to enter the hot spot and cool the core, contributing to the observed lower reactions than predicted levels. A plasma transport package, Transport for Inertial Confinement Fusion (TICF) has been implemented into the Radiation Hydrodynamics code FLASH, from the University of Chicago. TICF has thermal, viscous and mass diffusion models that span the entire ICF implosion regime. We introduced a Quantum Molecular Dynamics calibrated thermal conduction model due to Hu for thermal transport. The numerical approximation uncertainties are introduced by the choice of a hydrodynamic solver for a particular flow. Solvers tend to be diffusive at material interfaces and the Front Tracking (FT) algorithm, which is an already available software code in the form of an API, helps to ameliorate such effects. The FT algorithm has also been implemented in FLASH and we use this to study the effect that divots can have on the hot spot properties.

The SENSEI Generic In Situ Interface

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ayachit, Utkarsh; Whitlock, Brad; Wolf, Matthew

The SENSEI generic in situ interface is an API that promotes code portability and reusability. From the simulation view, a developer can instrument their code with the SENSEI API and then make make use of any number of in situ infrastructures. From the method view, a developer can write an in situ method using the SENSEI API, then expect it to run in any number of in situ infrastructures, or be invoked directly from a simulation code, with little or no modification. This paper presents the design principles underlying the SENSEI generic interface, along with some simplified coding examples.

Pre-test analysis of protected loss of primary pump transients in CIRCE-HERO facility

NASA Astrophysics Data System (ADS)

Narcisi, V.; Giannetti, F.; Del Nevo, A.; Tarantino, M.; Caruso, G.

2017-11-01

In the frame of LEADER project (Lead-cooled European Advanced Demonstration Reactor), a new configuration of the steam generator for ALFRED (Advanced Lead Fast Reactor European Demonstrator) was proposed. The new concept is a super-heated steam generator, double wall bayonet tube type with leakage monitoring [1]. In order to support the new steam generator concept, in the framework of Horizon 2020 SESAME project (thermal hydraulics Simulations and Experiments for the Safety Assessment of MEtal cooled reactors), the ENEA CIRCE pool facility will be refurbished to host the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section to investigate a bundle of seven full scale bayonet tubes in ALFRED-like thermal hydraulics conditions. The aim of this work is to verify thermo-fluid dynamic performance of HERO during the transition from nominal to natural circulation condition. The simulations have been performed with RELAP5-3D© by using the validated geometrical model of the previous CIRCE-ICE test section [2], in which the preceding heat exchanger has been replaced by the new bayonet bundle model. Several calculations have been carried out to identify thermal hydraulics performance in different steady state conditions. The previous calculations represent the starting points of transient tests aimed at investigating the operation in natural circulation. The transient tests consist of the protected loss of primary pump, obtained by reducing feed-water mass flow to simulate the activation of DHR (Decay Heat Removal) system, and of the loss of DHR function in hot conditions, where feed-water mass flow rate is absent. According to simulations, in nominal conditions, HERO bayonet bundle offers excellent thermal hydraulic behavior and, moreover, it allows the operation in natural circulation.